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Date Declassified: February 12, 1959,

Consolidation of this material into compact form to permit
economical, direct reproduction has resulted in multiple fo-
lios for some pages, e.g., 10—-12, 27-29, etc.

LEGAL NOTICE

This report was prepared as an account of Government sponsored work, Neither the United
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A. Makes any warranty or representation, expressed or implied, with respect to the accu-
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This report has been reproduced directly from the best
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Printed in USA. Price $3.50. Available from the Office of
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25, D. C.

 

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ORNL-1845(Del.)

OPERATION OF THE AIRCRAFT REACTOR EXPERIMENT

By

W. B. Cottrell
H. E. Hungerford
J. K. Leslie

J. L. Meem

September 6, 1955

Work performed under Contract No. W-7405-Eng-26

Oak Ridge National Laboratory
Qak Ridge, Tennessee

 

MARTIN MARIETTA ENERGY SYSTEMS LIBRARIES

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ACKNOWLEDGMENTS ’

4 The experiment described herein was performed by the ARE Operations Group under

 

the immediate supervision of E, S, Bettis and J. L. Meem, with frequent support from
other groups in the ANP Project. In particular, W, R, Grimes and his crew effected the

carrier and concentrate loading operation, as well as the fluoride sampling; C., P.

* o

Coughlen and his crew effected the sodium loading and sampling; a group from the

Analytical Chemistry Division, under the supervision of J. C. White, performed all

 

 

 

analyses of fluoride and sodium samples; and E. B. Johnson obtained a reactor power E
calibration from analysis of a fuel sample. Craft foremen B, H. Webster and J. C, :;
Packard are both to be commended for the excellent craftsmanship in the construction :
and maintenance of the system, which made the operation of the system such a success; i
and all who were in the building during the operation are appreciative of the concern of E
the Health Physics Group under R. L. Clark. ’ :
The bulk of this report was prepared by the authors, but considerable assistance g
was received from W, K, Ergen in the analysis of the nuclear data. The work of . i
numerous others is referred to in the various analyses which are presented in the text t
and in the appendixes. In addition, the constructive criticism of W, H, Jordan, S. J.
Cromer, and E. S. Bettis, each of whom reviewed this report, is gratefully acknowledged.
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FOREWORD

The operation of the Aircraft Reactor Experiment (ARE) culminated four years of
endeavor af ORNL in the field of nuclear propulsion of aircraft, The final success
of the experiment is a tribute to the efforts of the 300-odd technical and scientific
personnel who constitute the ANP Project at ORNL, as well as those others who were
less prominently engaged in the fabrication, assembly, and installation of the experi-
ment. The project was capably directed toward this goal during its formative years by
Dr. R. C. Briant (now deceased), and subsequently by W, H. Jordan and S. J. Cromer,
currently Director and Co-Director, respectively.

This is the second in a series of three reports which summarize the ARE experience
and is concerned primarily with the nuclear operation of the reactor. The experiment
is described, and the data obtained during the time from the start of the critical experi-
ment until the reactor was shut down for the last time are analyzed. The material is
presented in essentially chronological order, and frequent reference is made to appended
material which includes important detailed data and information that may not be of

general interest.

 

 

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SUMMARY ..
1. INTRODUCTION ...

CONTENTS

..........................................................................................................................

..........................................................................................................................

2. DESCRIPTION OF THE REACTOR EXPERIMENT ... e

Reactor ..o,
Fuel System ........ccocoei.
Sodium System ....................
Process Instrumentation....
Nuclear Instrumentation an

Off-Gas System ...
Fuel Enrichment System ...

A oM IO S oo

3. PRENUCLEAR OPERATION oo s

Circulation of Sodium ........
Circulation of Fuel Carrier
Final Preparations for Nue

4. CRITICAL EXPERIMENT

Enrichment Procedure........

..........................................................................................................................

........................................................................................................................

[Ear OPErAHION ..o vv ittt

..........................................................................................................................

Early Stages of the Experiment ...
Subcritical Measurement of the Reactor Temperature Coefficient ...

Approach to Criticality......
Analyses of Fuel Samples

Calibration of the SHim RoOds oo e e e

Measurement of the Reactivity-Mass Ratio ...

5. LOW-POWER EXPERIMEN

I S ....................................................................................................................

Power Determination From Fuel ACtivation ..o oot et

Radiation Surveys .............

..........................................................................................................................

Regulating Rod Calibration from Fuel Additions ........................... e
Regulating Rod Calibration from Reactor Periods ...

Calibration of Shim Rods v

s Regulating Rod ...

Fuel System Charaeteristics ..ot
Effect of Fuel Flow on Reactivity ..o
Low-Power Measurement of the Temperature Coefficient ...
Adjustment of Chamber Position ...

Approach to Power ............

HIGH-POWER EXPERIMENTS ... s

..........................................................................................................................

Temperature Coefficient MEQSUrEMENtS ... ..ot e

Fuel and reactor tempera
Sodium temperature coeff
Moderator temperature co
Measurement of the Xenon

ture COBFFIC NS (oo
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efficient .o e e e
PoiSoning . ..cocooeiriiiiie el ettt

Power Determination from Heat Extraction ... RUUTITT

Reactor Kinetics ................
Reactor control by tempe

..........................................................................................................................

PAURE COBFFI T N oo e e

Startup on demand For POWET . ..o e

Effect of one dollar of re
Power cycling of reactor

A VY oo e e

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10

14
14

19
19
20
21

23
23
23
30
31
35
36
36

40
40
40
43
48
49
52
54
55
57

58
58
61
62
65
66
67
68
69
69
73
78
81

 

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REGC O FrANSIENES oo e ‘ 87
" Calculated power change resulting from a regulating rod movement ... 90
Reactor temperature differential as a function of helium blower speed............................... 90
The phenomenon of the 1ime 1ag ..o, 91
Reactivity effects of transients in the sodium system ... 92
Reactivity following @ SCPAM ... e 93
Final Operation and SRUtdOWN ... 95
7. RECOMMEND AT IONS et 97
APPENDIXES
B. SUMMARY OF DESIGN AND OPERATIONAL DATA e 107
DS CrIP ION oo e 107
MatEriQls oo e 111
REACOr PRy Sies oot e 114
S TG oo 115
Reactor Control .o e e e 116
System Operating Conditions ... o e, 120
MiSCEIIANEOUS ..o e e 123
C. CONTROL SYSTEMDESCRIPTION AND OPERATION .. 125
Control System Design ..o, 125
Instruments DesCriprion ..o 125
Controls DesCriPHOn ..o 126
Console and Control Board Description ... e 129
Control OPerations ..ot e 130
REAEEOr OP@ration .......ooo it 135
D. NUCLEAR OPERATING PROCEDURES ... e, 144
Addition of Fuel Concentrate ..o e, 144
Suberitical Experiments ... 145
Initial Criticality oo e e, 147
Rod Calibration vs Fuel Addition ... 147
Low-Power Experiments ..., 147
Approach to POWer ..o, e, 150
Experiments at Power ..., 152
E. MATHEMATICAL ANALYSIS OF APPROACH TO CRITICALITY .o, 156
F. COLD, CLEAN CRITICAL MASS .. oo e, 158
G. FLUX AND POWER DISTRIBUTIONS ..o oo, 160
Neutron Flux Distributions ..., 160
Fission-Neutron Flux Distributions oo 161
Power Distribution ..o e 162
H. POWER DETERMINATION FROM FUEL ACTIVATION ...l 165
R OrY e e 165
Experimental Procedure ... ..o e, 166
[. INHOUR FORMULA FOR A CIRCULATING-FUEL REACTOR WITH SLUG FLOW ... 168
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J. CALIBRATION OF THE SHIM RODS .o ittt ettt e 172
Calibration from Critical Experiment Data ..o 172
Calibration Against the Regulating Rod ..., 174
Calibration by Using the Fission Chambers ..., 177

K. CORRELATION OF REACTOR AND LINE TEMPERATURES ... 180

L. POWER DETERMINATION FROM HEAT EXTRACTION ..o 184

M. THERMODYNAMIC ANALY S S it oottt 188
Insulation Losses, Heater Power Input, and Space Cooler Performance ... 188
Experimental Yalues of Heat Transfer Coefficients ... 188

N. COMPARISON OF REACTOR POWER DETERMINATIONS ... 190

0. ANALYSIS OF TEMPERATURE COEFFICIENT MEASUREMENTS ... et 192
Importance of the Fuel Temperature Coefficient ... 192
Effect of Geometry in the ARE .. 192
Time Lag Considerations ..., e e 193
Suberitical Measurement of Temperature Coefficient ... 193
Low-Power Measurements of Temperature Coefficients of Reactivity ..., 197
High-Power Measurements of Temperature Coefficients of Reactivity ... 199

P. THEORETICAL XENON POISONING ... et 200

Q. OPERATIONAL DIFFICULTEIES oot e 202
E i MENT SYS ORI oottt 202
Process INStrumentation ..o e e e 204
Nueclear Instrumentation and Controls .o e 205
ANUNCIGEOTS oo ettt e s e e, 205
Heaters and Heater Comtrols o oo et ettt ettt s e 205
System COmPONENTS ... e e 205
L OK S oo e e 206

R. INTEGRATED POWER oo et 208
AP ACEEA P OWET oot et e e 208
N EAE P OWET o oo e oo ettt e et e et e e e et a e e 210

S. INTERPRETATION OF OBSERVED REACTOR PERIODS DURING TRANSIENTS ... 212

T NUCLE AR LG oo oo oottt et e ettt b ettt ettt b e sttt 213

U, THE ARE BUILDING.............co e bR R 232

B B L IO G R A PHY oot s 235

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OPERATION OF THE AIRCRAFT REACTOR EXPERIMENT

SUMMARY

The Aircraft Reactor Experiment (ARE) was oper-
ated successfully and without untoward difficulty
in November 1954. The following statements sum-
marize the notable information obtained from the
experiment.

1. The reactor became critical with a mass of
32.8 Ib of U?35, which gave a concentration of
23.9 Ib of U233 per cubic foot of fluoride fuel.
For operation at power, the U233 content of the
fuel mixture was increased to 26.0 Ib/ft3, and thus
the final composition of the fuel mixture was 53.09
mole % NaF, 40.73 mole % ZrF,, and 6.18 mole %
UF,.

2. The maximum power level for sustained oper-
ation was 2.5 Mw, with a temperature gradient of
355°F; the maximum fuel temperature at this level
was 1580°F. Temperatures as high as 1620°F were
recorded during transients.

3. From the time the reactor first went critical
until the final shutdown, 221 hr had elapsed, and
for the final 74 hr the power was in the megawatt
range (0.1 to 2.5 Mw). The total integrated power
was about 96 Mw-hr.

4. While at power the reactor exhibited excellent
stability and it was easily controlled because of
its high negative temperature coefficient of re-
activity, which made the reactor a slave to the

load placed upon it. The fuel temperature coef-
ficient was -9.8 x 10™° (Ak/k)/°F, and the over-
all coefficient for the reactor was -6.1 x 1077,

5. Practically all the gaseous fission products
and probably some of the other volatile fission
products were removed from the circulating fuel.
In a 25-hr run at 2.12 Mw the upper limit of the
reactor poisoning due to xenon was 0.01% Ak/k.
No more than 5% of the xenon stayed in the molten
fluoride fuel.

6. The total time of operation at high tempera-
ture {1000 to 1600°F) for the sodium circuit was
635 hr, and, for the fluoride fuel system, 462 hr.
During most of the operating period the sodium
was circulated at 150 gpm and the fuel at 46 gpm.

7. The fabricability and compatibility of the
materials system, i.e., fluoride fuel, sodium coolant,
and Inconel structure, were demonstrated, at least
for the operating times, temperatures, and flux
levels present.

8. All components and, with few exceptions, all
instrumentation performed according to design
specifications. The performance of the pumps was
particularly gratifying, and the low incidence of
instrumentation failure was remarkable in view of
the quantity and complexity of the instruments
used.

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1. INTRODUCTION

The Aircraft Nuclear Propulsion (ANP) project
at the Oak Ridge National Laboratory was formed
in the fall of 1949, at the request of the Atomic
Energy Commission, to provide technical support
to existing Air Force endeavors in the field. The
ORNL effort gradually expanded and, following the
recommendation of the Technical Advisory Board
in the summer of 1950, was directed toward the
construction and operation of an aircraft reactor
experiment. A complete description of the ARE
falls naturally into three categories that correspond
to the three phases of the project: (1) design and
installation, (2) operation, and (3) postoperative
examination. Each of these phases is covered by
a separate report, ORNL-1844, ORNL-1845, and
ORNL-1868, respectively. Much detailed infor-
mation pertaining to the selection of the reactor
type and to the design, construction, and pre-
nuclear operation of the reactor experiment will be
presented in ORNL-1844. As the title of this
report (ORNL-1845) indicates it is concerned
primarily with the operation of the experiment, and
only insofar as they are necessary or useful to the
understanding or evaluation of the nuclear oper-
ation are design and preliminary operational data
included herein. The third report {ORNL-1868)
will describe the aftermath of the experiment, with
particular reference to corrosion, radiation effects,
and the decay of activity ~ effects that cannot be
evaluated at this time because of the high level of
the radioactivity of the equipment.

The specific operating objectives were to
attain a fuel temperature of 1500°F, with a 350°F
temperature rise across the reactor, and to operate
the system for approximately 100 Mw-hr. Other
objectives of the experiment were to obtain as
much experimental data as possible on the reactor
operational characteristics. The extent to which
each of these objectives was fulfilled is described
herein, and a measure of the success of the program
is thus provided.

Although it was initiclly planned to use a

sodium-cooled, solid-fuel-element reactor, the
reactor design evolved first to that of a sodium-
cooled, stationary-liquid-fuel reactor and, finally,
to that of a circulating-fuel reactor employing
sodium as a reflector coolant. These evolutionary
processes left their mark on the experiment, par-
ticularly in that the reactor had to incorporate a
moderator geometry that was originally specified
and ordered for the sodium-cooled reactor. The
adaptation of this moderator geometry to the
circulating-fuel reactor resuited in a reactor in
which the fuel stream was divided into six parallel
circuits, each of which made numerous passes
through the core. These fuel passages were not
drainable — a condition which caused considerable
concern throughout the course of the experiment.

Although it is not the purpose of this report to
give a detailed history of the design, construction,
or preliminary testing of the reactor system, the
final design and pertinent prenuclear operation are
briefly described. The bulk of the report concerns
the operation of the experiment from the time
uranium was added to the fuel system on October 30
until the evening of November 12, when the reactor
was shut down for the last time. In addition to
the description of the various experiments and
onalyses of the data which are presented in the
body of the report, the entire nuclear operation,
as recorded in the “Nuclear Log,'” is given in
Appendix T. The report also includes a number
of recommendations based on the operating ex-
perience, and much detailed supporting data and
information not appropriate for inclusion in the
body of the report are given in the other ap-
pendixes.

The various experiments that were performed

on the ARE were designated as E, L, or H series,
depending upon whether they occurred during the
critical experiment, low-power operation, or high-
power operation, respectively. Each of these
experiments, as listed below, is discussed in
this report.

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Experiment
Series No.
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E.2
L-1
L-2
L-3
L-4
L-5
L-6
L-7
L-8
L-9
H-1
H-2
H-3
H-4
H-5
H-6
H-7
H-8
H-9
H-10
H-11
H-12
H-13
H- 14

Type of Experiment

Critical experiment

Subcritical measurement of reactor temperature coefficient
Power determination at 1 w {nominal)

Regulating rod calibration vs fuel addition

Fuel system characteristics

Power determination at 10 w (nominal)

Regulating rod calibration vs reactor period

Calibration of shim rod vs regulating rod

Effect of fuel flow on reactivity

L ow-power measurement of reactor temperature coefficient
Adjustment of chamber position

Approach to power: 10-kw run

Test of off-gas system

Approach to powef: 100-kw to 1-Mw runs

High-power measurement of the fue! temperature coefficient
High-power measurement of the reactor temperature coefficient
Reactor startup on temperature coefficient

Sodium temperature coefficient

Effect of a dollar of reactivity

High-power measurement of reactor temperature coefficient
Moderator temperature coefficient

Xenon run at full power

Reéc_fiv'ity effecis‘ of sodium flow

Xenon buildup at. one-tenth full pofier

Operation at maximum power

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2. DESCRIPTION OF THE REACTOR EXPERIMENT

The ARE consisted of the circulating-fuel
reactor and the associated pumps, heat transfer
equipment, controls, and instrumentationt required
for its safe operation. A schematic arrangement of
the reactor system is shown in Fig. 2.1. The
major functional parts of the system are discussed
briefly below. The physical plant is described in
Appendix U. A detailed description of the reactor
and the associated system may be found in the
design and installation report.’ A summary of the
design and operational data, including a detailed
flow sheet of the experiment, is given in Ap-
pendix B.

REACTOR

The reactor assembly consisted of a 2-in.-thick
Inconel pressure sheli in which beryllium oxide
moderator and reflector blocks were stacked around
fuel tubes, reflector cooling tubes, and control
assemblies. Elevation and plan sections of the
reactor are shown in Figs. 2.2 and 2.3. The
innermost region of the lattice assembly was the
core, which was a cylinder approximately 3 ft in
diameter and 3 ft fong. The beryllium oxide was
machined into small hexagonal blocks which were
split axially and stacked to effect the cylindrical
core and reflector. Each beryliium oxide block in
the core had a 1.25-in. hole drilled axially through

 

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periment, ORNL-1844 (to be issued).

  

 
   

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its center for the passage of the fuel tubes. The
outer 7.5 in. of beryllium oxide served as the
reflector and was located between the pressure
shell ond the cylindrical surface of the core. The
reflector consisted of hexagonal beryllium oxide
blocks, similar to the moderator blocks, but with
0.5-in. holes.

The fuel stream was divided into six parallel
circuits at the inlet fuel header, which was located
above the top of the core and outside the pressure
shell. These circuits each made 1] series passes
through the core, starting close to the core axis,
and progressing in serpentine fashion to the
periphery of the core, and finally leaving the core
through the bottom of the reactor. The six circuits
were connected to the outlet header. Each tube
was of 1.235-in.-OD seamless Inconel tubing with
a 60-mil wall. The combination of parallel and
series fuel passes through the core was largely
the result of the need for assuring turbulent flow
in a system in which the fluid properties and tube
dimensions were fixed.

The reflector coolant, i.e., sodium, was admitted
into the pressure shell through the bottom. The
sodium then passed up through the reflector tubes,
bathed the inside walls of the pressure shell,
filled the moderator interstices, and left from the
plenum chamber ot the top of the pressure shell.
The sodium, in addition to cooling the reflector
and pressure shell, acted as a heat transfer medium

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in the core by which moderator heat was readily
transmitted to the fuel stream.

FUEL SYSTEM

The fuel was a mixture of the fluorides of sodium
and zirconium, with sufficient uranium fluoride
added to make the reactor critical. While the fuel
ultimately employed for the experiment was the
NaF-ZrF4-UF4 mixture with a composition of

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

53.09-40.73-6.18 mole %, respectively, most pre-
liminary experimental work (i.e., pump tests,
corrosion tests) employed a fuel containing some-
what more UF,. The fuel was circulated around
a closed loop from the pump to the reactor, to the
heat exchanger, and back to the pump. An isometric
drawing of the fuel system is given in Fig. 2.4.

The fuel pump was a centrifugal pump with a
vertical shaft and a gas seal, as shown in Fig. 2.5.

 
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

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SCALE IN INGHES

Fig. 2.2. The Reactor (Elevation Section).

 

 

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SCALE IN INCHES

Fig. 2.3. The Reactor (Plan Section).

The fuel expansion volume around the impeller
cavity provided the only liquid-to-gas interface
in the fuel system. While speeds up to 2000 rpm
could be attained with the 15-hp d-c pump motor,
the desired fuel flow of 46 gpm was attained at
a speed of 1080 rpm. Although only one pump
was used in the experiment, a spare fuel pump,
isolated from the operating pump and in parallel
with it, was provided.

From the pump the fuel flowed to the reactor,’

where it was heated, then to two parallel fuel-to-
helium heat exchangers, and back to the pump.
The cycle time was about 47 sec at full flow, of

which approximately 8 sec was the time required
for the fuel to pass through the core. The two
fuel-to-helium heat exchangers were each coupled
to a helium-to-water heat exchanger. The heat
extracted from the fuel was transferred via the
helium to water, and the water — the ultimate
heat sink ~ was discharged. The helium flow rate
in the fuel-to-helium heat exchanger loop was
controlled through a magnetic clutch that coupled
the blower to a 50-hp motor. Control of the helium
tlow rate in this manner permitted smooth control
at any reactor power at which the heat generation
was great enough for the temperature coefficient
to be the controlling factor. '

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As shown in Fig. 2.4, the fuel system was
connected with two fill tanks (only one of which
was used) and one dump tank which had provisions
for removing the afterheat from the fuel.

The relatively high melting point of the circu-
lating fuel (about 1000°F for the NaF-ZrF ,-UF,
fuel containing 6.18% UF,) required that all
equipment within which it was circulated be
heated sufficiently to permit loading, unloading,
and low-power operation. This heating was ac-
complished by means of electrical heaters attached
to all components of the fuel and sodium systems;
i.e., pressure shell, heat exchanger, pumps, and
tanks, as well as all fuel and sodium piping.

In addition to the heaters and insulation, all
fuel and sodium piping was surrounded by a 1- to
]l/2-in. annulus (inside the heaters) through which
helium was circulated. The helium circulated
through the annuli was monitored at various
stations around the system for evidence of leaks
and was, in addition, a safety factor in that it
could be expected to keep hot any spots at which
heater failures occurred.

SODIUM SYSTEM

The sodium circuit external to the reactor,
shown in Fig. 2.6, was similar to that of the fuel.
The sodium flowed from pump to reactor, to heat

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exchanger, to pump. The sodium pumps were
- gas-sealed centrifugal pumps of the same design
as the fuel pump, except that a smaller expansion
volume was provided. Again, there were two sodium
pumps, one of which was a spare in parallel with
the operating pump.

The sodium was circulated through the two
parallel sodium-to-helium heat exchangers after
being heated in the reactor. Again the heat was
transferred via the helium to water in two helium-
to-water heat exchangers. The sodium system was
equipped with heaters, as was the fuel system,
and all piping was surrounded with the helium
annulus for leck monitoring and heat distribution.

PROCESS INSTRUMENTATION

Since a basic purpose of the ARE was the
acquisition of experimental data, the importance
of complete and reliable instrumentation could
not be overemphasized. Therefore there were
27 strip temperature recorders (mostly multipoint),
5 circular temperature recorders, 7 indicating flow
controliers, 9 temperature indicators (with up to
96 points per instrument), about 50 spark-plug
level indicators, 20 ammeters, 40 pressure gages,

16 pressure regulators, and 20 pressure transmitters.

In addition there were numerous flow recorders,
indicators, alarms, voltmeters, tachometers, and
assorted miscellaneous instruments.

Most ARE process instrumentation was installed
to permit observing and recording rotational speeds,
flow rates, temperatures, pressures, or liquid
levels. The operating values of temperature, pres-
sure, and flow at various stations around the ARE
fluid circuits are given in Appendix B. Since most
commercially available instruments for measuring
flow and pressure are subject to temperature
limitations considerably lower than the minimum
operating temperature of the ARE and they employ
open lines in which the ZrF, vapor could con-
dense, they were not suitable for ARE applications.
Therefore a bellows type of device was adapted
for use as a pressure indicator, and a fluid-
immersed inductance type of instrument was
developed for measuring sodium and fuel liquid
level and fuel flow.2 Sodium flow was measured
by an electromagnetic flowmeter. Conventional
chromel-alumel thermocouples welded to the pipe
walls were used to determine temperatures. The

 

2Fqy greater detail see ANP Quar. Prog. Rep. Sept.
10, 1952, ORNL-1375, p 23.

10

instrumentation flow sheet is shown in Fig. 2.7.
Most of the indicating and recording instruments
were located in the control room. While numerous
important temperatures were recorded in the control
room, about 75% of the thermocouples were indi-
cated (or recorded) only in the basement.

NUCLEAR INSTRUMENTATION AND CONTROLS

Detailed information on the important nuclear
instrumentation and controls is presented in Ap-
pendix C. Briefly, the nuclear instrumentation
(such as fission chambers and ion chambers), the
electronic components (such as preamplifiers and
power amplifiers), and the control features of the
ARE were similar, and in many cases identical,
to those of the MTR, the LITR, and other reactors
now in operation. However, since the fission
chambers were located in a high-temperature
region within the reflector, it was necessary to
develop  special high-temperature chambers.
Helium was used to cool them to below 600°F.

The locations of the two fission chambers in the
reactor may be seen from examination of Fig. 2.3;
the locations of the remaining ion chambers are
shown on Fig. 2.8. Since the fission chambers
moved through sleeves in the reactor which were
parallel to those for the regulating and shim rods,
it was convenient to group the drive mechanisms
with those for the rods in an igloo outside the
shielding above the reactor, as shown in Fig. 2.8.
All other ion chambers were external to the reactor
pressure shell; they viewed the reactor at mid-
plane; and they were mounted so that they would
move horizontally. These chambers included a
BF3 counter (for the critical experiment), two
paraliel circular plate (PCP) chambers, and two
compensated chambers.

There were two separate elements to the control
system - the single regulating rod and the three
shim rods. The locations of these rods in the
reactor may be seen from examination of Fig. 2.3.
The regulating rod had a vertical movement of
12 in. about the center of the reactor and was
fabricated of stainless steel. Several such regu-
lating rods, with varying amounts of metal, were
fabricated, and the one finally used in the experi-
ment had a total value of 0.4% Ak/k for the 12-in.
movement and could be moved at a rate of 0.011%

(AR/R)/sec.

The three safety rods were located in the core
on 120-deg points at a radius of 7.5 in. from the

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center of the reactor. Each rod was made up of
slugs of a hot-pressed mixture of boron carbide
and iron; the slugs were canned in stainless steel.
The canned sections were slipped over a flexible
tube. The total rod travel was 36 in., and 13 min
was required for withdrawal of the rod. Thus, the
5.8% Ak/E in each rod (total for the three was
approximately 17% At/E) could be withdrawn at
a rate of 0.45% (Ak/k)/min per rod.

A small amount of the helium used to cool the
fission chambers was directed through the shim
and regulating rod holes, but the helium flow was
so low that not much cooling was effected. This
helium was circulated in an independent circuit,
known as the rod cooling system, which had two
two-speed helium blowers in parallel and three
helium-to-water heat exchangers. The helium
circulated from the blowers through the rod sleeves
and fission chamber sleeves to the heat ex-
changers and back to the blowers, and it removed,
in the process, up to 35 kw of reactor heat. Part
of the helium returning from the reactor was
diverted through pipe annuli before reaching the
heat exchanger.

OFF-GAS SYSTEM

The off-gas system was designed to permit the
collection, holdup, and controlled dischgrge of
the radioactive fission products that were evolved
as a consequence of reactor operation. Although
only the gases above the fuel system were ex-
pected to have significant activity, the gas from
the sodium system was also discharged through
the off-gas system. As shown in Fig. 2.9, there
was, in addition to the primary off-gas system, an
auxiliary system for discharging gases from the
pit through nitrogen-cooled charcoal tanks.

In the primary system the off gases from both
the fuel and the sodium were directed to a common
vent header. From this header, helium plus the
volatile fission gases passed through a NaK
scrubber, which removed bromine and iodine, and
then into two holdup tanks, which were connected
in series, to permit the decay of xenon and krypton.
From these the gases were released up the stack.
The release of gases to the stack was dependent

14

upon two conditions: (1) a wind velocity greater
than 5 mph and (2) radioactivity of less than
0.8 pc/ecm®. The activity was sensed by a monitor
which was located between the two holdup tanks.

FUEL ENRICHMENT SYSTEM

The fluoride fuel used in the ARE was amenablie

to a convenient enrichment technique in which the

fuel system was first filled with a mixture of the
fluorides of sodium and zirconium. When this
mixture, NaZrFs, had been circulated for a suf-
ficient time to ascertain that the reactor was ready
to be taken critical (no leaks, etc.), uranium in
the form of molten Na,UF, was added to the
fluorides then in the system. The temperature
contour diagram of the NaF-ZrF -UF, system is
shown in Fig. 2.10. The melting point of Na,UF
was 1200°F, and there were no mixtures of higher
melting points on the join between Na,UF, and
NaZrF ., which is shown in Fig.2.11. The addition
of Na,UF raised the melting point of the mixture
only slightly above that of NaZ¢F . (955°F).

It was initially intended that the enrichment
operation would consist of the remote addition of
Na,UF, to the system from a large tank which
contained all the concentrate. From the large
tank, the concentrate was to pass through an
intermediate transfer tank which would transfer
about 1 gt or less at a time. The intermediate
tank was suspended from a load beam so that the
weight of the concentrate could be determined
before each addition of enriched material. However,
this system was discarded when preoperational
tests proved the temperature control of the system
to be inadequate; in addition, the accuracy of the
weight-measuring instrumentation on the transfer
tank was uncertain.

In lieu of the original enrichment system, a less
elaborate, more direct method of concentrate ad-
dition was employed. The enrichment procedure
actually used involved the successive connection
of numerous small concentrate containers to an
intermediate transfer pot, which, in turn, was
connected to the fuel system by a line which
injected the concentrate into the pump tank above
the liquid level.

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15

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-,

 

 

 

 

 

 

 

 

 

 

 

 

DWG. 244534
780
30
0 . 20
N o\ ?oo 675 |
>’22> 625 650 S
A 5 X % e S
OO ANNCES NGy [ 7T A E N,
/NN NN AN e e 0D A SN SN N\
e A WL NGNS A TN AN
o 7 s wio Ceps O A * % ° ’
NazZrF, NapZrFy NaZrFg NagzZrgFyg

Fig. 2.10. The System NaF-Z¢F .UF .

17

 

 

e e n -

e e e

- ¢ g T e e e

v T Ry TR

 
 

 

 

 

 

 

 

675 ORNL—LR-DWG 6223
‘. L
. ’ 4/ b
650 s ~® b
./ -

L~ y
/ o o
N ® L
625 g
® ~ &
&
£ 600 | ) .
w / :
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o ® b
= ¥
<L &
x

& 575 / s
s 7/ ’ .
= / £

o

 

 

550
e

s

\

 

 

 

 

 

 

 

 

 

 

 

 

 

®
525 ° .
/ w
® E
b
500 *
NaZrFy 40 20 30 40 50 60 70 80 90 Na,UFg s

Na,UFg (mole D}

-y

Fig. 2.11. The Psevdo-Binary System NaZrFs-NazUFé.

 

<

 

18

   
RT NI RERE Ty

o Ak bR -

Lo G

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Lo

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=

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3. PRENUCLEAR OPERATION

It is not the purpose here to describe in detail
the preliminary checks and shakedown tests which
preceded the nuclear operation of the system be-
cause this information is covered in the design
and installation report.! However, some under-
standing of the scope and significance of these
“preliminary’’ tests is desirabie, particularly since
at the conclusion of this phase of the operation
the fuel and sodium systems were left filled with
clean carrier and sodium, respectively, and the
fuel system was ready for the fuel enrichment
operation which comprised the critical experiment.

The preliminary tests accomplished many pur-
poses, the more important being the cleaning and
leak-checking of the systems, performance and
functional testing of components, checking and
calibrating of instruments, and the practicing of
operational techniques. The system, as an integral
unit, was sufficiently assembled by August 1 for
the operating crews to be placed on a three-shift
basis so that intensive tests on the various sys-
tems, initially with water as the circulating fluid,
could be started. There were, however, frequent
delays while changes and modifications that the
testing indicated as being desirable or necessary
were made. The operation with sodium, in par-
ticular, was responsible for major changes in the
sodium vent system and the elimination of the
sodium purification system. Tests with sodium
and carrier in the systems were concluded on
October 30, at which time the system was ready
for the critical experiment.

CIRCULATION OF SODIUM

By the last week in September, all temperature
tests preliminary to operation of the sodium system
had been completed. It was necessary to operate
the sodium system before operating the fuel system
because the sodium was used to heat the reactor
to above the fuel melting point. In preparation for
filling the sodium system, the temperature of the
enfire system was brought to 600°F and sodium
was transferred from portable drums into the
system fill tanks. The sodium used had been
carefully filtered to minimize the oxygen content,
which at the time of the filling averaged about
0.025 wt %.

 

‘Desz’grz and Installation of the Aircraft Reactor Ex-
periment, ORNL-1844 (to be issued).

The system was filled with sodium on the morning
of September 26. No particular difficulties were
encountered during the filling or when the sodium
pumps were operated at design speed, although
there was evidence of trapped gas in the system.
The electromagnetic flowmeters, which were at
first inoperative, functioned properly after suf-
ficient time had elapsed for wetting of the pipe
wall to occur. On the following afterncon o ledk
developed in a tube bend in the sodium purification
system, so the sodium was dumped.

Subsequent analyses revealed that the leak
occurred where a thermocouple pad had been
heliarc welded to the tube. The excessive weld
penetration which caused the leak was attributed
to a lack of instruction to the welder that a vari-
ation in pipe wall thickness existed in this section.

As a consequence of this leak and difficulties
experienced during draining of the sodium, a
number of changes were made in the sodium sys-
tem. The purification systems were eliminated
and provisions were made for better heat control
for all vent lines and valves to prevent freezing
of the sodium and piugging of the iines. [t was
concluded that the sodium purification system
could be removed without endangering the experi-
ment because experience in operating the system
had proved that the oxygen contamination was very
low; furthermore, there was no danger of the oxide
plugging the heat exchanger tubes because of the
low temperature differential in the system and the
large diameters of the tubes in the heat exchangers.
[t was felt, also, that the purification system
represented the weakest link in the fluid circuit
because thinner walled tubing had been used there
than anywhere else in the system, except for the
reactor tubes, which were assembled with extreme
care. Even so, the thin-walled tubing in the puri-
fication system would probably not have presented
a problem had the thermocouples been properly

welded.
On October 16 the sodium was recharged into

the system at 600°F, circulated, and dumped four
times to thoroughly check the operability of the
sodium system. Analyses of the sodium taken
both at 600 and 925°F showed the oxygen content
to be satisfactorily low (of the order of 0.026 wt %),
and therefore the initial batch of sodium was not
replaced. The sodium was circulated for several
days at 1300°F before fuel carrier, NaZrF, was

19

e

- R ——

b

T -

e

 

o TR

 

 
 

 

 

added to the fuel tank. During the period of sodium
circulation, the typical system characteristics
were those given in Figs. 3.1 and 3.2. The data
were comparable to those obtained during the water
tests.] Such discrepancies as existed were within
experimental error and could easily have been due
to changes in the system that were made during
the time between the two tests, such as the re-
moval of the purification systems and the removal
of the bypass around the reactor during the water
tests,

During and after the second sodium loading
operation the copper gauze in the helium stream
was monitored and no evidence of sodium leakage
was found. With helium in the fuel system the

. temperatures in the fuel and sodium systems were

ORNL-LR—-DWG 6390C

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

  
  
    

 

       
  
  

 

 

 

 

 

 

2800 ‘ ; l
2400 ; | ;
MAIN No PUMP !
TEST $-2, OCT. 20 :
= 2000 |- i e // T
S f B
had i \ i
8 1800 [ b e
ut )
a i i : :
@ 4200 [ b |
= ‘ i
S \ : !
o i : ! |
800 [T e e <
; : . . \ : '1 !
400 e S
! ! :
' i
i i i
0 f '
0 20 40 60 80 {00 120 140 160 180 200
FLOW RATE {gpm)
Fig. 3.1. Pump Speed vs Sodium Flow Rate.
56 ‘ ) ORNL-LR-DWG 6394
| MAIN Na PUMP )
48[ TESTs-2, 0cT. 20,1954 |
| PRESSURE HEAD MEASURED |
= 40 [-—— FROM PUMP SUCTION TO ]
= | REACTOR DISCHARGE |
31332 - J[—f—%w—w——l——a‘—fl -
z . | -
% 24| DATA CORRECTED FOR _{ |/~ /7 ,
a GALCULATED INSTRUMENT o t !
i ZERO SHIFT ; o  T-ACTUAL DATA
o 56 b ¢ -+ e m e -1
| A
i
8 e i - ' o
| ! Lt 4 { j {
§ L : i i
ol 1 T | |
O 20 40 60 80 100 120 140 160 180 200

FLOW RATE (gpm)

Fig. 3.1. Prerssure Head vs Sodium Flow Rate.

20

then gradually raised (about 10°F/hr) to 1300°F.
After the 1300°F temperature was reached, the
helium pressure in the fuel system was reduced
to 0.5 psig and sufficient krypton was added to
bring the system pressure up to 5.0 psig. The
helium in the annuli was then sampled and analyzed
for krypton, and no evidence of a leak was found.

CIRCULATION OF FUEL CARRIER

With the system isothermal at 1300°F, the fuel
carrier, NchFs, was loaded into the fill tank at
1200°F, and the fuel system was evacuated to
facilitate raising the fluoride into the system.
This filling operation, which occurred on October
25, required only a few minutes once the desired
vacuum was attained. During the filling operation
it was possible to follow the progress of the
fluoride through the system by watching the thermo-
couple indications along the pipe because of the
100°F temperature difference between the initial
fluoride temperature and the system temperature.
In contrast to the sodium system, there did not
appear to be any gas trapped in the fuel system
after loading; evidence of this was that there was
no liquid level change when the pressure in the
pump was changed, and the temperature differential s
across the six parallel fuel passages in the reactor
changed simultaneously when a temperature pertur-
bation was put into the system.

Five samples of the carrier were taken, and the
analytical results from each showed the concen-
trations of impurities and corrosion products given
in Table 3.1. '

There was twice as much carrier available as
was required to fill the system, and if chemical
analyses of samples of the carrier after it had
been circulated for about 50 hr had indicated too
high a corrosion-product buildup {as could have
resulted if the system had not been adequately
cleaned), the carrier in the system would have
been dumped and replaced with the extra carrier.
However, from the very favorable analyses of the
first two samples it was apparent that replacing
the carrier would not be necessary. (The chromium
content would have had to have been 500 ppm to
have necessitated the replacement. The corrosion
mechanism that defines the upper limit of the
chromium content is described in detail in ORNL-
1844.") |

During the time the carrier was being circulated
and before the enrichment operation commenced,

the system characteristics were determined. The

B

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.

vy

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T A T e e R N ey W

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b ki

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TABLE 3.1. ANALYSES OF IMPURITIES AND CORROSION PRODUCTS IN CARRIER SAMPLES

 

" Running Time

Impurities and Corrosion Products Found (ppm)

 

 

Date Time (he) - e "
10/25 1650 9.7 90 <5 50
10/26 0214 19.1 81 <5 40

1911 36.1 81 <5 6
10/27 1919 60.2 90 18 12
10/29 0935 98.5 102 30 20

 

ORNL-LR-DWG 6392

1200 I/

1000 - T
MAIN FUEL PUMP i

TEST -3, OCT. 25, 1954 !
800 i o

/o/
600 »

400 /

 

 

 

 

 

PUMP SPEED (rpm)

 

 

200 t

 

 

 

 

 

 

 

 

0 10 20 30 40 50 60
FLOW RATE (gpm)

Fig. 3.3. Pump Speed vs Fuel Carrier Flow
Rate.

data obtained in terms of pump speed vs flow and
system head vs flow are given in Figs. 3.3 and
3.4, respectively. These data were considerably
different from those to be expected from an ex-
trapolation of the water data. The discrepancies
are analyzed in more detail in ORNL-1844,' but
the only completely satisfactory explanation is
that the system head during the water test was too
high, possibly due to the temporary water Rotameter
installation or to one or more of the reactor tubes
not being completely filled so that only partial
flow was obtained through the reactor during the
water test, '

FINAL PREPARATIONS FOR
NUCLEAR OPERATION

By this time the neutron source had been placed
in the reactor and the nuclear instrumentation

ORNL-LR-DWG 8393

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

MAIN FUEL PUMP
TEST C-3, OCT. 25, 1954 /
o
25 i E 7
R PRESSURE HEAD MEASURED FROM ./
+ 20 = PUMP SUCTION TO REACTOR INLET.| /
& (PRESSURE ZERO ERROR< 0.5 psi)
w {5 /
« /
2 °
0 /
Ll
& o
10 e
*
5 .
0
0 10 20 30 40 50 60

FLOW RATE (gpm)

Fig. 3.4. Pressure Head vs Fuel Carrier Flow
RQ'I'e.

checked out. A BF, counter was installed in one
of the instrument holes external to the reactor in
order to check the fission chambers during the
critical experiment. Pulse height and voltage
curves were measured on all three fission chambers
(the two regulars plus a spare) in order to establish
the operational plateau for each chamber.

The mechanical opergtion tests of the regulating
rod and of the three safety (shim) rods were made;
in all, over 25 rod drop tests were performed in
order to ensure the reliability of the rod insertions
at high (1200 to 1300°F) temperatures. The rod
drives were also continuously cycled in order to
check the performance of the gear trains and

21

 

 
 

 

 

 

 

motors. During this time the rod magnet faces
were cleaned, and, in order to maintain more uni-
form drop currents, it was decided to keep the
magnet faces engaged at all times when the rods
were not in use.

Before the fuel concentrate (NQZUF6) was charged
to the enrichment system, a number of practice
operations were performed with carrier in the en-
richment system. The temperature control of the
system was inadequate and the operation of the
weigh cell on the transfer tank was such that
reproducible weights couid not be obtained. Al-
though the enrichment system probably could have
been made to work, considerable time would have
been required. Accordingly, the original enrich-
ment system was abandoned, and a manually oper-
ated system was improvised. The concentrate
was first batched down into a number of containers.

22

The concentrate in each of these containers was
then added to the fuel system at the pump after
first passing through an intermediate transfer pot.
The amount of concentrate added was determined
by weight measurements of the concentrate con-
tainers before and after use. While this system
was being set up, the operating crews were engaged
in leak testing the various auxiliary systems,
determining operability of heater circuits, checking
and reading thermocouples, establishing the per-
formance of the pump lubrication and cooling sys-
tems, checking the operation of the annulus system
helium blowers, monitoring the annulus helium for
leaks, and checking the operability of the various
off-gas systems, instrumentation, and the like.
By the morning of October 30, these tests were
sufficiently well in hand for the critical experi-
ment to commence.

¥

-

 

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4, CRITICAL EXPERIMENT

On October 30 the sodium had been circulating
275 hr, and the fuel carrier about 110 hr. Both
the sodium system and the fuel system were at an
isothermal .temperature of 1300°F, and the fuel
system was ready for the addition of concentrate
(N02UF6) to the carrier (NaZrFs). The addition of
the concentrate was necessarily time-consuming
because of the cautious manner in which each
addition had to be made during the approach to
criticality; however, this operation would have
been completed more rapidly had it not been for
unforeseen difficulties. The reactor did not reach
criticality until 3:45 PM, November 3, some four
days after the enrichment operation was started.
Unfortunately, much of the four days was spent in
removing plugs and in repairing leaks which oc-
curred in the enrichment line. Since these leaks
were largely the result of the improvised nature
of the enrichment mechanism, they were not of
serious consequence.

The chronology of the critical experiment is
shown in Fig. 4.1, and a detailed description of
the experiment, including a subcritical temperature
coefficient measurement, is presented below. The
critical experiment, as well as all nuclear oper-
ations, followed, in general, the pattern prescribed
in the “"ARE Operating Procedures, Part 1l, Nuciear
Operation,””! which is included in this report as

Appendix D.

ENRICHMENT PROCEDURE

By the afternoon of October 30 the chemists had
installed and checked out the injection apparatus,
and the systems were ready for the enrichment
operation to commence. The loading station was
located directly above the main fuel pump. The
injection rig consisted of an oven into which the
batch cans fitted, an intermediate, heated transfer
pot capable of holding about 5.5 Ib of concentrate,
a resistance-heated transfer line connecting the
batch can with the transfer pot, and the electrically
heated injection line running from the transfer pot
into the pump, as well as various auxiliary equip-
ment, such as helium supply and vent lines,
vacuum pump, pressure control, and electrical
heating equipment. The resistance-heated transfer
line was left exposed to the air so that blow

 

]J. L. Meem, ARE Operating Procedures, Part Il
Nuclear Operation, ORNL CF-54-7-144 (July 27, 1954).

torches could be applied in case of a plug. A flow
sketch of the injection system is shown in Fig.
4.2,

The fuel concentrate was batched into small
cans prior to loading. The three batch sizes are
given in the following tabulation:

Approximate

Can Size Concentrate Weight
(Ib)
A 30
10
0.5

The A and B size cans were used during the criti-
cal experiment. The C size cans were held for
use in calibrating the regulating rod. Table 4.1
lists the batch cans by numbers in the order in
which the cans were used in the critical experi-
ment, the net amount of concentrate and uranium
added, and the U233 content.

Prior to an injection a batch can was selected,
set in the oven, and heated to 1400°F. Meanwhile
the transfer lines were being heated. When proper
temperatures were reached, an injection was
attempted.  The intermediate transfer pot was
filled from the batch can by pressurizing the batch
can and venting the transfer pot. At the same time
the pump pressure was maintained above that of
the transfer pot so that there could not be an inad-
vertent transfer into the pump. When the spark
plug probes indicated that the transfer pot was
full, the pressure in the batch can was released,
and the material in the transfer pot was forced into
the pump by helium pressure. This method of
transfer was used so that only a measured amount
of concentrate (not more than 5.5 1b) could be
added at a given time and the system pressures
could be kept low. With low system pressures
there was less need for venting and possibly
clogging the vent lines. Figure 4.3 shows a photo
of the chemists preparing for an injection during
the critical experiment.

EARLY STAGES OF THE EXPERIMENT

At 1500 on October 30 the critical experiment
was started. The main fuel pump was first trimmed
to its minimum prime level and the shaft speed
was reduced to 500 rpm (20-gpm flow). At this

23

 
S s g - D ke .
it ek

 

 

 
el i s i 1 S i e i o s s . ..

TIME OF DAY

OCTOBER 30, 1954

   

 

0000 -
0300
0600
L CARRIER CIRCULATING;
PUMP SPEED, 500 rpm
osoc |
1200
1500 LEVEL OF CARRIER IN PUMP TRIMMED TO
MINIMUM OPERATING LEVEL
FIRST FUEL ADDITION: CAN NO. 6, 30.28 Ib
ADDED IN SIX BATCHES
FUEL PUMP SPEED INCREASED FROM
500 TO 1100 rpm
SAMPLE 6 REMOVED FOR ANALYSIS
1800
STARTED TO ADD FIRST 5-1b BATCH OF
CAN NO. 7
TRANSFER LINE PLUGGED; NO TRANSFER
2100
2400

i icaiis - i s i i eaciddl

CHRONOLOGY OF THE CRITICAL

OCTOBER 31, 1954

~ ATTEMPTED TO UNPLUG TRANSFER LINE

 

 

 

DECIDED TO INSTALL NEW TRANSFER
LINE; THEREFORE THE FIRST 5-ib
CAN OF CONCENTRATE FROM CAN
NO. 7 COULD NOT BE ADDED

L INSTALLING NEW TRAMNSFER LINE

INSTALLATION OF NEW TRAMNSFER
-7 LINE COMPLETED

SAMPLE 7 TAKEN FOR ANALYSIS

 

STARTED ADDITION OF SECOND 5-!b BATCH
OF CONCENTRATE FROM CAN NO. 7
{(FIRST BATCH OF REMAINING 25 ib)

W TRANSFER LINE PLUGGED AGAIN

 

T T TR < W T T oy e vy TR TR ' jiciaal e T

TR A T TR T R W T A T

 

| i e SRR i i .t il - i i il v 2 i il el e s R i s voime e o SRR -t L . i

NOVEMBER 1, 1954

UNPLUGGING INJECTION FITTING;
COMNNECTED VENT LINE TO SPARE

- INJECTION FITTING ON PUMP TO

MAINTAIN CONTINUOCUS GAS BLEED

THROUGH TRANSFER LINE

 

 

  
 
  
 
  
 

PLUG REMOVED FROM TRANSFER LINE

 

COMPLETED ADDITION OF FUEL FROM
CAN NO. 7 (25 |b ADDED)

}

SAMPLE TAKEN FOR ANALYSIS

THIRD FUEL ADDITION; 30.48 |b ADDED
FROM CAN NO. 8 IN 51X BATCHES

‘SAMPLE 9 REMOVED FOR ANALYSIS

FOURTH FUEL ADDITION; 30.27 ib ADDED
FROM CAN NO, ? IN SIX BATCHES

F ig. 4-]0

EXPERIMENT

s ol il el S .y w

NOVEMBER 2, 1954

FIFTH FUEL ADDITION; 30.11 Ib ADDED
FROM CAN NO. 10 IN SIX BATCHES

SIXTH FUEL ADD!TION; 29.17 ib ADDED
FROM CAN NO. 12 :

{

FIRST 5-1b BATCH OF SEVENTH FUEL
ADDITION TRANSFERRED FROM CAN
NO. 5; LEAK OCCURRED; 0.2 ib OF
FUEL LEFT IN LINE

SUBCRITICAL MEASUREMENT OF REACTOR
TEMPERATURE COEFFICIENT OF REACTIVITY
(EXP. £-2)

STARTED REPLACING SECTION OF TRANSFER
LINE CONTAINING LEAKY SWAGELOK
FITTING

} MEASUREMENT OF COUNT RATE V5.
SHIM ROD POSITION

SAMPLE 10 TAKEN FOR ANALYSIS

1

SAMPLE 11 TAKEN FOR ANALYSIS

 

ORNL-LR-DWG 6394

NOVEMBER 3, 1954

LEAK IN TRANSFER LINE REPAIRED, BUT
SPARE VENT LINE ON PUM? PLUGGED;
GAS BLEED MAINTAINED THROUGH
REGULAR PUMP VENT LINE

BALANCE OF SEVENTH FUEL ADDITION
INJECTED; TOTAL FUEL ADDED FROM
?IX BATCHES FROM CAN NO., 5 WAS

7.90 1b

   

} EIGHTH FUEL ADDITION; 9.%0 ib FROM
CAN NO. 11.IN TWO BATCHES

NINTH FUEL ADDITION; 14.11 |b FROM
CAN NO. 22 [N THREE BATCHES

FIRST RADIOACTIVITY OF REACTOR NOTICED;
RADIATION LEVEL 6 mr/hr

TENTH FUEL ADDITION; 13,76 |Ib FROM CAN
NO. 31 IN THREE BATCHES

ELEVENTH FUEL ADDITION; 11,0 Ib FROM
CAN NO. 20 IN TWO BATCHES

REACTOR CRITICAL (1545)

J

  

 

 

_\ CONTROL GIVEN TO SERVO
RADIATION SURVEY MADE; LEVEL AT REACTOR,
750 mr/hr; LEVEL AT FUEL PUMP 10 mr/he
REACTOR SHUT DOWN; CRITICAL
] EXPERIMENT COMPLETED

 

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1 —

 

TR TenT T

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TO VACUUM
PUMP
VENT
» [ ——
T-ELECTRICAL CONNECTIONS
5 FOR RESISTANCE HEATING

TO HELIUM SUPPLY ORNL-LR-DWG 6395

VENT
TO HELIUM SUPPLY

TRANSFER LINE

   
   

 

 

 

 

ELECTRIC HEATERS

 

 

|t

I aaan]

 

 

prd MW
TRANSFER POT | ]

MYVWM =———— GVEN

 

 

 

 

 

 

—

 

 

 

 

 

 

il ]
Tl |2
H_JE

 

 

 

 

iTO HELIUM SUPPLY

——><——— VENT

 

 

 

FLOOR LEVEL

 

  
  
  

 

 

 

 

 

AUXILIARY
VENT
SAMPLING LINE FUEL PUMP
DISCHARGE
SUCTION
Fig. 4.2. Equipment for Addition of Fue! Concentrate to Fuel System,
TABLE 4.1. FUEL CONCENTRATE BATCHES ADDED DURING CRITICAL EXPERIMENT
Batch Concentrate Added Uranium Added Weight of U235 Added
Date Time Can —_—
No. g Ib wt % g g b

10/30 1625 A-6 13,609 30.002 59.548 8104 7569 16.687
1N 1415 A-7 11,510 25,375 59.513 6850 6398 14,105
1804 A-8 13,852 30.538 59.530 8246 7702 16.980
2203 A-9 13,806 30.437 59.587 8227 7684 16.940
11/2 0213 A0 13,638 30.066 59.454 8108 7573 16,695
- 0610 A-12 13,241 29.191 59,531 7882 7362 16.230
0831 A-5 8,221 18.124 59.637 4903 4579 10.095
11/3 0523 A-11 4,505 9.932 59.671 2688 2511 5.536
0941 B-22 6,432 14,180 59.702 3840 3587 7.908
1245 B.31 6,312 13.915 59.637 3764 3516 7.751
1536 B-20 4,567 10.068 59.529 2719 2540 5.600

27

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R T P T T —T

e

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=

B d A st
o

 

Fig. 4.3. Chemists Preparing for Enrichment Operation During Critical Experiment,

TR ———— I B X o p-ray oy R e mr oo R R porre e R e P e v F?WW‘ ooy, e e T WE o g g RS R U FEE AP R S gm0 ET D gty T e P ol e e Trr e g g
il kBN

e i Bk

L e

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4wt g kb b

i, e PN

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g

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.

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time the volume of carrier in the fuel system was
calculated to be 4.82 %, and the weight was
927 ib. At 1507, concentrate transfer was started
from batch can A-6 to the intermediate tank, and,
at 1539, transfer of the first 5.5 Ib of fuel to the
system was accomplished. At 1554, the second
5.51b of fuel had been added. These fuel injections
very noticeably affected the fission chamber re-
corders. Mixing of the concentrate with the carrier
did not occur rapidly, and therefore each time the
enriched slug entered the reactor it produced a
multiplication that was observable on the fission
chamber count-rate recorder. Figure 4.4 is a photo-
graph of the trace of fission chamber No, 1; the
pips that occurred during addition of the first ond
second slugs of fuel are readily observable. From
the time interval between pips, which was almost

 

COUNTING RATE (arbitrary scate)

2The first five samples taken prior to this time were
for carrier impurity analysis. These five analyses were
discussed in chap. 3, **Prenuclear Operation,”’

   

TRACE OF FISSION CHAMBER NO. § RECORDER OCT. 30,1954 - — = *

exactly 2 min, and the volume of the fuel system,
a check on the rate of fuel flow was obtained:

4.82 (%) x 7.48 (gal/#)

f {gpm) = 3 {min)

 

= 18.03 .

This checked fairly well with the value of about
20 gpm read from the fuel flow recorder.

The remaining four injections from can A-6 were
accomplished smoothly. The pumps were speeded
up to an observed fuel flow rate of 46 gpm to
obtain better mixing of the fuel. At 1720, fuel
sample 6 was removed for analysis.?

At 1920 the first 5.5-lb injection from can A-7
was started, but the transfer line from the inter-
mediate pot to the pump clogged due to concentrate
freezing in it. After several hours of unsuccessful
attempts to free the line, a gas leak occurred in
the transter line at the intermediate tank pot, and

ORNL~LR-DWG 3853

PUMP AT MINIMUM SPEED

SECOND SLUG THROUGH SECOND TIME
SECOND SLUG THROUGH FIRST TIME
! i

e

U'FIRST SLUG THROUGH SECOND TIME

&
,5
oL
ul
o
z
g
]
w
45
Lz
u-

 

—e————— TIME

Fig. 4.4. Passage of Enriched Slugs Through Reactor,

29

 

¢
r

S T

T e

 
 

 

 

it was then decided to install a new transfer line
and transfer pot. At this time it was reported that
no transfer of concentrate had been made to the
pump from can A-7. The first 5.5 Ib of can A-7
was lost to the experiment through the leaks and
in the plugged line.

At 2000 on November 1, the new transfer line and
tank were installed and checked out ready for use.
At 2006, fuel sample 7 was drawn for analysis,
and, at 2300, the critical experiment was resumed.
When the first 5.5-Ib slug was injected, periodic
pips were again observed on the fission chamber
recorder. These pips had a period of 47 sec which,
together with a pump speed of 1080 rpm, yielded
a calculated fuel flow of 46 gpm, which corre-
sponded to the flow observed on the fuel flowmeter.

At 2315, while attempting the next 5.5-lb transfer,
the line again froze. The line was disconnected
and found to be plugged at the injection fitting
through the pump flange. |t was therefore decided
to increase the current to the resistance-heated
fitting and to add a separate vent line from the
pump so that the chemists could continue to *‘blow
through’’ the transfer line and out the new vent
line without raising the system pressure.

By 1320 on November 1 the new vent line had
been instalied and the transfer system was again
in operating condition. The transfer of the re-
maining 20 |b of concentrate in four batches from
can A-7 took place with no difficuity, and the
transfer was completed by 1415. At 1617, fuel
sample 8 was taken for analysis, and, at 1707,
the first batch from can A-8 was transferred into
the system. The remaining 5 batches were trans-
ferred smoothly, and the transfer from can A-8 was
completed by 1804. At 2028, fuel sample 9 was
removed for analysis. The next three additions of
fuel from cans A-9, A-10, and A-12 were accom-
plished with little difficulty.

At 0830 on November 2, while transferring the
first 5-1b batch from can A-5, a leck occurred in
the injection line just below the floor level of the
loading station {see Fig. 4.3). Examination of the
injection line showed the leak to have occurred at
a Swagelok fitting. Approximately 0.2 Ib of concen-
trate was lost from the experiment as a result of
the leak. Since the line had clogged, it was neces-
sary to install a new section of line. During the
time the repairs were being made to the injection
system, the original schedule was changed, and a
subcritical measurement of the temperature coef-
ficient of reactivity was made. In addition, samples
10 and 11 were taken for analysis.

30

SUBCRITICAL MEASUREMENT OF THE
REACTOR TEMPERATURE COEFFICIENT

Approximately 100 Ib of U?3° (180 Ib of Na,UF )
had been added at the time the leak occurred on
November 2, at 0831, and it was estimated that
this was about 80% of the total fuel needed. Al-
though it had been planned to make a preliminary
measurement of the temperature coefficient of
reactivity with about 90% of the fuel added (see
“*“Nuclear Operating Procedures,’”’ Appendix D),
the plans were revised and the measurement was
made at this time.

At the start of the experiment the reactor mean
temperature was 1306°F. The fuel heat exchanger
barrier doors were raised (at 1003); two minutes
later the fuel helium blower was started and its
speed increased to 275 rpm. The resultant cooling,
with time, as traced by the reactor fuel mean
temperature recorder, is shown in Fig. 4.5, with
the counting rate from fission chamber No. 2 super-
imposed. The data for both fission chambers are
given in Table 4.2. The counting rate of the
fission chambers monitors the neutron flux.

The counting rate was observed to rise with
decreasing temperature and thus indicated a nega-
tive temperature coefficient of reactivity, Counts
were taken for 40 sec during every minute. At
1010, when the reactor temperature reached 1250°F,
the helium blower was stopped, and soon thereafter
the reactor temperature began to rise again, with
a corresponding decrease in counting rate.

ORNL-LR-DWG 6396

 

 

 

 

 

 

 

 

 

 

 

 

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1006 ——
1004
1002

® e ¥ w9
e 2 & g 2
TiME OF DAY

Fig. 4.5. Subcritical Measurement of the Reactor
Temperature Coefficient. (Plot of fission chamber
No. 2 counting rate superimposed on fuel mean
temperature chart.)

 

Lo e pre T e e

et gy F P e v ¥ Creem o B

- vt

R

e prmr Y o

T T

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o W T

L e 1

EEreRTY

i

 
e K . -

Tt b o

b B LA

B o,

e Rk we molbgle s

i,

il s

TABLE 4.2, SUBCRITICAL MEASUREMENT OF THE
REACTOR TEMPERATURE COEFFICIENT

 

 

Fission  Fission

* No. 1 No. 2

) (counts/sec) (counts/sec)
1004 1306 486.4 723
1005 1304 492.8 736
1006 1302 512.0 755
1007 1290 531.2 780.8
1008 1275 537.6 793.6
1009 1260 537.6 793.6
1010 1252 544,0 800.0
1011 1244 531.2 787.2
1012 1240 531.2 774.4
1013 1246 518.4 768.0
1014 1252 518.4 768.0
1015 1258 512 761.6
1154 1275 502 736.0
1319 1284 490.9 725.0
1422 1290 484 713.0

 

The reactor temperature coefficient, as estimated
from the data presented in Fig. 4.5, was of the
order of ~5 x 10~3 (Ak/k)/°F. The data were
adequate to show that the reactor would be easy
to control; therefcre the experiment was allowed
to proceed. |t will subsequently be observed that
the temperature data recorded in the data room did
not exhibit the full temperature drop across the

reactor in any run where heat was being abstracted

from the system. Consequently where the ab-
stracted heat is one of the parameters of the
experiment, as in the above temperature coefficient
measurement, it would be expected that some
temperature correction would be in order. The
existing data for the case in question are, how-
ever, too meager to justify correction, although
based on correlations (app. K) with data taken
during the low- and the high-power runs, the con-
trol room temperature indications were lower than
the actual temperature, and therefore the estimated
temperature coefficient given above is too high.

It may aiso be noted from examination of Fig.
4.5 that the maximum counting rate and the temper-

ature minimum did not occur simultaneously. A-

lag of about 2.5 min in the response of the reactor
temperature was observed. The reason for this
time lag is not well understood, but it may have
some connection with the fact that the thermo-
couples were outside the reactor and hence did
not ‘‘see’’ the changes immediately. Details of
this and later measurements of the fuel and reactor
temperature coefficients are given in Appendix O.

APPROACH TO CRITICALITY

At 0130 on November 3, the injection system was

again in operation, and the fuel additions during .

the remainder of the critical experiment occurred
without mishap. The remaining portions of can
A-7 and all of cans B-22, B-31, and B-20 were
added during this time.

Throughout the course of the experiment, the
progress toward criticality was observed on the
neutron detectors. With every fuel injection the
counting rates of the two fission chambers and the
BF, counter were simultaneously clocked and
recorded. The approach to criticality could readily
be seen by plotting the reactivity, &, against the
concentration of the fuel in the system. The re-
activity was obtained from the relationship

1
k=1-—,
M
where M is the subcritical multiplication, which
is determined from the expression

where N is the counting rate after a fuel injection
and Ny is the initial counting rate. As criticality
is approached, 1/M approaches zero, and therefore
at criticality, & = 1.

Figure 4.6 shows a plot of & vs U235 concen-
tration, where & is determined from three neutron
detectors, i.e., two fission chambers and o BF3
counter. The reactivity, as observed on the two

fission chambers, showed a rapid increase during

the early stages of the experiment and then leveled
off as the critical condition was near. The BF,
counter, on the other hand, showed a more uniform
approach to criticality. The differences in the
responses of the two types of detectors are dis-
cussed in Appendix E. Table 4.3 presents the
data from which Fig. 4.6 was drawn.

A condensed, running uranium inventory during
the critical experiment is given in Table 4.4, The

31

 

 

 
 

 

 

ORNL-LR-DWG 3852B

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

-~ | |
FISSION CHAMBER NO. 2 -~ EXP. E-f, NOV. 3
, | b,
0.9 //A ,
FISSION CHAMBER NO. { i //
V8
1.0
0.8 7’
/
' // 0.98
0.7 /. e e e e
7
r;
/f
~
> / *
~ / o 0.96
—
; 0.6 - £
g :
gg )
—-BF5 COUNTER @
: 0.94 e
05
0.92 _
0.4 /
A
16 i8 20 22
U235 CONCENTRATION IN FUEL (In/f3)
03 ——- S
0.2
0 4 8 12 16 20 24 28 32 36 40

U235 CONCENTRATION IN FUEL (ib/ft3)

Fig. 4.6. Approach to Criticality: Reactivity vs Fuel Concentration.

figures of columns 4, 6, and 11 were supplied by
the chemists.®

At 1545 on November 3 upon the addition of the
second batch of fuel from can B-20, a sustaining
chain reaction was attained - the reactor was
critical. Figure 4.7 shows a photo of the control
room just as the critical condition was reached.

At 1547 the reactor was given over to the servo
mechanism at an estimated power of 1 watt, and
during the next ¥ hr a brief radiation survey of
the reactor and heat exchanger pits was made (cf.,
chap. 5, “Low-Power Experiments’”). At 1604 the
reactor was shut down, and thus the initial phase
of the ARE operation program was completed.

 

3). P. Blekely, Uranium in the ARE, ORNL CF-55-
1-43 (Jan. 7, 1955).

32

The uranium inventory (Table 4.4) showed that
the critical concentration of U232 was 23.94 b/,
which corresponded to a total weight of 133.8 Ib
of U233 in the fuel system. At this time the total
weight of the fuel was 1156 Ib, and its volume was
5.587 #°. The per cent by weight of U?3% was
calculated to be 11.57.

The critical mass of in the reactor was
calculated from the ratio of volumes:

U235

 

Vv
7
M =M .
r 5 V
5
where

M, = critical mass of U235 in reactor core,
M, = mass of U233 in system,
V = critical volume of reactor,

NPT VR & WEITTT T e b T TERTEON T e, TR M e

o ey

* o

B ETr NS e wTRCMmcR MR g Meow SRR o

FETY T ry T e, (TSN S 7 TR e

A W
i, .

CebiRe Ra

R

~ i

ke i

Ca el e sgieha.

Yy

L

[

-

TABLE 4.3.

APPROACH TO CRITICALITY:

REACTIVITY FROM VARIQUS
NEUTRON DETECTORS vs FUEL CONCENTRATION

 

Fission Chamber No. 1

Fission Chambet No. 2

BF, Counter

 

 

I&:i Concli::rf;fion Counting Multiplication, R . Counting Multiplication, R . Counting Multiplication, R o
No. (b/f%) Rate, N M ecc:vlty, Rate, N M eacway, Rate, N M eqclzlvufy,
{counts/sec} (N/Ng) {counts/sec} (N/Ng) (counts/sec) {N/NG)
1 0 9.46* 1 0 16.28* i 0 4.49* 1 0
2 3.39 26.7 2.82 0.645 42,5 2,61 0.617 7.04 1.57 0.381
3 6.16 47.44 5.02 0,801 71.54 4.39 0.772 9.19 2.05 0.510
4 9.37 86,96 9.19 0.891 128.5 7.90 0.873 13.43 2.99 0.667
5 12.45 147.2 15.6 0.936 217.2 13.34 0.925 19.73 4,39 0.780
6 15.36 251.9 26.6 0.962 367.6 22,6 0.956 29.5 6.57 0.848
7 18.09 450.9 47.6 0.97% 649.9 39.9 0.975 49,0 10.3 0.908
8 19.74 648.2 68.5 0.985 977.4 60,0 0.983 70.5 15.7 0.936
9 20.63 859.3 90.8 0.989 1317 80.9 0.988 93.7 20.9 0.952
10 21.88 1489 157.4 0.994 2316 142.3 0.993 161.3 35.9 0.972
n 23.08 4028 425.6 0.998 6730 413.5 0.998 442.0 98.5 0.990
12 23,94 Critical

 

*Initicl counting rate, N,

TABLE 4.4. URANIUM INVENTORY DURING CRITICAL EXPERIMENT

 

 

 

 

 

Samples Removed Fuel Mixture Uranium-235
Run Fuel Concentrate Added for Analysis Total Weight ol . Total
Date Time No. Weight Yolume Fuel y23s Concentrate Tm Density v’i:'jh; Weight Cencentration
(Ib) (£t3) Removed Removed Plus Carrier o"éme (tb/Ft3) 8¢ in System |b/f% Wt %
(1b) (1) (Ib) (#) () (i)
10/30 1425 1 927.3 4.820 1924 0
1625 2 30.00 0.1046 957.3 4,925 194.4 16.687 16.687 3.388  1.743
1740 2.447 0.0427 954.8 4,912 16.644
10/31 2013 253 0.0441 952.3 4,900 16.600
1171 1415 3 25.37 0.0885 $77.7 4.988 196.0 14,105 30.705 6.156 3.141
1635 2.286 0.0716 977.5 4.976 30.633
1804 4 30.54 0.1065 1008 5,082 198.3 14,980 47.613 9.368 4.723
2040 2.670 0.1261 1005 5.069 47.487
2203 5 30.44 0.1062 1036 5175 200.2 16.940 64.427 12,45 6.220
12 0213 6 30.07 0.1049 1066 5.280 2019 16.695 81.122 1536  7.610
0610 7 29.19 0.1018 1095 5.382 203.5 16,230 97.352  18.09 8.890
0831 8a 5.50 0.0192 1101 5.401 203.8  3.058 100.410 1859  9.123
1625 2.562  0.2337 1098 5.388 100.176
2025 2.716 02478 1095 5.375 99.928
/3 0256 85 1262 0.0440 ’ 1108 5.419 2045 7.037 106965 19.74  9.654
0523 9 9.93 0.0346 1118 5.454 205.0 5,536 112,501 20,63 10.06
0941 10 14,18 0,0495 1132 5503  205.7 7.908 120,409 21.88 10.64
1245 W 13.92 0.0486 1146 5.552 2064 7751 128,160 23.08 11.18
1536 12 10.07 0.0351 1156 5.587 206.9 5.600 133.760 23,94 1157
1545 Criticality reached 7 7
/4 0910 2.679  0.3100 1153 5.574 133.450
0915 2718 0.3145 15 5.561 133,135
1310 2.482  0.2872 1148 5.549 132,848
1315 0.732 0.0847 1147 5.545 132,763
33

 

 

TR T

 

 
 

 

B g

< e

 

ik b

 

 

34

R

e

e

R

 

icality,

Control Room at Cri

ig. 4.7.

F

 

mr vvos

" g

s by - T W ATIARENTERT R B Mo el v @ T FET

L

t

 

o PR SRR BT e FT e peert nnpe e e

o m————

 

 

™y

"o

 

y T TR R ee FEER

 
V_ = system volume.
Therefore
1.37 #2
M, = 133.81b [—— | =3281ib .
5.587 f+°

The calculated, cold, clean critical mass of the
reactor, as obtained from the subsequent rod cali-
brations, was 32.75 Ib (cf., app. F).

The reactor was not instrumented to permit the
measurement of the flux or power distributions
through the reactor, but measurements of these
distributions were made on a critical mockup of
the reactor at the ORNL Critical Experiment Fa-
cility about two years before the operation of the
ARE.? These measurements represent the best

 

4p. Callihan and D. Scott, Preliminary Critical As-
sembly for the Aircraft Reactor Experiment, ORNL.-1634
(Oct. 28, 1953).

information that is available and, because of the
general interest therein, typical axial and radial
flux and power distribution curves are given in
Appendix G.

ANALYSES OF FUEL SAMPLES

in addition to the fuel samples taken, as noted,
during the critical experiment, four more samples
were removed on November 4 after initial criticality
was reached. A list of all samples taken and the
results of the chemical analyses are presented in
Table 4.5. Besides showing the analyses of the
percentages of U and U?3% by weight in the fuel
system and a comparison with the U233 (wt %)
content obtained from the criticality data, the
table also lists the results of the analyses for the
impurities and corrosion products Fe, Cr, and Ni
to provide information on the purity of the fuel and
the corrosion rate of the fuel system.

TABLE 4.5. CHEMICAL ANALYSES OF FUEL SAMPLES

 

 

 

U235 ¢ 235
vure Tim, Sample  Impurities and Corrosion Products (ppm) UT::L'M Chemical _from Critical
No. Cr Fe Ni (wt %) Analysis Experiment Data
(wt %) (wt %)
10/25 1650 1 90 200 50
10/26 1414 2 81 <5 40
1911 3 81 <5 6
10/27 1919 4 90 18 12
10/29 0935 5 102 30 20
10/30 1740 6 100 20 10 2.47 + 0.01 2.31 1.74
2013 7 150 15 10 1.84 * 0.04 1.72 1.74
11/1 1635 8 190 15 15 3.45 * 0.01 3.22 3.14
2040 9 200 10 17 5.43 * 0,01 5.07 4.72
/2 1625 10 210 9.58 + 0.08 8.95 9.12
2025 n 205 20 20 9.54 + 0,08 8.91 9.12
1174 0910 12 310 | 12,11 20,10 11,32 11.57
0915 13 300 12,21 £ 0,12 11.41 11,57
1310 14 320 12.27 £ 0,08 11.46 11.57
1315 15 310 | 12,24 £ 0,12 11.43 11.57
1/5 1100 16 372 5 <5 1254 £ 0.07 1172 11.72
11/6 17 f‘ 1257 2007 1175 11.79
3 0535 18 420 12.59 £ 0,12 11.77 11.79
11/7 0423 19 445 13.59 + 0.08  12.70 12.38

 

35

 

 

 
 

3
13
. it
.
A
4

:i

 

 

The first five samples were removed from the
system prior to enrichment, and samples 16 through
19 were taken after the critical experiment and
during the low power runs at the time of the cali-
bration of the regulating rods against fuel addition.

For the most part, the U233
chemical analysis agreed to within a few per cent
with that obtained from the running inventory.
When the sample for analysis was withdrawn from
the system too soon after a fuel addition, there
was a tendency for the analysis to be low, un-
doubtedly, because of inadequate mixing of the
additive with the bulk of the fuel. The chemical
analysis of sample 1, however, was obviously in
error, whereas that of sample 9 gave a uranium
content that was about 7% higher than the inventory
showed. The other sample analysis figures were
within about 2% of the percentages given by the
inventory. The analyses of samples 12, 13, and
14, which were taken after the critical experiment,
agreed to within about 1% with the inventory calcu-
fation. '

The increase in the buildup of chromium in the
fuel system with time is shown in Fig. 4.8 to give
an indication of the corrosion rate of the system.
The initial corrosion rate was quite small, but
after enriched fuel had been added to the system,
the analyses showed a chromium content increase

of about 50 ppm/day.

content as given by

CALIBRATION OF THE SHIM RODS

As outlined in the ‘‘Nuclear Operating Pro-
cedures,”” Appendix D, during the first fuel ad-
ditions the shim rods were withdrawn all the way

out of the reactor to obtain the multiplication.

After about 100 Ib of U%%® had been added to the
system, the procedure was altered in order to
obtain a shim rod calibration. After each fuel
addition, all three rods were simultaneously with-
drawn to positions of 20, 25, 30, and 35 in. out;
total movement was 36 in. The counting rate of
each of the neutron detectors was recorded for
each rod position. Figure 4.9 shows the reciprocal
multiplication and the reactivity as a function of
U?35 content of the fuel system for the various
rod positions. The data ysed for plotting Fig. 4.9
were obtained from the BF3 counter and are pre-
sented in Table 4.6. A cross-plot of reactivity vs
shim rod position determined the rod calibration
detail s, which are given in Appendix J. The rods

36

ORNL-LR-DWG 63387

500 .
= o .
— o 12.4 %
£ = 2
s 400 O .8 z
O O
g Hr 2
[ < H
= 4
w w! e £
E 300 2 =
o
5 2 2
2 200 [ > 51 ©
= Eo 3y @
o g dom
& < 17 °4
X o g
© 00 *D_: f
QO

 

0

25 28 34| 4 4 8
OCT, = NOV.

Fig. 4.8. Increase in Chromium Concentration in
Fuel as a Function of Time.

were found to be nonlinear, with (Ak/k)/in. and
total Ak/k varying in the manner shown in Appen-
dix D. From the calibration it was found that each
rod was worth a total of about 5.8% Ak/k.

As a check on the general shape of the reactivity
curves of the shim rods, a series of counts were
taken on the BF3 counter and the fission chambers
for various rod positions for two different uvranium
concentrations. These data are also discussed in
Appendix J.

MEASUREMENT OF THE
REACTIVITY-MASS RATIO

Prior to the ARE operation it had been estimated
that when the critical mass (assumed to be 30 Ib)
was reached, the value of the ratio (Ak/R)/(AM/M)
should be 0.232. This valve was obtained from a
calculation of the ratio for various amounts of
U233, as shown in Fig. D.1 of Appendix D. From
the data taken during the critical experiment, it
was possible to establish an experimental curve
for the ratio and to verify the value given above
for the ratio at the critical mass.

In order to find (Ak/k)/(AM/M) experimentally,
use was made of the curve of Fig. 4.9 which gives
E in terms of the U233 content of the fuel in the
system for various shim rod positions; i.e., for any

Ak/k (/_\.k M <
AM/M  \NAM/ \&/ '

which is the reactivity-mass ratio. The value

point,

 

v

poer

mops s g

 
TABLE 4.6. REACTIVITY vs U235 CONTENT OF FUEL SYSTEM EOR VARIOUS SHIM ROD POSITIONS

 

 

E xperiment u23s5 Shim Rod Counting Rate,
E-1 in System Positions™ BF3 Counter NO/N k=1-— (NO/N)
Run No. (1b) (in. out) {counts/sec)

8 A 100.41 20 32.0 0.140 0,860

25 38.4 0.117 0.883

30 46.9 0.0956 0.904

35 54,3 0.082 0.918

8B 103.13 20 34.1 0.131 0.869

25 40.5 0.111 0.889

30 51.2 0.0877 0.912

35 57.8 0.0774 0.923

8 C 106.18 20 36.3 0.123 0.877

25 46.9 0.0954 0,905

30 57.8 0.0774 0.923

35 66,1 0,0676 0.932

8 D 106.97 20 36.3 0.123 0.877

25 46,9 0.0954 0.905

30 59.7 0.0749 0.925

35 68.3 0.0655 0,935

9 A 110,14 20 38.4 0.117 0.883

25 51,2 0.0874 0.913

30 68.3 0.0655 0,934

35 83.2 0.0538 0.946

9B 112,50 20 40.5 0,111 0.889

25 55.5 0.0808 0.919

30 76,8 0,0582 0.942

35 93.9 0.0479 0.952

10 A 115,58 20 44.8 0.0998 0.900

25 59.7 0.0749 0.925

30 83.2 0.0538 0.946

35 110.9 0.0403 0.960

10 B 118.63 20 42.7 0.105 0.895

25 68,3 0.0655 0.934

30 93,3 0.0479 0.952

35 136.5 0,0327 0.967

0C 120.41 20 51,2 0.0873 0.913

‘ 25 68.3 0.0655 0.934

30 110.9 0,0403 0,960

35 162,1 0.0276 0,972

1A 123.43 20 51.2 0.0873 0.913

25 85,3 0.0524 0.948

30 128.0 0,0349 0,965

35 - 213.3 0.0209 0.979

*Position of all three shim rods.

37

Y i

 

 

:
|

 
 

 

TABLE 4.6 (continued)

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Experiment u23s Shim Rod Counting Rate,
E-1 in System Positions* BFS Counter Ng/N k=1-—- (NO/N)
Run No. (Ib) (in. out) (counts/sec)
11 B 126.48 20 59.7 0.0749 0.925
25 93.9 0.0476 0.952
30 162.1 0.0275 0.973
35 307.2 0.0145 0.985
11cC 128,16 20 59.7 0.0749 0.925
25 93.9 0.0476 0.952
30 196.3 0.0228 0.977
35 443.7 0.0101 0.990
12 A 131.08 20 59.8 0.0748 10,925
25 110.9 0.0403 0.960
30 256.0 0.0175 0.983
35 955.7 0.00468 0,995
12B 133,76 20 68.3 0.0655 0,935
25 136.5 0.0327 0.967
30 375.5 0.0119 0.988
35 Critical
*Position of all three shim rods.
RUN NUMBER ORNL—LR—-DWG 6398
8A 88 8C SD 9B {0A 10B 10C 11A B 14C 12A 12B
— J e
0.6 ' l | | | ! 0 | l’l ioolS l\‘\’ ?lss'm-__,____-L 1
N EXPERIMENT E-1, NOV 3, 1954 - s\—\“\gR :I;rl .
~ BF, COUNTER DATA —T
= 014 //..“/ l . 100 0.g8
\O -8 e ,’-f'.’. ODS AT
= |_le=—Toum R l L.
< o012 .t l - 096
- onS AT F’ ‘[“‘
= e R x
S 040 ] 9"‘]‘ML | ow. |- 094 -
o ol o) |SHME ] S
- i -
Qo8 sl 092
é ® //f/ OSH - — 5
= f M ROpg AT 20‘;"‘--.3~_. @
g 006 Tl SHIM| e 0.90
5 004 ol | ‘? "“r‘ n__~___ 088
3 ol LTl s g
— | ‘ T - ‘(:)“ES AT 30in '
002 SOLID LINES SHOW REACTIVITY - Ol | SHiMg F?oo?,fi’"*i% 086
e — T 35 . I
DASHED LINES SHOW RECIPROCAL MULTIPLICATION n.
1 1 L] i ? L1 ll_l_ ] - 1 Ull_ ! JL 1 llo ] ’ TNI\T'?"'TE-- 0.84

 

 

 

 

 

 

 

 

 

 

 

 

 

 

0
100 102 104 106 108 110 12 144 416 418 120 422 124 126 128 130 432 134 136
>IN FUEL SYSTEM {ib)

Fig. 4.9. Reactivity and Reciprocal Multiplication vs U235 Content of Fuel System for Yarious Shim
Rod Positions,

38

»

! R T B ¢

i T T R T E e W T IR O R

e

s

TE R PR R CURE T P T

¥

v

B g R L A L L A
ORNL~LR~DWG 6399 (AR/AM) is the slope of the & vs M curves shown
in Fig. 4.9. For various small sections of the
curves the slopes approximated straight lines.
From the average values of M and & over these
increments of the curves and the measured slopes,

0.35

 

 

g 0.30 the ratio R was determined for each increment.
. 3 Figure 4.10 shows a plot of R vs the U23% content

> of the fuel in the system.

3

<

0.25 The value of R for the experimental critical mass

(133.8 ib) was calculated to be

Ak/k
AM/M
for shim rod positions of 25, 30, and 35 in. Since

e e e el s s, e e

= 0.236

 

 

0.20
25 26 27 28 29 30 3 32 33 34
235

 

MASS U IN REACTOR (10) this value (0.236) was in excellent agreement with

: the calculated value (0.232), it was used through-

Fig. 4.10. Reactivity-Mass Ratic as a Function out the experiment. The data are tabulated in
of the U235 Content of the Fuel in the Reactor, Table 4.7.

TABLE 4.7. REACTIVITY-MASS RATIO AS A FUNCTION OF THE AMOUNT OF U235 IN
THE REACTOR FOR VARIOUS SHIM ROD POSITIONS

 

Reactivity-Mass

 

 

 

 

Mass Shim Rod M, Mass of .
' . Ratio,
® Range* Position** U235 i Reactor An Reactivity, Ak AL/k
(1b) (in.) (1b) (1b) k
AM/M -
100 10 106 25 25,24 1.47 0,891 0.0172 0.331
30 25.24 1.47 0.913 0.0175 0.329
35 25,24 1.47 0.924 0.0174 0.322 ~_
| Average 0,327 i
114 +o 120 25 ) | 28,66 1.47 0,929 0.0137 0.287 E
30 28,66 - 1.47 0.951 0.0136 - 0.279
35 28.66 1.47 0.963 0.0135 0,273
Average 0,280 ]
128 t0 134 25 32,10 44 0,961 6.0110 0,257
' : 3O 32,10 | 1.49 0,982 0.0100 0.219
35 : - 32,10  1.40 0.995 0.0100 0,231
| 7 . Average 0,236
*These rfiass rarfgeg refer to the curves of Flg 4.9 on .\n'.?hich these calculations are based. |
. **Position of all three rods. | E

P T

39

 

 
 

 

 

 

 

5. LOW-POWER EXPERIMENTS

The reactor was operated, after initial criticality
was attained, for about 20 min at a nominal power
of 1 w and then shut down. The next morning a
series of low-power experiments was started. This
series of experiments lasted from the morning of
November 4 until the morning of November 8. The
chronology of this low-power operation is given in
Fig. 5.1.

Two of the early experiments, L-1 and L.-4, were
devoted to a determination of the reactor power
from measurements of the fuel activation. These
runs were each of 1 hr duration with the reactor at
a nominal power of 1 and 10 w, respectively. After
each of these runs, fuel samples were taken and the
activity counts were made from which the reactor
power was determined., The method was inaccurate,
as subsequently evidenced by power calibrations
from the extracted power, but did indicate that the
nominal power estimates were low,

Radiation surveys of the entire experimental
system were made during both the nominal 1- and
[0-w runs except that a survey of the reactor pit
was not attainable during the 10-w run because
the pit was sealed at that time. Most of the equip-
ment and many of the instruments were, however,
located in the heat exchanger pits to which access
was maintained throughout the low-power operation.

Most of the time of the low-power operation was
devoted to calibration of the regulating and shim
rods. The two essentially independent methods
used to calibrate the regulating rods were cali-
bration against fuel addition and calibration against
reactor periods by using the inhour equation (app. I}).
The shim rods were then calibrated against the
regulating rod.

Also as an integral part of the low-power opera-
tion the earlier measurement of the temperature
coefficient was checked, and the fuel system per-
formance characteristics were determined. Finally,
in preparation for the high-power experiments, the
ion chambers, which had been positioned close to
the reactor during the critical and low-power ex-
periments for greater sensitivity, were withdrawn
so that they would register sensibly at the higher
powers,

POWER DETERMINATION FROM
FUEL ACTIVATION

Reactor power measurements are usually made
by exposing gold or indium foils to the neutron

40

flux in a reactor soon after initial criticality is
attained. In the ARE it was simpler to draw off a
sample of the uranium-bearing fuel after operation
and measure its activity with an ion chamber. Com-
parison with a similar sample of known activity
gave a determination of the flux level and, hence,
the power level. The procedure is described in

Appendix H.

In run L-1, the reactor was operated for 1 hr at
an estimated power level of 1 w, and a sample was
drawn off for measurement of the activity. The
specific activity was too low for a reliable de-
termination and, accordingly, the experiment was
repeated in run L-4 at a nominal 10-w power level,

It was subsequently found from the heat balance
at high power that the actual power was 27w during

run L-4, However, the fuel sample activity was:

only on the order of one-half to two-thirds the
activity to be expected from operation at 27 w.
Apparently a considerable amount of the gasecus
fission products was being given off from the fuel.
A comparison of the data obtcined from fuel acti-
vation with those obtained from heat balances is
given in Appendix N,

RADIATION SURVEYS

The primary purpose of experiments L-1 and L-4
was to calibrate the estimated reactor power against
the power determined from the radioactivity of fuel
samples, but, at the same time, radiation dose
levels were measured at various locations around
the reactor and the tank and heat exchanger pits.
These data will subsequently be of interest in
evaluating the radiation damage to various com-
ponents of the system.

Most of the radiation measurements were made
with a “‘Cutie-Pie’’ (a gamma-sensitive ionization
chamber), but a GM survey meter, a methane pro-
portional counter, an electroscope, and a boron-
coated electroscope were used for some measure-
ments, in particular, those made close to the
reactor. While the GM survey meter only provided
a check on the gamma dose as measured by the
‘‘Cutie-Pie,” the other instruments provided meas-
urements of the fast- and thermal-neutron doses.
The proportional counter measured the fast-neutron
dose, and the thermal-neutron flux was calculated
from the difference of the two electroscope readings.

 

I PR T T

 

 

 

 
 

TIME OF DAY

T

0000

0300

0600

Q900

1200

1500

1800

2100

2400

M .o, Raiaieiii u., it

e e

NOVEMBER 4, 1954

L]
I

 

 

 

 

 

SAMPLES 12 AND 13 REMOVED FOR ANALYSIS

REACTOR MADE CRITICAL AND POWER
INCREASED TO 1t w

REACTOR AT~! w FOR DETERMINATION
OF POWER LEVEL

SAMPLES 14 AND 15 REMOVED FOR ANALYSIS

START OF CALIBRATION OF REGULATING ROD
FROM FUEL ADDITION

RUN O
PENGUIN NO. 11 INJECTED}

RUN 1}
REACTOR AT 1 w

EXP.

PENGUIN NO. 19 INJECTED

RUN 2
REACTOR AT 1 w

REGULATING ROD FOUND TO 8E TOO
LIGHT; CHANGED TC ROD WITH MORE kex

FUEL SYSTEM CHARACTERISTICS RUN; EXP, L-3

} EXP. L-1

L-2

CHRONOLOGY OF THE LOW-POWER EXPERIMENTS

NOVEMBER 5, 1954

   
  
   
  
 
   
    
 
 

 

 

 

 

 

o~
:_/ REACTOR AT 1 w o
PENG UIN NO. 10 INJECTED; GAS RUN 11| -
LEAK IN INJECTION SYSTEM >
- DEVELOPED 5
REACTOR AT | w
PENGUIN NO. 4 INJECTED GAS LEAK REPAIRED
REACTOR AT 1 w RUN 34
REACTOR AT 1 w
PENGUIN FOUND TO HAVE NO
DIP LINE; NO FUEL INJECTED POWER INCREASED
REACTOR AT 1 w REACTOR AT 10 w; 1-hr RUN FOR POWER
PENGUIN NO. 14 INJECTED DETERMINATION; RADIATION SURVEY
© (NJECTED + RUN 38 o MADE; EXP. L-4
REACTOR AT 1 w o
REACTOR AT 1 w L~ SAMPLE 17 REMOVED FOR AMALYSIS
o
PENGUIN NO. 16 INJECTED b RUN 4 % SAMPLE 18 REMOVED FOR ANALYSIS
—\ REACTOR AT 1 w — -
REACTOR AT | w FOR RUN 1 OF EXP. L-5:
1
REACTOR AT ‘”] ROD CALIBRATION AGAINST PILE PERIOD
PENGUIN NO. 13 INJECTED r RUN 5 PERIGD EXCURSION TO ~100 w; RUN 2
REACTOR AT 1 w
REACTOR AT T w F—1—  INJECTION LINE PLUGGED APPARENTLY “
PENGUIN NO., 5 INJECTED » RUN 6 AT FITTING INTO PUMP o
REACTOR AT 1 w p ] REACTOR AT 1 w -
>
DURING RUN 6, HIGH GAMMA RADIATION -J FUEL PUMP STOPPED &
NOTICED AT PUMP; REACTOR MADE REACTOR EXCURSION TO ~50 w RUN 3
SUBCRITICAL DURING INVESTIGATION PUMP STARTED
[} PUMP LEVEL TRIMMED BY REMOVING
= ABOUT 71 1b OF FUEL MIXTURE N#gkiAIRNoPPIETe;AHON STOPPED FOR FINAL
[ S
SAMPLE 16 REMOVED FOR ANALYSIS INJECTION FITTING AT PUMP FOUND TO BE
_/ PLUGGED BECAUSE OF ELECTRICAL SHORT
- — IN RESISTANCE-HEATED FITTING; FINAL
INJECTION TO BE MADE THROUGH FUEL
SAMPLE LINE
REACTOR MADE CRITICAL
— 10-min RUN AT 1 w FOR RADIATION
MR SyURVEY; ACTIVITY AT PUMP FOUND
TO BE DUE TO PRESENCE OF GAS LINE ‘
— NEAR DETECTOR F— ;
—1—  REACTOR PIT SEALED 1
:
i
. .
EXP. L-2 RESUMED
L REACTOR AT 1 w 1 > PREPARATIONS BEING MADE FOR FINAL
SHIM RODS RESET ADDITION OF FUEL CONCENTRATE
RUN 7 THROUGH SAMPLE LINE; FINAL CHECKS
REACTOR AT 1 w MADE PREPARATORY TO SEALING PITS
PENGUIN NO. 15 INJECTED
REACTOR AT 1 w
o~
REACTOR AT | w 0
PENGUIN NO, 18 INJECTED [ RUN 8 p .
a.
REACTOR AT 1 w X
REACTOR AT 1 w
PENGUIN NO, 17 INJECTED [ RUN 9
REACTOR AT 1 w
REACTOR AT 1 w
PENGUIN NO. 12 INJECTED p RUN 10
REACTOR AT 1 w y
Fig. 5.1.

 

 

 

 

 

 

 

 

 

 

 

NOVEMBER 6,1954

" T T T m— T

 

i b, it e il e s e o o Biie ki i e e e

ik dsdmass o e i b eanm

NOVEMBER 7, 1954

FINAL FUEL ADDITIDON; REACTOR AT 1 . w

START OF INJECTION OF 22.16 |b OF
CONCENTRATE FROM CANS 25 AND
32 PLUS 38.2 |b FROM THE 70 Ib
WITHDRAWN

COMPLETION OF INJECTION

 

)

]
|

EXP. L-é

|
|

FINAL FUEL SAMPLE (19) REMOVED FOR
ANALYSIS

 

ALL INJECTION AND SAMPLING LINES
AND EQUIPMENT REMOVED FROM PUMP
AND PITS

TANK PIT SEALED

'

REACTOR PERIOD EXCURSION
TO 100 w; RUN 4

REACTOR PERIOD EXCURSION
TO 100 w; RUN 5

ALL PIT WORK COMPLETED

REACTOR AT 1 w

PUMP STOPPED

REACTOR PERIOD EXCURSION TO 460 w
=== PUMP STARTED

~ HEAT EXCHANGER PIT SEALED; ALL PITS
NOW SEALED

REACTOR PERIOD EXCURSION
TO 100 w; RUN 7

g

 

 

EXP., L-6: SHIM ROD CALIBRATION AT | w

PERIOD EXCURSION TO 200 w; RUN 8; END
OF EXP, L-5

 

 

 

REACTOR AT 1 w FOR SHIM-ROD CALIBRATION

i e il -

 

L-5

EXP.

e i - sad oy ambia

  
  
  

" REGULATING ROD POSITION

 

 

 

 

ORNL-LR~-DWG 6400

NOVEMBER 8,1954

EXP, L-7: EFFECT OF FUEL FLOW RATE ON
DELAYED NEUTRONS; REACTOR AT | w;
FUEL FLOW VARIED O TO 46 gpm

START EXP. L-8: MEASUREMENT OF 7
TEMPERATURE COQEFFICIENT AT LOW
POWER; REACTOR AT 1 w

START HELIUM BLOWERS
RAISE THERMAL BARRIER DOORS
INCREASE HELIUM BLOWER SPEED TO 255 rpm

READIJUST SHIM RODS; LOW HEAT EXCHANGER
REVERSE; HELIUM BLOWER OFF

READJUST SHIM RODS

L-8

EXP.

OBSERVE HEAT UP AND CHANGE OF

 

END OF EXP, L-8; START OF EXP, L-9:
READJUSTMENT OF CHAMBERS FOR HIGH
POWER LEVEL

ACCIDENTAL SCRAM '

POWER 40 TO 500 w DURING ADJUSTMENT
OF CHAMBERS

PERIOD SCRAM TO CHECK SAFETY CHAMBERS;
END OF EXP. L-9

END OF LOW-POWER EXPERIMENTS

.

EXP

20-kw RUN; APPROACH TOQ POWER; EXP, H-1

REACTOR SCRAMMED; ACTIVE GASES IN
BASEMENT

4]

 

| e

 

T =
i s i s i D . i e b s s i eiiia i s oo, .l ol S o s 1) i s 1, .,

 

 

 

TRY TR T R T Y WV T e T T T Y e, T " W AR
B

oy

kb e,

e iy

i b

SRR

L

e w
4

m.‘k

Lo

it i SR Bl ok 3 L

The radiation dose data from the two runs during
which the dose in the pits was measured are pre-
sented in Tables 5.1 and 5.2, Table 5.1 gives the
gamma and the fast- and thermal-neutron doses in
the reactor pit during the 2.7-w run; Table 5.2 gives
the gamma doses at various stations in the heat
exchanger pit during both the 2.7- and 27-w runs.
Within the accuracy of the measurements, there is
excellent agreement between the doses per watt
from the two runs.

REGULATING ROD CALIBRATION FROM
FUEL ADDITIONS

Twelve small cans, or “‘penguins” (so-called
because of their shape), had been prepared for
adding small amounts of fuel to the system in order
to calibrate the regulating rod (exp. L-2) after
initial eriticality had been reached. Each penguin
contained from 0.2 to 0.5 |Ib of fuel concentrate
(0.1 to 0.3 [b of U233), The contents of the pen-
guins were injected directly into the pump rather
than through the intermediate transfer pot that was
used in the critical experiments. Table 5.3 lists
the penguins in the chronological order in which
they were used, together with the amount of yranium
injected from each can,

The method used to calibrate the regulating rod
was to note the difference between the position of
the rod at criticality before and after each injection
of fuel, with the shim rods held constant. Before
each injection the reactor was brought from sub-
critical to a nominal power of 1 w, and the shim
and regulating rod positions were recorded. The
reactor was then brought subcritical on the regu-
lating rod, with the shim rods in position, and the

fuel was injected from the penguin, Upon comple-
tion of the injection the reactor was again brought
critical to l-watt power with the regulating rod.
The positions of the rods were again recorded.
The worth of the rod was then obtained from the
relation

Ak AM
k M

and the known increment of fuel added to the sys-
tem; the proportionality factor, 0.236, was obtained
experimentally during the critical experiment

(cf., chap. 4).

The first run of experiment L-2 started at 1651
on November 4. After two injections had been
completed, it was apparent that the worth of the
12 in. of vertical movement of the regulating rod in
terms of Ak/k was about 0.24%. It was desirable,
for reasons of safety and also for convenience in
conducting high-power transient experiments, fto
have a rod worth about one dollar of reactivity
(which for the ARE was 0.4% rather than 0.76% as
in stationary-fuel reactors, because of delayed-
neutron loss in the circulating fuel). A number of
spare rods of varying weights had been made up to
take care of such a contingency, and one was
selected and installed which had more nearly the
desired weight. The original rod weighed 19.2 g/em
of length, and the new rod weighed 36 g/cm.

While the rod was being changed (a delay of
about 5 hr), the fuel system characteristics of
pressure head vs flow were obtained. These
measurements are described in a following section
of this chapter,

TABLE 5.1. RADIATION LEVELS IN REACTOR PIT DURING 2,7-w RUN

 

 

Gamma Fast-Neutron Thermal-Neutron Total
Position Dose Rate Dose Rate Dose Rate Dose Rate*
{mr/hr per w) (mrep/hr per w) {mrep/hr per w) (me/hr)
Space cooler No. 1 48 67 8 760
Top center of reactor : 280 280 35 (est.) 3200
(1'in. above thermal shield)
Side of reactor at mid-plane 440 180 22 {est.) 2350
(at surface of thermal
shield)
Manhole into pits 12 10 2 130

 

*The total dose rate was obtained from the weighted sum of the preceding columns by using an RBE of 10 for fast

neutrons and 5 for thermal neutrons.

43

 

o T T e T

T

 
el L

TABLE 5.2. GAMMA-RAY DOSE IN TANK AND HEAT EXCHANGER PITS

 

» i s e e

Doses per Watt
During 2,7-w Run

Doses per Watt
During 27-w Run

 

 

 

 

(mr/hr) (me/hr)
Locations Surveyed in Dump Tank Pit
Motor for space cooler No, 3 0.3 0.3
Vent valve U-113 for tank No. 6 0.1 0.15
- Motor for stack vent on hot fuel dump tank 0.2 0.15
Vent Valve U-112 for hot fuel dump tank 0.1 0.07
: Helium inlet valve U-100 0.07 0.07
Motor for space cooler No. 4 0.07
Air-operated valve for tank No, 5 0.07
Locations Surveyed in Fuel-to-Helium Heat Exchanger Pit
Operator’s position at fuel sampler 6 4
Pressure transmitter PXT-6 22
Top plate of main fuel pump 150 120
Top bearing of pump motor 45 80
V-belt at pump . 7 6
V-belt at pump motor 7 4
Line 120 under valve U-3 110 12
Bend No, 2 of line 303 105 45
Motor for space cooler No. 6 10
West side of heat exchanger No. 1 135 120
Bottom of fuel flowmeter 95 105
Line 112 between valves U-63 and U-1 220 200
Lubrication pump for pump 65 90
Top of fuel storage tank 4
Helium analyzer dryers 0.3 1
3 East side of heat exchanger No. 1 230
West side of heat exchanger No. 2 200
Sheet metal can around pump 185
L ocations Surveyed in Scdium-to-Helium Heat Exchanger Pits
Top of standby sodium pump 1.5 1.4
Y-belt above standby pump 1.5 1.7
Line 310 under vaive B-141 13 15
Motor for space cooler No. 7 4
Top of main sodium pump 1.5 0.7
V-belt above main pump 0.7 0.8
Line 309 at valve U-21 2.2 1.8
Line 313 at bend No. 5 1.5 1.5
Motor for space cooler No, 8 1.5
North side of front heat exchanger 15 16
Kfnney pump No. 1 4 4
Magnetic clutch on 50-hp motor 1 13
North side of back heat exchanger 1 6
North end of rod-coolant blower No. 1 6 7
EM flowmeter on line 305 3.7 5
Standby pump lubrication system 9
Main pump lubrication system 3

44

 

.

o

- T T * T '?'r‘!‘?"*“’&fl_”'

.

T

© g

It R T o e e e ety v e e B poy e Pormrr v v v B e 2 e o

 
i & kL

-t s b

~anbod R A

i LR A .

hboeg

S

R ek K R R el

e, ok =

- s R, i,

o B

i

TABLE 5.3. FUEL CONCENTRATE BATCHES ADDED DURING LOW-POWER E XPERIMENTS

 

 

 

_ Ponguin Concentrate Added Uranium Added U235 Added
Date Time
No. g wt % g g b
11/4 1710 C-11 227 0.5004 59.671 135 126 0.2778
2009 C-19 722 1.5917 59.671 43 403 0.8885
11/5 0354 C-14 247 0.5445 59.671 147 137 0.3020
0439 C-16 285 0.6283 59.671 170 159 0.3505
0529 C-13 131 0.2888 59.671 78.2 73.0 0.1609
0619 C-5 123 0.2712 59.671 73.3 68.5 0.1510
2045 C-15 92 0.2028 59.671 55 51.4 0.1143
2204 C-18 87 0.1918 59.671 52 48.6 0.1071
2258 C-17 457 1.0075 59.671 273 255 0.5622
2345 C-12 84 0.1852 59.671 50 46.7 0.1030
11/6 0100 C-10 84 0.1852 59.671 50 46.7 0.1030
11/7 0231 120% 21,681 47.798 27.557 5975 5581 12.304

 

*Container No. 120 was not a penguin but a specially prepared batch of concentrate which provided the excess
uranium required for the experiments and to compensate for burnup.

At 0142 on November 5 the installation of the
second rod was completed and the experiment was
resumed. Four more injections of fuel were made
and the rod movement was noted. At 0630, during
the sixth fuel injection, an unexpectedly high
burst of gamma-ray activity (55 mr/hr) was ob-
served on a ‘‘Cutie-Pie’’ monitoring the fuel addi-
tion at the injection station over the pump. Further
fuel addition was stopped until the cause of the
high gamma-ray reading could be ascertained.
While investigation was under way the level of the
pump was trimmed to avoid any possible hazard
from the uranium held in the pump. About 71 .Ib of
fuel was removed from the system. Later in the
day the cause of the high gamma-ray activity was
determined to be a vent line passing close to the
position of the ‘'Cutie-Pie.””  Apparently, the
“‘Cutie-Pie”" had been held close enough to the
vent line to read the fission gases (evolved during
operation) as they were discharged through the

vent line.

At 1947 the experiment was resumed and five
more injections were made uneventfully, A photo-
graph of the control room taken during this time is
presented in Fig. 5.2; it shows two members of the
ARE operating group examining the fission chamber
recorders after a fuel injection for rod calibration;
other members of the evening crew are shown in

their nominal operating stations.

An inventory of the uranium added during this
experiment is given in Table 5.4. The final large
amount (47.8 Ib) of concentrate containing 12.30 ib
U235 that was added after the experiment is also
shown in the table, as well as a record of the
samples and withdrawals, The final amount of
U235 jn the system was 138.55 Ib, with a system
volume of 5.33 fi3,

The data obtained for calibration of the regulating
rod as a function of fuel addition are given in
Table 5.5, which lists the values of M, AM, AM/M,
the recorded movement of the rod during each fuel
addition, the average position of the rod for each
fuel addition, and the calculated value of (Ak/k)/in.

The results of the calibration are shown in Fig,
5.3, where (Ak/k)/in. is plotted as a function of
rod position. The movement of the rod for each
point is shown as a horizontal line through the
point. The volues of (Ak/k)/in. for both rods were
used, the values of reactivity for rod No. 1 being
corrected by the ratios of the weights of the two
rods. From these data it is evident that there is no
good reason for presuming that the value of
(Ak/E)/in, over the whole length of rod is not
constant. Therefore, the average value taken over
the first ten runs gives

(Ak/E)/in. = 0.033%/in,

45

N

[ —

-l

 
R e R

 

   

9w

 

 

rro® coep T e T s egmerrpt T e T g e e T e e B e et

 
b ke

el -

Bl ., g

iRk ke oas s

S

Coulle k.

W .

[

b

o amRS

g

" A

N om

kit g .

R

LS

SR

Wb bk

TABLE 5.4. URANIUM INVENTORY DURING LOW-POWER EXPERIMENTS

TR T ———

 

 

 

 

 

Samples Removed Fuel Mixture Uranium-235
Fuel Concentrate Added for Analysis Total Weight _ Total
Date Time NL:: Weight Yolume Fuel y23s Concentrate Vi?::wle Density X;:fi}:f Weight Concentration
(Ib) (f13) Removed Removed Plus Carrier (F13) (Ib /53 (Ib) in System  |b/#3  wt%
(ib) {ib) (tb) (b}
/4 1315 0 ) 114745 55454 206.9 132,763 23.94 1157 F
1710 1 0.5004 0.0017 1147.95 5.5471 2069  0.2778 133.041 23.98 11.59
2009 2 1.5917 0.0056 1149.54 5.5527 207.0 0.8885 133.930 2412 11.65 !
/5 035¢ 3 0.5445 0.0019 1150.08 5.5546 207.05 0.3020 134232 2417 11.67 '
0439 4 0.6283 0.0012 1150.71 5.5558 207.1 0.3505 134.583 2422 11.70
0529 5 0.2888 ¢.0010 1151.00 5.5568 207.1 0.1609 134.744 2425 11.71
0619 6 0.2712 0.0009 1151.27 5.5577  207.1 0.1510 134.895 24.27 11.72 (
1015 53.486* 6.269 1097.78 5.2944 207 128.626  24.27 11.72 E
1020 17.523* 2.064 1080.26 5.2148 207.1 126.562 2427 11.72 ;
1100 2.5807 0.2%1 1077.68 52023 207.1 126.271  24.27 1172 i
2045 7 0.2028 0.0007 1077.88 5,2030 207.2 0.1143 126385 2429 11.73 :
2204 8 0.1918 0.0004 1078.07 52034 207.2  0.1071 126492 2431 1173
2258 9 1.0075 ¢.0020 1079.08 5.2054 207.3 0.5622 127.054 2441 11.77
2345 10 0.1852 0.0004 1079.27 52058 2073 0,030 127,157 2443 1178
11/6 0160 11 0.1852 0.0004 1079.45 5.2062 2073 0.1030 127.260 24.44 1179 ’
0530 2.686 0.317 1076.76 5.1932 2073 126.943 2444 1179 l
0535 2.839 0.335 1073.92 51795 2073 126.608  24.44 1179 ~
/7 0231 12 47.798**  0.1667 1121.72 5.3462 209.8 12304 138912 2598 1238
. 0423 2.918 0.361 1118.80 5.3323 209.8 138.551 2598 1238

 

*Two large batches of fluoride mixture were removed from the system in order to frim the liquid level in the pump.
**+0f this amount, only 15.419 |b was N02UF6, the remainder being some of the fluoride mixture that was removed from the system when the
pump level was trimmed.

ORNL-LR-DWG 6401

 

© e e e e

 

 

 

 

 

REACTIVITY (% /in))

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

0.05

| o —ROD COMPLETELY ROD COMPLETELY

INSERTED , WITHDRAWN \ !
0.04 —~— o -
WEIGHTED AVERAGE [0
L~ D AVE
0.03 —
it
0.02
C.01
0 :
2 3 4 5 6 7 8 9 10 " i2 13 14

REGULATING ROD PQOSITION (in.)

:
k
L
E

Fig. 5.3. Calibration of Regulating Rod from Fuel Addition,

47

 
 

 

 

\

TABLE 5.5. REGULATING ROD REACTIVITY CALIBRATION FROM FUEL ADDITION E

(Exp. L-2) F

|

M, 235 AM, 23 Average Rifl::;?:; F:Zd Reactivity, k

Run  Weight of U Weight of U233 Am/m Ak/k Regulating Rod During Fuel (Ab/k)in.

No. in System Added (%) Position Addition (%/in) .k

(Ib) (1b) (in.) , .

(in.) ;

] 133,041 0.2778 0.00209  0.0493 11.80 2.4 ©0.0308* . E

2 133.930 0.8885 0.00663 0,156 9.35 7.3 | 0,0400% E

3 134.232 0.3020 0.00225  0.0531 10.70 1.6 0.0335 fi

4 134.583 0.3505 0.00260  0.0614 8.80 1.9 0.0323 E

5 134,744 0.1609 0.00119  0.0281 6.80 0.85 0.0330 E

6 134.895 0.1510 0,00112  0.0264 6.00 1.0 0.,0264 é

7 126.385 0.1143 0.000904  0.0213 6.35 0.7 0.0304 b
8 126,492 0.1070 0.000847  0.0200 6.00 0.6 0.0333
9 127.054 0.5622 0.00442  0.104 5.10 3.6 0.0289
10 127.157 0.1030 0.000810  0.0191 12.25 0.5 0.0383

11 127.260 0.1030 0.000809  0.0191 2.95 0.67** 0.0305%+*

 

Average reactivity, weighted according to rod movement 0.0331

 

*Corrected by the ratio of the weight of the new rod to the weight of the original rod: 36/19.2 = 1.88,

** Average of three readings.

The value from run 11 was not included in this  scrammed. From the induced period and the known S
average because three different values of rod final flow, the excess reactivity introduced during the £
position were recorded in different places for this  run was obtained from the inhour formula, This '
run. number was then divided by the travel of the regu- £
lating rod to find the value of (Ak/k)/in. for the i

REGULATING ROD CALIBRATION FROM run, A total of eight reactor excursions of this ¥
REACTOR PERIODS nature were made, six at 48-gpm flow, and two at -

The calibration of the regulating rod by use of  zero flow. g
reactor periods (exp. L-5) utilized the inhour equa- Some of the reactor excursions during this ex- ‘
tion in a form in which reactivity is given in terms periment are shown in Fig. 5.5, as recorded by the #
of the reactor period for a given rate of fuel fiow, log N recorder. The straight lines that indicate E
Figure 5.4 shows a plot of the reactivity {Ak/k) as  the slope of the power increase reflect periods on
a function of reactor periods for 0- and 48-gpm the order of 20 to 25 sec. Figure 5.6 presents some F
fuel flow, as calculated from the inhour relation,  of the period traces recorded on the period recorder :
Details of the calculational method used are pre-  during the experiment. The initial high peaks for E
sented in Appendix I, each rod movement are due to the transient condi- k
In the experimental procedure followed, the regu-  tion, the interpretation of which is given in ;
lating rod was suddenly withdrawn a known dis- Appendix S. In plotting the period for a particular “

tance, while the shim rods were in a set position
and the reactor was operating at a nominal power
of about 1 w. The reactor was then allowed to
rise in power on a constant period to a peak of
50 to 100 w, at which time it was manually

48

rod motion, the average of the periods obtained
from the log N and the period recorders was used,
A plot of (Ak/k)/in. as a function of rod position
as obtained from this experiment is shown in
Fig. 5.7. The horizontal lines on each side of the

R

T U ey - wy R T

 
Cakii .

R e e

BA. -

ol o

b B ia

ehcd Gomdges.d sl g

ORNL—-LR—DWG 8402

 

0.0035

 

0.0030 \

 

0.0025 .

ZERO FLOW
L

 

0.002C \\

0.0015 \

 

REACTIVITY, Ak/k

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

0.0010 ]
\\
—
0.0005 T~
\
—
0
0 10 20 30 40 50 60 70 80 20 100

REACTOR PERIOD (sec)

Fig. 5.4. Regulating Rod Reactivity vs Reactor Period for Fuel Flow Rates of 0 and 48 gpm.

experimental points show the extent of the rod
movement for that point. Again, there is no con-
clusive evidence that the reactivity of the rod per
vnit length was not constant.  Therefore, the
average of the 48-gpm flow data weighted over the
rod movement for each run gives

(Ak/R)/in. = 0.029%/in.

For the two zero-flow points the rod was moved
over the middle 10 in. of its 12-in. travel in two
6-in. overlapping runs covering the upper and lower
halves of the rod, respectively, The weighted
average of these points gave a value of

(Ak/k)/in. = 0.033%/in. ,

which is in excellent agreement with the fuel
addition calibration, This value was settled upon
as the best value of reactivity per inch over the
whole length of the regulating rod, The values for
the reactivity vs rod position are given in Table

5.6.

CALIBRATION OF SHIM RODS VS
REGULATING ROD

With the reactivity of the regulating rod known,
the shim rods could be calibrated against it (exp.
.-6). This calibration could then be compared
with the previous calibration made during the fuel
addition. It was assumed that the three shim rods
were enough alike that a calibration of one rod
would be representative, and rod No. 3 was chosen
for the calibration,

The calibration procedure consisted of setting
shim rods Nos. 1 and 2 in a fixed position, and
then, with the reactor at a power level of 1 w on
the servo mechanism, rod No. 3 was moved until a
specified travel of the regulating rod was obtained.
At the start of the experiment the rods were ad-
justed so that shim rod No. 3 was nearly all the
way out (position indicator at 35 in.) and the regu-
lating rod nearly all inserted (position indicator at
3 in.). After recording the position of all rods,

49

= T g

 
 

tn
S

ORNL-LR-DWG 6403
1000

2 —— RUN2,NOV. 6, 0715 RUN 3, NOV. 6, 0950 — RUN 5, NOV. 7, 1810
FUEL FLOW, 46 gpm FUEL FLOW, O gpm FUEL FLOW, 46 gpm

RUN 8, NOV. 7, 2355
FUEL FLOW, 46 gpm

100

REACTOR POWER (watts)

1 1w, NOMINAL POWER

 

0.1

—~—— TIME

Fig. 5.5. Typical Reactor Excursions Recorded by Log N Recorder During Period Calibration of Regulating Rod. Experiment L-5,

L4 ¥

 

wee s WF R OT P ) M wh o e CEETTE CUpcin e g e, TEPECCRIE T TEehr mertrie R gt ST Yo omEEs “‘"‘!’m"m Cr o Mg e fEEUY R O EENEEUres o EUTTET T e rer Y g S tr. glewr Tt R T rEmpgmes g P »
16

REACTOR PERIOD, (sec)

RUN 3, NOV. 8, 0715
FUEL FLOW, O gpm

RUN 4, NOV. 7, 1610
8 FUEL FLOW, 46 gpm

10

20
30

100

—{00

-30

 

RUN 5, NOV. 7, 1808
FUEL FLOW, 46 gpm

RUN 6, NOV. 7, 2034
FUEL FLOW, Ogpm

- -

TIME

ORNL-LR-DWG 6404

RUN 7, NOV. 7, 2208
FUEL FLOW, 46 gpm

RUN 8, NOV. 7, 23
FUEL FLOW, 46 gpm

 

Fig. 5.6. Reactor Periods Recorded by the Period Recorder During Calibration of Regulating Rod, Experiment L-5.

 

 

- e T T - R TR =~ T
 

 

 

ORNL-LR-DWG 6405

 

0.40

FUEL FLOW, Cgpm

 

0.35%

 

 

 

REACTIVITY (% /in.}

 

ROD COMPLETELY
INSERTED

 

 

 

 

 

0.25 I

 

ROD COMPLETELY
WITHDRAWN ——_,_ |

 

 

 

 

 

 

 

0.04 ‘ I

WEIGHTED AVERAGE

 

 

 

003 —F—TF—— —— —

 

 

’
!

 

*

 

 

0.02

REACTWITY (% fin.)

FUEL FLOW, 46 gpm

 

0.04

 

 

 

 

 

 

 

 

 

 

 

 

 

 

2 3 4 5 6 7

8 9 10 1 12 13 14

REGULATING ROD POSITION (in.}

Fig. 5.7. Calibration of Regulating Rod from Reactor Period Measurements,

shim rod No. 3 was inserted until the servo action
had withdrawn the regulating rod a compensating
distance of about 10 in. The action was then
stopped and the new rod positions were recorded.
With shim rod No. 3 set, shim rods Nos. 1 and 2
were then adjusted to bring the regulating rod back
to its original position, and again the rod positions
were recorded, Again rod No. 3 was inserted until
a 10-in. withdrawal of the regulating rod was
attained.  This process was repeated until rod
No. 3 had been inserted about 14 in,, and each
new position had been recorded. This calibration
agreed well with the calibration against fuel addi-
tion. Details of the results are given in Appendix J.

FUEL SYSTEM CHARACTERISTICS

The fuel system characteristics with the fuel
concentrate in the system were determined for the

52

first (and only) time the day after the reactor first
became critical (exp. L-3). The final amounts of
concentrate (i.e., the excess to provide for burnup
and poisoning during the subsequent power runs)
had not yet been added to the system, Conse-
quently, the fuel density at the time the data were
taken was 3.32 g/cm3 as compared with 3.36 g/cm?
for the final fuel composition, and 3.08 g/cm® for
the carrier; all densities were determined at 1300°F.

The measured and calculated flow data as a
function of pump speed are shown in Fig. 5.8. The
flow was calculated for two pump speeds by using
the time between pips on the fission chamber as
concentrate was first added to the system. The
straight line joining the two calculated points is
considered to be a good approximation because of
the regime in which the pump was operated.

e b oy

e gy e "o

»e

et ey

s

.

P e g

Theer ot ot

-

—p . e TTT F oo oy o

e o s

wErmeToomes mewer WET Y roeerze MF

TRE ey T meey P

 
ke, i bl

e g o R S,

g,

R%

T Y

el

s iR, B R

o e, Sk

S

b A BREREE |

TABLE 5.6. REGULATING ROD REACTIVITY CALIBRATION FROM REACTOR PERIOD MEASUREMENTS

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

  
 

 

 

 

 

 

 

 

 

 

 

 

(Exp. L-5)
A
Evel FI Average Ak/E R Ivu;-;l:ageR 4 Movement of Reactivity,
Run No. U? ow Period from Inhour eg; a .“Tg ° Regulating Rod (Ak/R)/in.
gpm) (sec) Equation os'tt|§n (in.) (%/in.)
. (in.)
1 48 48 0.00036 7.5 1.1 0.0318
2 48 22,2 0.00060 7.6 2.C8 0.0289
3 0 21 0.00206 9.6 6.05 0.0330
4 48 26 0.00054 3.8 2.0 0.0270
S 48 22 0.00061 11.1 2.2 0.0278
6 0 21.6 0.0020 5.6 ‘ 6.1 0.0328
7 48 20.1 0.00065 5.2 2,08 0.0313
8 48 22.2 0.00060 9.3 2.06 0.0291
Average for 0-gpm flow  0.0329
Average for 48-gpm flow 0.029
1200 ORNL-LR-DWG 6408 ORNL-LR—DWG 6407
L 30
FLOW CALCULATED FROM PIPS /O/'
1000 [— AT 1080 AND 520 rpm ASSUMING //
CIRCULATING VOLUME TO BE //' 25 ACTUAL PRESSURE DATA (FLOW CALCULATED FR(!M
3 - P .
E s00 482 f17{ASIN Fig. 4.4) /e/ TIME BETWEEN PIPS ON FISSION CHAMBER )~ /
2 //; °/
& 600 A& / //
v /< 20 7/
% /P/ FLOW FROM FLOWMETER = /
o. P ] =
400 v g //
WRE .,/
200 {7~ 7 , .
& o /
T /
0 S
0 10 20 30 40 50 60 10 7
FLOW RATE (gpm) /‘/ .//
,o/ /
//
Fig. 5.8. Fuel Flow Rate as a Function of Pump 5 4!\
Speed. ' o | “PRESSURE CORRECTED
- (FLOW FROM PIPS)
‘ . A '
The flow data as a function of the system head 0l
0 {0 20 30 40 50

loss from the reactor inlet to the pump tank are
given in Fig. 5.9. The flow data are plotted di-
rectly vs head (by using the flow rates determined
from pump speed as given by the upper curve in
Fig. 5.8) and then as corrected for a ‘‘live’’ zero

FLOW RATE {gpm)

Fig. 5.9. Fuel Flow Rate vs the Pressure Head
from the Pump Suction to the Reactor Inlet.

53

 

 
on the pressure transmitter from which the head
values were obtained. The latter data fit the
theoretical curve very well, Furthermore, these
system data fit the pump performance datq, if it is
assumed that there is a 7-psi drop from the pump
discharge to the reactor inlet at design flow (i.e.,
46 gpm); there is no experimental measurement of
this pressure drop. The pump performance charac-
teristics determined in a test stand run with fuel at

1300°F are given in Fig. 5.10,

“ EFFECT OF FUEL FLOW ON REACTIVITY

, An experiment was performed in which the effect
of fuel flow on reactivity was observed (exp.L-7).
’ For this experiment the reactor was brought to a
nominal power of 1 watt and given over to the
flux servo mechanism after full fuel flow of 46 gpm
had been established. After recording the positions
of the regulating and shim rods the flow rate was
reduced step-wise until zero-flow was reached.
Each reduction of the flow was accompanied by a
change of position of the regulating rod. Reducing

 

 

the flow had the effect of allowing more delayed
neutrons to contribute to the flux level in the
reactor, and therefore the servo mechanism inserted
the regulating rod to compensate for the excess
reactivity created by the delayed neutrons. Each
rate of fuel flow and the corresponding regulating
rod position were recorded. The results of the
experiment are shown in Fig, 5.11, where AL/k is
plotted as a function of flow rate. It is observed
that 12 in. of rod movement, or 0.4% Ak/k, was
needed to compensate for the reactivity introduced
when the flow rate was reduced from full to zero
flow. This curve is comparable to that shown in
Fig. D.2 of Appendix D. The data for this experi-
ment are tabulated in Table 5.7,

Although not a part of experiment L-7, additional
insight into fuel activity may be obtained from
examination of Fig. 5.12, which shows activity as
recorded by fission chamber No. 2. Before the
data shown in the figure were recorded the reactor
had been operating at a power of about 1 w with
the fuel pump stopped. As shown, the reactor was

ORNL—-LR-DWG 4524
60

 

 

 

 

 

 

 

 

 

 

 

 

60
50 {400 rpm| 50
. \ 20
e EFFICIENCY AT 1200 rpm
{200 rpm g’
— ey <
T / ;
O P —
30 30 o
SCJ \ S
’ E
ui

 

/ 800 rpm

 

// 1000 rpm
——
20

A T

\
\‘_ 20
T~

 

 

 

 

 

 

 

 

 

 

 

 

54

 

600 rpm
\
o - 0
0 0 20 30 40 50 80 70 80 90
' FLOW (gpm)

Fig. 5.10. Fuel Pump Performance Curves.,

ey P

 

F e T

g - T

.l.
W

;W T W’F"@'W‘”W*’ »W' e T

” "'WF =

 
o SRR

e e e

P

L

s o

%

ORNL-LR-DWG 6408
04 {

 

-+ 105

03t 9.0

 

<4 7.5

;
\ | a5

\. REACTIVITY CHANGE DUE TO
(o] ) SRR REE . STOPPING SODIUM FLOW
\. |
\.‘1
\l —H 15
i\\a} ®
0 5 10 15 20 25 30 35 40 45 50
FLOW RATE (gpm)

 

 

 

 

 

 

30

 

 

EXCESS REACTIVITY Ak fk (%)
o
av
&
=}
INSERTION OF REGULATING ROD (in)

 

 

 

 

 

 

 

 

 

 

0

 

Fig. 5.11. Effect of Fuel Flow Rate on Reac-
fiVifYQ

TABLE 5.7. EFFECT OF FUEL FLOW RATE
ON REACTIVITY

 

 

(Exp. L-7)
Fuel Sodium Movement of Change in
Run Flow Flow Regulating Rod  Reactivity,
e (gpm)  (gpm) (in.) Ak/k (%)
1 45,5 149 0 0
2 41 149 0.7 0.0231
3 35 149 1.4 0.0462
4 28.5 149 1.9 0.0627
5 238 149 2.4 " 0.0792
6 18 149 3.0 0.099
7 13.5 149 3.45 0.114
8 0 149 12.05 0.398
9 13.5 149 3.45 0.114
10 0 149 12,05 0.398
11 13.5 149 3.65 0.120
12 13.5 149 3.65 0.120
13 43.5 149 0.9 0.0297
14 41 149 0.96 0.0317
15 41 0 1.12 0.0370

16 41 149 0.96 0.0317

 

then scrammed! and the fuel pump started, The
fuel which had been in the reactor before the scram
was more active than the remainder of the fuel and

 

]Scrum, as used in this report, refers to an intentional
shut down of the chain reaction by suddenly dropping
the contrdl rods into the core.

ORNL-LR-DWG 3854

—T— REACTOR SCRAMMED ‘

. DATA FROM FISSION CHAMBER NO. 2
: 3 EXP. L-5, RUN 6, NOV. 7,1954 ==
T IEITETT (READ FROM RIGHT TO LEFT)

1) i PUMP STARTED

 

Fig. 5.12. Circulating Fuel Slugs after Starting
Pump Following a Reactor Scram.

it was this excess activity which showed up as the
fuel was circulated after the reactor was scrammed.

LOW-POWER MEASUREMENT OF THE
TEMPERATURE COEFFICIENT

A measurement of the temperature coefficient of
reactivity (exp. L-8) was made next. With the
reactor isothermal at 1312°F and controlled by
the flux servo mechanism, the heat barrier doors
were raised and the helium blower turned on to
200 rpm to cool the fuel in the heat exchanger.
As the temperature of the reactor decreased, the
servo began to drive the regulating rod in to com-
pensate for the increased reactivity, The recording
of the regulating rod position vs time is shown in
Fig. 5.13. Beginning at 0218, the rod was in-
serted by the servo quite rapidly, and by 0221 the
With the
regulating rod on servo, one of the shim rods was
inserted as rapidly as possible. This shim rod
insertion overcompensated for the increase in
reactivity due to the temperature drop, and the

rod was approaching its lower limit,

55

- TP P T VT

[

T

.

v g T

T

T R e T e

 
4
i
i
4

¢

 

 

ORNL—LR-DWG €409

 

02492

 

 

 

0244 e e o b
0240 | ol WITHDRAWAL

0239 b — e INSERTION
—— -

 

G238
0237
0236
0235

 

EXP. L-8, NOV. 8, 1954
0234 .

0233
0232

 

 

0234
0230
0229
0228
0227

TIME OF DAY

Q226
0225

 

 

 

0224 ;
0223 '

 

\‘_\ ‘ L. UPPER END OF |
/ ROD TRAVEL
- |

 

 

 

 

 

 

 

 

 

 

-« BLOWER OFF -

 

 

 

 

 

 

 

 

 

I
e} ROD WITHDRAWN

 

 

0222 i

 

 

 

0220
0219
oz18

0217

 

 

0218 by
02145 ONE INCH
REGULATING ROD
0214 TRAVEL

Q213

 

 

 

0212 ‘ : 5 ; -

 

BY INSERTING SHIMS
TO OVERRIDE SERVO |

LOWER END OF
b\-\‘_L"T\ T ROD TRAVEL |,
| O

BLOWER ON {255 rpm)

 

 

 

 

T 0 | R —
.- oL _,}\START TO RAISE B
HEAT BARRIER DOORS

"

 

 

 

 

 

 

 

 

13 12 i 10 9 8 7

5 4 3 2 1 0

REGULATING ROD INSERTED (in.)

Fig. 5.13. Regulating Rod Position During Low-Power Measurement of the Reactor Temperature

Coefficient,

servo quickly withdrew the regulating rod. In-
sertion of the shim rod was stopped when the
regulating rod had been withdrawn to near its
vpper limit, and the regulating rod was again in-
serted by the servo to compensate for the reactivity
introduced by the cooling of the fuel. By 0230, the
outlet temperature in the heat exchanger was

-approaching its lower limit, 1150°F, so the helium

blower was stopped and the run was ended.

A trace from the chart for recording the mean fuel
temperature is shown in Fig. 5.14, and super-
imposed on it is a plot of the displacement of the

56

regulating rod as obtained from the previous figure,
A measure of the temperature coefficient may be
obtained by comparing the slopes of the two curves,
At 0228 the rod was moving at exactly 1 in./min,
corresponding to an increase in reactivity of
3.3 x 10~4 (Ak/k)/min. At the same time, the
mean temperature was dropping at the rate of
5.1°F/min. The resulting reactor temperature
coefficient, as obtained from these data, would
appear to be ~6.48 x 10=° (Ak/k)/°F. By com-
paring the slopes of the curves obtained earlier
during the run, a considerably higher coefficient

- e

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ORNL-LR-DWG 6410

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

[ 400
)
£
T T ERRER"
Ll o~ - b Le.
3z o LrN) 1 o
S« o £XP. L~8 NOV. 8,1954 = T
O & = = o
i o o “
2
— ‘3:“135 x Yo —1 1350 U
Sea |En - 2 - 2
- I w VES &1
= T
¢ - — - EN\ENL’ N Ll
= \’pC - n
q 0%~ b
5 16 A \—— -------- : 1300
» <t
o \ / L
o _ \ =
Q . i
@ w
. ! \\fUEL M o
An e
/ & L~
/
o P
0215 0220 0225 0230 0235
TIME OF DAY

Fig. 5.14. Regulating Rod Position Superimposed on Fuel Mean Temperature Chart,

can be obtained that might approximate that for the
fuel. In any case, the recorded mean temperature
is believed to be in error (as explained in Ap-
pendix K) in such a manner that the actual rate of
temperature change is about 40% higher than that
given in Fig. 5.14. The measured value of the
temperature coefficient was corrected accordingly
to give a value of —4.6 (Ak/k)/°F for the low power.
This value is comparable to that subsequently
obtained during high-power operation and, as noted
in the subsequent discussion, the data from experi-
ment H-8 yielded the best values for both the fuel
and the reactor temperature coefficients,

The apparent time lag between the reactivity and
the minimum mean temperature, as first noted during
the subcritical measurement of the femperature
coefficient, was still in evidence here. Possible
explanations of the time lag, as well as other

details of the various temperature coefficient

measurements, are given in Appendix O.

ADJUSTMENT OF CHAMBER POSITION

During the critical experiment and the low-power
operation, it was desirable fo have the nuclear
instruments positioned to read at as low a power

as possible. The chambers were therefore as

close to the reactor as possible. Before increasing
the power, it was necessary to adjust the positions
of the nuclear chambers so that they would read at
higher power levels. This was accomplished by
adjusting the power level of the reactor to about
100 w and then moving the compensated log N
chamber away from the reactor so that its reading
was changed from 0.032 to 0.001, or its upper limit
was extended by a factor of about 30, With the
reactor power at about 1 kw, the micromicroammeter
chamber was moved away from the reactor so that
its reading was changed from 41.4 on the 5 x 107
amp range to 54 on the 1 x 10~8 amp range. This
extended the upper limit of the meter by a factor of
about 40. While one chamber was adjusted the
reactor power was held constant by the other
chambers so that the correlation between instru-
ment readings was not lost. The nuclear power for
the experiment was determined on this basis.

in the final analysis, the correlation between
chamber readings and reactor power can be based

~on.the 25-hr xenon run during which a good value

for the extracted power was obtained. The micro-
microammeter read 52.5 on the 2 x 10~3 scale
when the log N read 22.5, and both corresponded
to an extracted power of 2.12 Mw.

57

 

 
 

6. HIGH-POWER EXPERIMENTS

The low-power tests were concluded by noon on
November 8. During the last few days of the low-
power tests all last minute details in preparing the
system for high-power operation had been under
way. These preparations included lubricating all
rotating equipment, refilling oil reservoirs, chang-
ing filters, replacing motor brushes, checking all
instrumentation, etc. By 1445 on November 8 the
pits had been sealed, i.e., covered with three
layers of 2‘/2-ft-thick concrete blocks and calked,
and the experiment was ready for the high-power
phase of the operation, the chronology of which is
given in Fig. 6.1.

Prior to the start of the high-power experiment
the final addition of fuel concentrate, which was
to provide the reactor with sufficient excess
reactivity to overcome burnup and fission-product

' poisoning, was made. This last addition consisted

of 15.419 Ib of concentrate diluted with 32.379 Ib
of partially enriched fuel (part of that removed from
the system earlier when the pump level was trimmed)
and was injected into the system through the
sample line on the morning of November 7. The
sample line was used for this final enrichment
operation because the transfer line through which
the concentrate had hitherto been injected into the
system had again developed a leak. It was neces-
sary to lower the melting point of the concentrate
(by dilution) because it was not certain that the
sample line could otherwise safely maintain the
high temperature required.

High power, which may be defined for the ARE
as anything over 0.1 Mw, was approached in a
series of successive steps of increasing power.
The power regime was not attained at once, because
leakage of fission activity from the system into the
pits and subsequently from the pits into the oc-
cupied parts of the building required that provision
be made to maintain the pits at a subatmospheric
pressure. The experiment could then continue
safely, and a power of greater than 2 Mw was first
attained early during the evening of November 9.

The remaining time (70 hr) of the ARE operation
was devoted to several experiments to determine
the high-power behavior of the reactor; these ex-
periments are described in detail in the following
sections. Starting in the evening of November 9,
the fuel temperature coefficient of reactivity (Exp.
H-4) ond the over-all reactor temperature co-
efficient (Exp. H-5 and Exp. H-9) were measured.

58

The early hours of November 10 were used to take
the reactor to power from subcritical by making use
of the negative temperature coefficient (Exp. H-6).
Later in the day the sodium temperature coefficient
was measured (Exp. H-7), some other reactor
kinetics were studied (Exp. H-8), and the moderator
temperature coefficient was measured (Exp. H-10).
During this time the operating crews had opportunity
to familiarize themselves with the handling of the
reactor at power.

At 1835 on November 10, a 25-hr full-power xenon
run was started. For this run the reactor was held
in steady-state operation at 2.12 Mw by the negative
temperature coefficient and at a mean temperature
of 1311°F. Since equilibrium temperature condi-
tions prevailed during this run, a very good meas-
vrement of extracted power could be made. The
run ended at 1935 on November 11, Later that
night, the effect on the power of stopping the
sodium flow was observed (Exp. H-12), and at
2237 a ]/m-power run (Exp. H-13) was started to
observe the effect of xenon buildup and continued
for a period of 10 hr. During this run the reactor
power was approximately 200 kw. At 0835 on
November 12 no effect of xenon buildup had been
observed, and therefore the experiment was termi-
nated.,

The morning of the last day, November 12, was
devoted to some more reactor kinetics studies and
a measurement of the maximum extracted power
available (Exp. H-14), which turned out to be about
2.5 Mw. On the afternoon of November 12 the
building was opened to visitors with proper clear-
ance, and the behavior of the reactor was demon-
strated to them. At 2004, an estimated total oper-
ating time of 100 Mw-hr having been accumulated,
the reactor was scrammed and the ARE experiment
was terminated.

APPROACH TO POWER

The initial phase of high-power operation of the
ARE began at 1445 on November 8. As shown in
Fig. 6.1, the power was raised in a series of
successive steps. At each power level, when
steady-state conditions had been obtained, a
complete set of control room data was taken. The
first experiment, H-1, was a 20-kw run. During
this run the safety chambers were withdrawn and
reset to scram at about 260 kw., At 1619 the
reactor was shut down because of the presence of

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CHRONGLOGY OF THE HIGH-POWER EXPERIMENTS

e

TIME OF DAY

 

 

NOVEMBER 8, {954

 

 

 

 

NOVEMBER 9, 1954

 

NOVEMBER 10, 1954

   
    
    
  
 
 
   
   
   
 
  
 
 
  

NOVEMBER 14, 1954

NOVEMBER 12, 1954

    
   
  
 

 

 

0000 ] REACTOR SUBCRITICAL
EXP. H-6: START UP BY USING TEMPERATURE
COEFFICIENT; POWER TO 200 kw
END OF RUN
REACTOR MEAN TEMPERATURE
L | OFF-GAS SYSTEM COMPLETED AND WORKING; ACCIDENTAL SCRAM DURING CALIBRATION DURING RUN 1365°F
RADIATION LEVEL IN BASEMENT BACK TO OF A FLOWMETER REACTOR OUTLET TEMPERATURE 1522°F
NORMAL REACTOR AT 200 kw REACTOR INLET TEMPERATURE 1209°F
0300 |— - REACTOR AT 313°F
FUEL FLOW RATE 46 gpm
REACTOR OPERATED AT POWER LEVELS FROM gp
200 kw TO 1.5 Mw TO FAMILIARIZE OWL- 2oDIum JEAN TEMPERATURE ek
SHIFT CREW WITH OPERATIONS T A ow RATE 152 gpm
REACTOR AT 200 kw
REACTOR OPERATED AT POWER LEVELS FROM
100 TO 500 kw TO FAMILIARIZE DAY SHIFT
0600 +— [ WITH OPERATION
—t— JET EJECTOR TESTED TO GET CORRELATION REACTOR AT 100 kw
iECTTV{SIET'\{,P” PRESSURE AND BASEMENT EXP. H-7: MEASUREMENT OF SODIUM
TEMPERATURE COEFFICIENT; PEAK POWER
300 kw END OF EXP. H-13
EXP. H-8: EFFECT OF A DOLLAR OF DEMON
REACTIVITY; REACTOR AT 100 kw v ONSTRATION RUN
0900 — REACTOR AT 2 Mw
REACTOR AT T Mw; REGULATING ROD
TOTALLY INSERTED - DEMONSTRATION RUN
. REACTOR AT 2 Mw; EXP. H~14, RUN 1
20-kw RUN TO DETERMINE EFFECTIVENESS OF ROD WITHDRAWN FULL TRAVEL OF 12 in. \ s ’
TMPROVISED OFF-GAS SYSTEM DURING ROD INSERTED . BSERVATION OF REACTOR KINETICS;
OPERATION— EXP., H-2 ED FULL TRAVEL OF 12 in. REACTOR AT VARIOUS POWERS FROM
END OF EXP. H-8 | 100 kw TO 2 Mw
START OF EXP. H-3; APPROACH TO POWER ~ N
1200 —— END OF LOW-POWER EXPERIMENTS 200-kw RUN ACCIDENTAL SCRAM WHILE PULLING CHARTS ) MRAUXI\:N;UM POWER OF 2,5 Mw; EXP, H-14,
FUEL AND SODIUM HEAT BARRIER DOORS REACTOR AT 300 kw
RAISED
REACTOR OPERATED AT VARIOUS POWER
HELIUM BLOWER IN FUEL SYSTEM STARTED; LEVELS FROM 100 kw TO 2.5 Mw NOTICED SLIGHT FLUCTUATION ON ALL
SPEED INCREASED TO 300 rpm NUCLEAR AND PROCESS INSTRUMENTS
REACTOR SCRAMMED AT 260 kw BY SAFETY E?:A;\PHEI{?TE\?S:SCU;::AFEI?ITESF gg:g;fgg
CHAMBERS T
START OF HIGH-POWER EXPERIMENTS OPERATED AT POWER LEVELS FROM CONTINUOUS DEMONSTRATION OF THE
1500 REACTOR AT POWER 100 kw to 2.5 Mw REACTOR FOR VISITORS; OBSERVATION OF
|/ BEMAVIOR OF REACTOR UNDER POWER
20~kw RUN; APPROACH TO POWER; EXP. H-1 SCRAMMED AGATN BY SAFETY CHAMBERS 0 EXP. H-10: MEASUREMENT OF MODERATOR CYCLING; REACTOR CYCLED 21 TIMES
REACTOR CRITICAL L TEMPERATURE COEFFICIENT, REACTOR ig?{“‘PSOUVez;AN”“” LOW POWER TO
- OPERATED AT 800 kw: RUN 1
REACTOR SCRAMMED; ACTIVE GASES IN REACTOR AT 100 kw ON SERVO; SAFETY o
ASEMENT
8 CHAMBERS RAISED s DEMONSTRATION OF REACTOR TO AIR FORCE
500 kw REACHED; DATA TAKEN PERSONNEL; REACTOR AT 2 Mw
1 EACHED; DATA TAKEN
1800 |— Mw REA EXP. H-10; RUN 2; REACTOR AT 2 Mw
SODIUM SYSTEM HELIUM BLOWERS TURMED END OF FULL-POWER XENON RUN; EXP. H=11
;i S 00
| | CAUSE OF GAS LEAK THOUGHT TO BE ON; SET AT 500 rpm
DEFECTIVE SEALS IN JUNCTION PANEL DATA TAKEN START EXP, H-11: 2,12-Mw RUN FOR 25 hr 7O — DEMOMNSTRATION FOR AIR FORCE PERSONNEL
DETERMINE EFFECT OF XENON POISONING N
REACTOR AT 2.1 Mw L\ EXP. H-14; RUN 3; MAXIMUM POWER, 2.5 Mw
2 Mw REACHED; DATA TAKEN
. FINAL SCRAM AT 2004
AIR JET EJECTOR AND COMPRESSORS BEING )
2100 INSTALLED WEST OF BUILDING AND END OF RUN; REACTOR AT 200 kw ~ S;’?é‘l};PO“ZGEXSPéDI—:;JIN;I:FEFOFWECA'!’N%NFSELA(.;];,(.;RE?AF
CONNECTED TO TERTIARY OFF-GAS SYSTEM START EXP. Hed: ME T
TO MAINTAIN A NEGATIVE PRESSURE 1N e eraTunE ot T REMENT OF FUEL HELIUM BLOWERS; REACTOR AT 10 kw TO 2 END OF ARE OPERATION
THE PITS; 10600 £t OF 2-in. PIPE CONNECTED Mw e A e
TO JET EXHAUST AND RUN OVER HILL INTO END OF EXP. H-4
VALLEY SOUTH OF BUILDING END OF EXP. H-12
- START EXP, H-5: MEASUREMENT OF OVER-ALL
F—t— H,O0 MANOMETER INSTALLED IN BASEMENT REACTOR TEMPERATURE COEFFICIENT START OF EXP, H-13: ONE-TENTH POWER
TO MEASURE NEGATIVE PIT PRESSURE XENCON RUN; REACTOR AT 200 kw
2400 END OF EXP. H-§

 

T

 

Fig. é.1.

 

59

 

 

g
(LN ] N = e T

p T TR TR Ty, R T
B,

 

=

B

 

T

 

T TR OIS U ey
e s e e

Bk . -

L A

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airborne radioactivity in the basement, Apparently
the gas fittings to the main fuel pump were leaking
and fission-product gases and vapors were leaking
into the pits and from the pits into the basement
through defective seals in some of the piping and
electrical junction panels connecting the basement
with the pits.

In order to prevent the leakage of activity into
the basement it was decided to operate the pits at
a slightly subatmospheric pressure. Accordingly,
a 2-in, pipeline was run in a direction south of the
ARE building for a distance of 1000 ft and termi-
nated in anuninhabited valley, Then, by means of
portable compressors and a jet, the pit pressure
was fowered by about 6 in. H,0O, with the exhaust
from the compressors being bled into the 2-in.
pipeline, The pits were maintained at a negative
pressure for the balance of the experiment.

At 1125 on November 9 the reactor was again
brought to 20-kw power (Exp. H-2) for the purpose
of testing the new off-gas system to determine
whether or not the negative pressure of 6 in. H,0
in the pits was enough to keep the hot gases from
diffusing into the basement. During this run the
pit activity, as recorded on the pit monitrons, was
watched, and its decay after shutdown of the
reactor at 1246 was observed. From this run it was
ascertained that the subatmospheric pit pressure
would prevent pit activity from leaking into the
building.

At 1520 on November 9 the approach to power
was started again (Exp. H-3). At 1523 the reactor
was brought to a nominal power of 200 kw on a
17-sec period and leveled out. Six minutes later
the barrier doors were raised and the first at-
tempt to extract power was made. At 1529 the fuel
helium blower was turned on and the blower speed
was increased to 300 rpm. The nuclear power
again began to increase, but the safety chambers
had not been withdrawn far enough and they
scrammed the reactor when the power level had

reached about 260 kw.

After another false start, the safety chambers
were withdrawn again and set to scram at about 4
Mw. The reactor was then brought to 100 kw,
leveled out, and placed on flux servo at 1625, At
1626 it was taken off servo and the blower speed
was increased, this time to only 170 rpm. The
reactor power sfowly rose and at 1643 had leveled
off on the reactor temperature coefficient at about
500 kw. After a set of data had been taken and

the reactor mean temperature elevated, the blower

speed was increased to 400 rpm and the power level
continued to rise. By 1715 the power level had
feveled out at near 1 Mw.

The sodium blowers were turned on (at 1740) and
brought to a speed of about 500 rpm. Half an hour
later the fuel blower speed was increased to 1590
rpm, and in 10 min a steady-state power level of
about 2 Mw was reached, At this time the nuclear
and process instruments read as follows:

Log N power,] Mw 1.32
Micromicroammeter power,] Mw 1.38
Reactor inlet fuel temperuture,2 °F 1203
Reactor outlet fuel tempercnm'e,2 °F 1440
Reactor mean fuel temperofure,2 °F 1321

Fuel temperature difference across recu:'ror,2 °F 237
Fuel flow, gpm 46
Reactor sodium mean temperature, °F 1281
Sodium temperature difference across reactor, °F 10

Sodium flow, gpm 152

From these data the extracted power was calculated
as follows:

Power extracted from fuel, Mw 1.16
Power extracted from sodium, Mw 0.05
Power extracted from rod cooling, Mw 0.02
Total extracted powar,2 Mw 1.23

However, it was subsequently discovered that the
temperatures upon which this extracted power was
calculated were in error and, as discussed in
Appendix K, the actual temperature gradient at this
time was about 40% higher. Therefore, the ex-
tracted power was correspondingly higher than that
listed. A general discussion of the extracted power
appears in a following section.

The above-described power operation was main-
tained for about 1 hr, during which time no further
trouble was occasioned by the fission gases.
Power was reduced at 1900 and the reactor scrammed
at 1715, and thus the initial phase of the high-
power operation was concluded,

TEMPERATURE COEFFICIENT MEASUREMENTS

Several experiments were performed in an at-
tempt to measure the effect on reactivity of temper-
ature changes of the various components of the

 

1Set on the basis of the power calibration from fuel
activation.

2The temperatures and consequently the extracted
power values are in error as noted in the discussion
following the data.

61

 

- e g

 
i b1

 

 

 

 

 

reactor, In addition to the over-all reactor co-
efficient, experiments were run to obtain data on
the instantaneous fuel temperature coefficient and
the coefficients attributable to the sodium and the
beryllium oxide moderator. There were several
anomalies that developed in the course of these
experiments, the two most significant being (1) the
interpretation of the lag in the response between
the rod movement and the change in reactor temper-
atures as power was abstracted from the reactor
and (2} the discrepancy between the thermocouple
indications at the reactor (fuel and sodium inlet
and outlet temperatures) and those along the lines
to and from the reactor. These anomalies are
discussed in detail in Appendixes O and K,
respectively., [t was concluded that the time lag
was not real (i.e., the fuel temperature should show
changes as soon as the reactivity does) and that
the line temperature most accurately reflects the
actual stream temperat'ures.

In the determination of a temperature coefficient
by varying the power extracted from the reactor,
the recorded temperatures are subject to correction,
as described in Appendix K. However, when the
temperature coefficient was determined, as in
Exp. H-5, by withdrawing the shim rods so that the
reactor slowly heated itself while the extracted
power remained constant, no correction needed to
be made for the temperature. Furthermore, this
technique minimized the temperature lag anomaly,
and the fuel and reactor temperature coefficients
derived from Exp. H-5 are therefore believed to be
the best experimental values for these coefficients.

Fuel and Reactor Temperature Coefficients

The fuel and reactor temperature coefficients
were both obtained from the same experiment, the
fuel being the instantaneous coefficient, the
reactor the ‘‘equilibrium’’ coefficient, The initial
attempts (Exp. H-4) to measure these coefficients
were not satisfactory because of the difficulty in
interpreting the results, and therefore the measure-
ment was repeated (Exp. H-5) in a somewhat
different manner,

In the first experiment, H-4, the procedure was
similar to that described previously under ‘‘Low-
Power Measurement of the Temperature Coefficient,’’
and the data are presented in Figs. 6.2 and 6.3,
The first of these figures is the recording of the
position of the regulating rod as it was inserted by
the servo as the fuel was cooled by the helium flow
in the heat exchanger, Figure 6.3 is a recording of

62

the mean fuel temperature with the rod displacement
as determined from Fig. 6.2 superimposed, De-
pending upon the time at which the slopes of the
two curves are compared, the temperature coefficient
appears to decrease from a value of ~—1.,6 x 1074
(AR/k)/°F to ~4.43 x 107° (Ak/k)/°F toward the
end of the run. These values represent the recorded
data and are of course subject to the correction for
temperature, as discussed in Appendix K, even as
the data are subject to various interpretations be-
cause of the lag in the temperature data (see
Appendix Q). Furthermore, the changes were made
quite rapidly, the entire run lasting only 5 min.

In an attempt to minimize the effect of the temper-
ature anomalies, the temperature coefficients were
then determined from an experiment (Exp. H-5) in
which the extracted power was held constant and
the shim rods were withdrawn. Starting with the
reactor at a mean temperature of 1260°F, the with-
drawal of the shim rods allowed the reactor to
gradually heat the whole system to 1315°F. Since
the reactor was a slave to the heat extraction
system, which was not changed, the extracted
power remained virtually constant during this time.
This process took about 35 min during which the
nuclear power was held constant at a nominal power
of 200 kw by the flux servo. The regulating rod
gradually withdrew to compensate for the decrease
in reactivity as the reactor temperature increased.
The data are tabulated in Table 6.1, and a plot of
rod withdrawal vs mean reactor temperature is
presented in Fig. 6.4. Inspection of Table 6.1
shows that the rate of change of the inlet and
outlet fuel temperatures followed the change in
mean fuel temperature very closely, and thus gave
assurance that the fuel throughout the reactor was
changing temperature at a uniform rate,

The most reliable values for the temperature
coefficients of the ARE were obtained from fthis
experiment. For the first 6 min the plot of rod
withdrawal vs reactor mean temperature shows a
slope of 0.296 in./°F. Since 1 in, of rod movement
is equivalent to 3.3 x 107* Ak/k, the initial
temperature coefficientwas —9.8 x 1073 (Ak/k)/°F.
After the first 6 min and up until conclusion of the
run, the curve in Fig. 6.4 demonstrates a constant
value of the temperature coefficient of —6.0 x 107>
(AR/E)/°F. These are the best values for these
coefficients that were obtained. It should be noted
that the value for the over-all temperature coefficient
agrees to within 25% with the value previously

obtained from the low-power experiments. However, -

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TIME CF DAY

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2127 ’ ( l

~=—[NSERTIO

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WITHORAWAL = OFF SERVO CONTROL
2126 \\‘
2125 fifi\\\

EXP, H-4. NOV, 9, 1954 \“w\\
UPPER END OF
N / ROD TRAVEL
ered — | ROD WITHDRAWN |
___J___———————“‘—"J BY INSERTING SHIMS
B e TO OVERRIDE SERVO
: \
h“"\~=‘ LOWER END OF
2123 < ROD TRAVEL
—
\\
2122 T
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\
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ONE INCH —ea— BLOWER ON {350 rpm)
REGULATING ROD

TRAVEL
2121
2120

12 f 10 9 8 7 6 5 4 3 2 1 0

REGULATING ROD INSERTED {in.)

Fig. 6.2. Regulating Rod Position During Measurement of Reactor Temperature Coefficient. Experi-

ment H-4,

ROD DISPLACEMEN® (in.}

2121

BL.LOWER ON

 

2122

ORNL - LR-DWG 6413
1400

EXP. H-4 NOV. 9, 1954

BLOWER
OFF

-—

1350

1

RO

1300

ME
AN TEMPERATURE

REACTOR MEAN TEMPERATURE (°F)

1250
2123 2124 2125 2126 2127 2128

TIME OF DAY

Fig. 6.3. Regulating Rod Position Supetimposed on Fuel Mean Temperature Chart, Experiment H-4.

63

* T B T T T YT PO e

R T T e

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TABLE 6,1, REACTOR TEMPERATURES AND ROD POSITIONS DURING
100-kw MEASUREMENT OF TEMPERATURE COEFFICIENTS (EXP. H-5)

 

 

 

 

El petmmeee o W*W L. A W e TE T E

 

 

Regulating Rod Rod Reactor Mean Reactor Inlet Reactor Outlet 5
Time Position Withdrawal Temperature Temperature Temperature
(in.) (in.) °F) (°F) (°F) .
i
2223 3.72 0 1263.5 1262.5 1283 . -
24 3.84 0.12 1265 1263.5 1285
25 4,15 0.43 1266 1264 1286 b
26 4.43 0.71 1267 1265.5 1287 L
27 4.69 0.97 1268 1266 1288 i
28 4.96 1.24 1269 1267.5 1290 ‘
29 5.23 1.51 1270.3 1269 1291 .
30 5,50 1.78 1271.5 1271 1292
31 5.77 2.05 1273 1272 1293 *
32 6.02 2.30 1274 1273 1294
33 6.28 2.56 1275 1274.5 1296
34 6.49 2.77 1276.5 1275.5 1297
35 6.68 2.96 1277.5 1277 1298
36 6.90 3.18 1278.5 1278 1300
37 7.10 3.38 1279.5 1279 1301
38 7.35 3.65 1281 1280.5 1302
39 7.50 3.78 1282 1281.5 1303
40 7.69 3.97 1283 1282.5 1304
41 7.92 4.20 1283.5 1283.5 1305
42 8.14 4.42 1285 1284.5 1307 |
43 8,32 4.60 1286.3 1285.5 1308 ¥
44 8.55 4.83 1287.5 1286.5 1309
45 8.75 5.93 1288.5 1287.5 1310
46 8.94 5.22 1289.5 1289 1311 :
47 9.11 5.39 1291 1290 1312
48 9.31 5.59 1292.5 1291 1313
49 9.50 5.78 1293.5 1292 1314
50 9.68 5.96 1294.3 1293 1315 L
51 9.85 6.13 1295 1294 1316 :
52 10.05 6.33 1296.5 1295 1317 g
53 10.23 6.51 1297.5 1296 1318
54 10.45 6.73 1298.8 1297 1319 ‘i"
55 10.60 6.88 1299.5 1298 ' 1320 E
56 10.78 7.06 1300.5 1299 1321 Efi |
57 10.99 7.27 1301.3 1300 1322
58 | 11.17 7.45 1302.5 1301 1323 E -
59 11.41 7.69 1303.5 1302.3 1324 .
2300 11.54 7.82 1305 1303 1325 3
01 11.80 8.08 1306.3 1304 1326 g
02 11.95 8.23 1307 1305 1327
03 12.20 8.48 1308 1306 1328 * f‘
04 12.38 8.66 1308.8 1307 1329 :
05 12.57 8.85 1309.8 1308 1330 k
06 12.75 9.03 1310.5 1309 1331 | s E
07 12,90 9.18 1311.5 1310 1332.5 5
08 13.19 9.47 1312.5 1311 1333 L
09 13.25 953 | 1313.5 i
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10 -
/
9 - >
OVER-ALL TEMPERATURE COEFFICIENT: (é?) - 2N ose3inseF :
B fover-aLL S0°F * :
8 Ak -5 @& *
(A—f) = ~0.183 x 0.00032 = —6.0 x 107~ Ak /°F
OVER-ALL
2
(& ]
O
@
2 6 =
}-——
< |
- |
] |
o /‘//‘/“ i i
L
o
§ '/
g 4 >
O
z /
= //
o
3 /
2
. Ad —9in. .
INITIAL TEMPERATURE COEFFICIENT : (—) = = —0.296in./°F
1 /A// B ma - 30-8°F
w
/ : -
//0 oL = % = —0.296 x 0.00032 = ~0.98 x 10°% £k/°F
®
7/
0 o | | |
1260 1265 1270 1275 1280 1285 1290 1295 1300 1305 1310 1315

7, REACTOR MEAN TEMPERATURE (°F)

Fig. 6.4. Regulating Rod Movement as a Function of Reactor Mean Temperature., Experiment H-5.

the instantaneous value of —9.8 x 1075 (Ak/R)/°F
is much more important from the reactor control
standpoint, and it was this large fuel temperature
coefficient which made the ARE demonstrate the
excellent stability described below under the
subtitle ‘'Reactor Kinetics.”

The value for the reactor temperature coefficient
was subsequently confirmed in a later experiment
(Exp. H-9) in which the system temperature was
changed by gradually cooling the sodium, which in
turn, cooled the fuel and the reactor. While the
curve obtained from this experiment was very
similar to that given in Fig. 6.4, the initial slope
could not be that due to the fuel temperature co-
efficients, because the fuel is one of the last con-
stituents in the reactor to feel the temperature
change. The “'equilibrium’’ slope, however, should
be that determined by the reactor temperature co-

efficient, and a value of —6.3 x 1073 (Ak/E)/°F

was obtained. No temperature correction was re-
quired because the extracted power in the fuel
system did not change (although that in the sodium
system did). This value agrees very well with
that obtained in the preceding experiment.

Sodium Temperature Coefficient

In order to measure the sodium temperature co-
efficient an experiment {Exp. H-7) was performed
in which the sodium was cooled and the correspond-
ing changes in sodium temperature and reactivity
were measured, [n this experiment it was mando-
tory that the sodium be cooled rapidly and the
data recorded before the moderator had time to
cool. Otherwise, since the sodium bathes the
moderator, the moderator temperature would follow
the sodium temperature and introduce an extraneous
effect due to its (i.e., moderator) temperature co-
efficient. Accordingly, with the reactor at about

65

e s s

T T

R i T P

 

 

 
 

 

 

33 kw and no power being extracted from either
the fuel or the sodium, the sodium blower speeds
(2 blowers) were quickly raised from 0 to 2000
rpm. The reactor period immediately started to
change and went from infinity to 50 sec in 1.5
min, which, as shown in Fig. 5.4, corresponds to
a reactivity change of 3.5 x 1074 Ak/k. The con-
sequent rate of reactivity change was 2.33 x 1074
(Ak/E)/min. In this experiment the regulating rod,
as well as the shim rods, was held in a fixed
position. During the 1.5-min interval, the fuel
temperature was changing at a rate of about
~1°F/min and the sodium temperature was changing
at a rate of —2.3°F/min. (The rate of change of
fuel mean temperature was corrected for the dis-
crepancy in the recorded mean temperature, as
described in App. K, but no comparable correction
was necessary in the rate of change of the sodium
temperature,) By applying the previously mentioned
value of —9.8 x 1072 (AE/E)/°F for the instantane-
ous fuel temperature coefficient to give a rate of
reactivity change for the fuel of 0.98 x 1074
(Ak/E)/min and subtracting from the rate of re-
activity change observed upon cooling the sodium,
it is found that the reactivity change caused by
the decrease in the sodium temperature is

2.33 x 1074 (Ak/E)/min (observed)
- 0.98 x 1074 (Ak/k)/min (fuel)

= 1.35 x 1074 (Ak/E)/min .
Then, by applying the observed rate of change of
the sodium temperature, the sodium temperature co-
efficient is found to be
1.35 x 1074 (Ak/k)/min
~2,.3°F/min

 

=~ -5.88 x 1073 (AR/E)°F .

This value is valid if it is assumed that the
transient took place rapidly enough that there was
no appreciable change in the moderator tempera-
ture. The measurements were, however, subject
to considerable error because the temperature
changes involved were so small as to be quite
difficult to detect and a correction was necessary
in the fuel temperature because of thermal lags.

Moderator Temperature Coefficient

| An. .éxpérim'enf'('H-'IO) was conducted in order to
determine the temperature coefficient of the moder-

66

ator. During this experiment the fuel temperature
was held constant and the speed of the blowers
for cooling the sodium was increased to change
the temperature of the moderator coolant. Since
this was done very slowly, the moderator did cool
down, in contrast to the earlier experiment (H-7)
in which the sodium temperature was changed so
rapidly that the moderator temperature was unable
to follow the sodium temperature. The earlier
experiment gave information as to the temperature
coefficient of the sodium alone, whereas experi-
ment H-7 gave the combined effect of sodium and
moderator changes. The reactivity was indicated
by the position to which the regulating rod was
adjusted by the flux servo.

The changes in the sodium and the moderator
temperatures slightly increased the heat loss of
the fuel, and the power had to be increased slightly
to keep the fuel mean temperature constant. |In
fact, in the middle of the run the reactor mean
temperature was lower than at the beginning, but
before the final reading was taken, the reactor
mean temperature was brought back to its original
valve. Table 6,2 gives the pertinent data, The
sodium inlet temperature was taken from a record-
ing of the temperature of the reflector coolant inlet
4 in. from the bottom of the reactor and the sodium
outlet temperature was taken from a recording of the
temperature of the reflector coolant outiet 3 in. from
the top of the reactor. The time recorded in column
one is the time when a reading was taken. The
change of the blower speed preceded this time by
a few minutes to allow the temperature equilibrium
to be established. As an indication of the estab-
lishment of the equilibrium, use was made of the
leveling off of the trace on the sodium temperature
differential recorder.

As can be seen from the table, the decrease of
the average sodium temperature from 1273 to
1246°F corresponds to a withdrawal of the regu-
lating rod from the 8-in. position to the 8,9-in. posi-
tion. At the lower temperature the reactor was less
reactive and showed that the temperature coefficient
was positive. The magnitude of this temperature
coefficient was small, (0.9 in./27°F) x 3.33 x 1074
(Ak/EY/ine = +1.1 x 1073 (Ak/R)°F. This co-
efficient is, however, the sum of the sodium temper-
ature coefficient and the moderator temperature
coefficient. Since the sodium temperature coefficient
was ~5.9 x 1073, the moderator temperature co-
efficient must be +6.9 x 1073 (Ak/E)/°F. This

value is, of course, subject to all the inherent

<

 

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TABLE 6.2 MODERATOR TEMPERATURE COEFFICIENT DATA

 

 

 

 

Fuel Fuel Blower Sodium Regulating
Speed Temperature Nuclear
) Mean Temperature P P Rod
Time ) (rpm) ©F) o Power
Temperature Gradient rem Position (Mw)
' w
(°F) (°F) No. 1 No, 2 Inlet Outlet Average (in.)
1735 1313 206 960 1050 1265 1282 1273 8.0 1.98
1745 1314 206 1160 1050 1255 . 1280 1267 8.0 1.98
1752 1313 203 1170 1260 1248 1278 1263 7.6 1.98
1802 1308 200 1340 1250 1238 1272 1255 6.6 1.98
1809 Increased servo demand signal so as to withdraw rod
1812 1311 204 1480 1240 1230 1268 1249 8.4 2.12
1825 1313 205 1470 1480 1225 1268 1246 8.9 2.12

 

errors of the measurement of the sodium temperature
coefficient, as well as the additional errors in the
temperatures recorded for this particular experiment.

MEASUREMENT OF THE XENON POISONING

At 1825 on November 10 a 25-hr run at a power
of 2.12 Mw was started for the purpose of measuring
the amount of xenon built up in the fuel (Exp. H-11).
As discussed in Appendix P, the reactor should
have been poisoned by about 2 x 10~ Ak/k after
25 hr, if it is assumed that no xenon escaped from
the molten fuel. Not only did the 25-hr run demon-
strate that very little of the fission-product gas

 

remained in the fuel but, also, that the reactor
possessed phenomenal stability. Except for a
minute withdrawal of the regulating rod to compen-
sate for a barely detectable drop in the mean re-
actor temperature, all readings on both reactor and
process instrumentation held constant within ex-
perimental error for the 25 hr. It was assumed that
the rod withdrawal was due to xenon buildup in the
fuel. However, it could have been due to any of a
number of minor perturbations, and therefore the
experiment demonstrated an absolute upper limit
on the xenon poisoning.

An abstract of the log book and data sheets
during the experiment follows:

November 10, 1954

1825 Run started; reactor power, 2.12 Mw; rod position, 9.00 in.; reactor mean temperature, 1311°F

November 11, 1954

0635 Reactor mean terfiperature had décreased to 1309°F; rod withdrawn to 9,05 in.

0750 Reactor mean temperature up to 1310°F; rod withdrawn to 9,15 in.

1020 Reactor mean temperature 1310.5°F; rod withdrawn to 9.25 in.

1120 Reactor mean temperature 1311°F

1435  Reactor mean temperature down to 1310°F; rod withdrawn to 9.30 in.

1559 Reactor mean temperature up to 1311°F

1932 Reactor mean temperature up to 1312°F; rod still at 9.30 in.; end of run

/

67

g

 
 

 

{t is to be noted that the temperature recorded
was actually one degree higher at the end of the
experiment than at the start; this indicates that
the withdrawal of the regulating rod by 0.3 in. may,
indeed, have been too much. The withdrawal was
unquestionably an upper limit on the compensation
needed for xenon poisoning and corresponded to a
Ak/k of 1x 1074, This was ), or 5% of the value
to be expected if the xenon had not left the fuel
{see Appendix P). Removal of the xenon probably
occurred by means of the swirling action of the
fuel as it went through the pump. As mentioned
previously, fission-product gases that probably
came from a leak in the gas fittings were detected
in the pits early in the experiment.

At the conclusion of Exp. H-11 at 1935, the re-
actor was operated at various power levels for 2 hr
in an experiment (Exp. H-12) for determining the
effect of the sodium flow rate on the extracted
power. At 2237 the reactor was set at 200 kw (one-
tenth full power) and held there by the flux servo
for 'IO hr (Exp. H-13). If any appreciable amount
of xenon had been built up in the fuel from decay
of iodine formed during Exp. H-11, the poisoning
effect should have been observed by a compensating
rod withdrawal during this 10-hr period. No appre-
ciable rod withdrawal was observed, and therefore
it was concluded that if there was xenon poisoning
from Exp. H-11, it was negligible.

POWER DETERMINATION FROM
HEAT EXTRACTION

The most reliable method for determination of
actual reactor power was that based on the energy
removed from the reactor in the form of heat. The
inlet, outlet, and mean temperatures, the tempera-
ture difference of the fuel across the reactor, and
the temperature difference of the reflector coolant
were continually recorded. The rates of flow of
both the fuel and the sodium were recorded and
could also be determined from the speeds of the
pumps. Since the heat capacities of both the fuel
and the sodium were known, the power level of the
reactor could be determined by the sum of the
values obtained from the following relations:

P, =011q,AT,

Py, = 0.0343 4, AT,

N
“where
P_ = power from fuel heat extraction (Mw),
PNa = power from sodium heat extraction (Mw),

68

volume flow rate (gpm),

q

AT = temperature gradient across reactor of
the fuel or the sodium (°F).

Since both the fuel and the sodium were cooled
by helium passing across a liguid-to-helium heat
exchanger and the helium was cooled by passing
it over a helium-to-water heat exchanger, the re-
actor power could also be determined from the
water Hlow and temperature differences across the
helium-to-water heat exchangers. Since the fuel-
to-helium and sodium-to-helium heat exchangers
are close to their respective helium-to-water heat
exchangers, little heat was lost by the helium and
therefore practically all of the heat removed from
the fuel and sodium was transferred to the water.
The heat balance thus obtained was known as the
secondary heat balance, while that obtained directly
from the fuel and sodium systems was known as
the primary heat balance.

initial comparisons of the primary and secondary
heat balances revealed discrepancies of the order
of 50%. Subsequent investigation revealed that
the temperature drops of the primary heat balance
were in error. 1he temperatures were obtained from
a few thermocouples located on fuel and sodium
lines within the reactor thermal shield that read
considerably lower than several thermocouples
located external to the thermal shield on the fuel
and sodium lines to and from the reactor. This
anomaly is discussed further in Appendix K. All
reactor inlet and exit femperatures were subse-
quently based on the external thermocouple readings.

A detailed analysis of the extracted power was
made (App. L) for the 25-hr xenon experiment
primarily because of the certainty that equilibrium
conditions had been established; also, this high-
power run at more than 2 Mw contributed almost
two-thirds of the megawatt-hours logged during the
entire reactor operation. For this run the primary
heat balance showed an extracted power of 2.12
Mw, and the secondary heat balance gave an ex-
tracted power of 2.28 Mw. The difference between
the two determinations was 7%. Furthermore, for
this run about 75% of the power was removed by
the fuel, about 24% by the sodium, and about 1%
by the rod cooling system.

Heat balances and, hence, power determinations
were made for a number of other experiments during
the high-power operation; these are tabulated in

Appendix L. One such heat balance that was of

particular interest was obtained during the maxi-

 

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mum power run (Exp. H-14). For this run the mean
temperature was raised to 1340°F (to avoid a low-
temperature reverse), and the fuel and sodium
system blower speeds were increased to their
respective maximums, At equilibrium the power
levels indicated by the primary and secondary heat
balances were 2.45 and 2.53 Mw, respectively.
The line temperatures throughout the fuel and
sodium systems during this experiment are shown
in Figs. 6.5 and 6.6, respectively, The tempera-
tures indicated along the lines are the actual
thermocouple readings of each point at equilibrium.
It may be noted that the outlet fuel line tempera-
ture averaged about 1580°F. The temperature
measured at the reactor was considerably lower,
as shown in Fig. 6.7, which shows a portion of the
instrument panel during this experiment.

REACTOR KINETICS

" A distinctive control characteristic of any circu-
lating-fuel reactor is that the reactor power is
determined solely by that part of the system which
is external to the reactor, i.e., the heat extraction
equipment. In the power regime, control rods do
not appreciably influence the steady-state power
production, but the power extracted from the reactor
does influence the reactivity of the reactor and
hence renders the reactor controllable.

In order that such a system be acceptable as a
power reactor, it is requisite that the reactor have
a negative temperature coefficient of reactivity.
One of the most gratifying results of the ARE
operation was the successful demonstration of its
large negative temperature coefficient.  This
temperature coefficient made it possible for the
reactor to maintain a balance between the power
extracted from the circulating fuel and coolant and
the power generated within the reactor. The tem-
perature cycling of the fuel was the mechanism by
which equilibrium was maintained. ‘

A thorough understanding of control processes
in a circulating-fuel reactor with a negative tem-
perature coefficient is necessary for an appreciation
of the kinetic behavior of such a power reactor in
the power regime. An important purpose of the
ARE was the observation of the kinetic behavior
of the reactor under power coupling to its load
when perturbations in the reactivity were intro-
duced. Transient conditions could be induced both
by control rod motion and by variation of the
external power load. Information on the kinetic
behavior was obtagined from a number of experiments

that were conducted during the period of power
operation. These are described in the following
sections., Preceding the discussion of these experi-
ments is a detailed qualitative description of the
temperature cycling of the circulating fuel which
is inherent to all kinetic phenomena.

Reactor Control by Temperature Coefficient

The control of a circulating-fuel reactor with a
negative temperature coefficient can best be under-
stood by following the course of a *‘slug’’ of fuel
as it traverses the ARE system. For a description
of the course of a slug, it is assumed that at time
zero the system is in equilibrium, with isothermal
temperatures throughout; in this condition no power
is being extracted. At time t the fuel system
helivm blower is turned on. It is also assumed
that at this time the slug of fuel under observation
is just entering the heat exchanger. In passing
through the heat exchanger the slug is cooled by
the helium blowing through the heat exchanger.
About 20 sec later the cooled slug enters the re-
actor and is registered as a decrease on the inlet
temperature indicator. Because it is cooler than
the fuel it is displacing, its density is greater and
therefore the number of uranium atoms per unit
volume is greater. This results in a greater fission
rate and, hence, a greater reactivity which, in turn,
increases the power generated. The temperature
of the slug rises as it passes through the reactor
because of increased power generation. The rise
in temperature of the slug results in its expansion
and decrease in reactivity. This, in turn, lowers
the rate of power generation. Eight seconds after
the slug enters the reactor it passes out into the
outlet line and is registered as an increase in
temperature on the outlet fuel temperature indi-
cator. In 47 sec from the start of its journey it is
back at the heat exchanger to be cooled again.

The masses of fuel behind the initial slug follow
the same pattern so that, since the fuel is a con-
tinuvous medium, the power generated in the re-
actor® will rise until the increased reactivity due
to the incoming fuel and decreased reactivity due
to the outgoing fuel attain o balance. At this
point equilibrium is reached between power ex-
tracted by the helium and the power generated in

 

3With a power reactor it is appropriate to speak of two
types of power: the nuclear power, i.e., the total power
generated within the reactor; and the extracted or useful
power, i.e., the power removed from the fuel and coolant
by cooling.

69

 

 

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REACTOR —_
T . KEY:
. | T 100 SERIES NUMBERS ARE
i FUEL LINES.
| ALL FOUR DIGIT NUMBERS ARE
| 1563 TEMPERATURES IN °F
FREEZE VALVE
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Fig. 6.5. Fuel System Temperatures During Maximum Power Run. Experiment H-14.

 

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Fig. 6.7. Portion of Instrument Panel During Maximum Power Run. Experiment H-14.

the reactor, The reactor power can thus be varied
at will merely by changing the rate of cooling, that
is, the demand for power. This type of reactor is
said to be a slave to the demand.

The nuclear power is proportional to the neutron
flux for any given critical concentration. The
neutron flux is conventionally measured by o
neutron detector, such as a Log N chamber system,
which consists essentially of a neutron detector
coupled to an electronic system which has a
logarithmic-type of output signal. This signal is
proportional to the neutron flux (and, hence, the
power) level, and is indicated and recorded on a
logarithmic scale.

The extracted power, on the other hand, is pro-
portional to the product of either the fuel flow and
the temperature difference between the inlet and
outlet sides of the heat exchanger (or reactor) or
the secondary (or tertiary) coolant flow and its
temperature difference. In the ARE the fuel was
cooled by helium flowing across it in the fuel-to-

72

helium heat exchangers. The heat picked up by
the helium was then removed by water in a helium-
to-water heat exchanger. It was possible to measure
the extracted power in both the fuel loop and the
water loop, as explained in a preceding section.

When no power is being extracted from a power
reactor, the nuclear power is determined by the
control rod and shim rod settings, and the reactor
is controlled by some type of servo mechanism
which keeps the power at some desired level.
When the reactor is taken off the servo and allowed
to respond to the demand of the cooling produced
in the heat exchanger, it will seek a power level
determined by the rate of heat removal. Once an
equilibrium between nuclear and extracted power
has been established, changing the shim rod or
regulating rod settings merely changes the nuclear
power level but does not alter the amount of power
being extracted. This change in nuclear power
results in a raising or lowering of the reactor mean
temperature, and, once the temperature change has

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been effected and equilibrium re-established, the
nuclear power and the extracted power will again
be equal.4 _

The control of the ARE by extracted power demand
is illustrated in Fig. 6.8, which shows the tracings
made by several of the control room instrument
recorders during power operation between 1258 and
1329 on November 10, 1954. The reactor behavior
is demonstrated by reading the tracings right to
left, proceeding as follows: At 1258 the reactor
was in an equilibrium condition at about a 50-kw
power level. At 1259 the fuel system helium blower
was turned on and its speed was increased to 500
rpm. The reactor power, as noted on the Log N
chart, rose on a 10-sec period to slightly over 1
Mw, ‘‘overshot’’ its mark, fluctuated somewhat in
the manner of a damped oscillator, and, finally,
some 4 min later, came to an equilibrium power of
around 900 kw. Soon,s but not immediately, after
the blower was turned on, the reactor fuel inlet
temperature began to drop and the fuel outlet tem-
perature began to rise, while the reactor mean
temperature began to drop slightly. Each of these
temperatures showed the oscillatory phenomenon
noted with the Log N recorder. The reason for the
drop in the mean temperature was that both the
decrease and rate of decrease of the inlet tempera-
ture were greater than the corresponding increase
and rate of increase of the outlet temperature. The
inlet temperature dropped from 1335 to 1256°F% at
a rate of 1.54°F/sec and the outiet temperature
rose from 1405 to 1475°F at a rate of 1.36°F/sec;
as a result the mean temperature fell fram 1370 to
1365°F at a rate of about 0.1°F/sec. The fact that
the mean temperature dropped {and this was a
characteristic phenomenon noted throughout the
power operation) rather than remaining constant
can be explained, at least in part, by the distortion
of the flux patterns within the reactor. The highest
fluxes were toward the inlet fuel passages, and
the lowest fluxes were toward the outlet fuel
passages {(cf., App. O},

At 1310 the blower speed was increased from 500

 

4On several occasions the nuclear power, as indicated
on the Log N recorder, rose temporarily above 2.5 Mw,
but the extracted power, limited by the system capacity
for removing heat, never exceeded 2.5 Mw.

Errors in marking the charts could conceivably ac-
count for as much as 1/2 min of the time ditferences
noted. This phenomenon of time lag is discussed later
in this section, ‘

8T he temperature readings given here and elsewhere
in this section have been corrected by the method given
in Appendix K,

to 1000 rpm, and the power rose from 1 to 1.8 Mw
on a period of 7.5 sec, with corresponding changes
of the inlet, outlet, and mean temperatures. Al-
though the power increase in this case was compa-
rable to the first power increase, the rates of rise
of both the power and temperctures were much
less, being on the order of one-half as much for
the temperatures.

The effect of withdrawing the regulating rod 3 in.
is demonstrated with the next rise at 1312 in
Fig. 6.8. Up to this time the trace of the regu-
lating rod position was constant, since it was not
on servo. With the withdrawal of the rod the rise
of the nuclear power was immediate. After some
delay, all of the temperatures showed a rise.
Interestingly enough, the outlet temperature rose
16°F, the mean 12°F, and the inlet 9°F. This
pattern, with the outlet temperature showing the
greatest change, was characteristic of the temper-
ature changes incurred by shim rod and regulating
rod movement during power operation. After the
rod movement the temperatures leveled out at the
higher values, but the nuclear power level slowly
drifted back to the equilibrium position,

At 1321 the blower was shut off and the charts
show the result of the shift toward isothermal (no
power) conditions. Two minutes later, when the
nuclear power had decreased to about 600 kw, the
blower was again started and its speed was in-
creased to 1500 rpm. The nuclear power level rose
initially to 2.6 Mw, and it again went through an
oscillatory cycle before settling down to 2.2 Mw.
The corresponding temperature oscillations were
again noted. These oscillations were of sufficient
strength to determine an oscillation period of about
2.25 min. An interesting observation at this point
was that the time lag of temperature response for a
higher initial power (600 kw compared with 50 kw)
was only about one-half the lag during the first
rise to power at 1259. This phenomenon had been
noted previously in connection with the various
temperature coefficient experiments (cf., App. O).

Table 6.3 shows the data obtained from the
nuclear and process instruments during the typical
operation period shown in Fig. 6.8.

Startup on Demand for Power (Exp. H-6)

To illustrate how completely the reactor was a
slave to the load, an experiment was performed in
which the reactor was brought to subcritical and
then taken to critical on temperature coefficient
(i.e., by the power demand and without use of

73

 

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FUEL QUTLET
TEMPERATURE (°F) TEMPERATURE (°F)

FUEL MEAN

FUEL INLET
TEMPERATURE (°F)

REACTOR

POWER {Mw)

REGULATING
ROD POSITION (in.)

 

 

1300

1350

1300

1300

1250

1200

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mo- N

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S

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1328

BLOWER SPEED
REDUCED TO LEVEL OUT

POWER AT 2 Mw

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BLOWER STARTED; SPEED
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REACTOR FUEL OUTLET TEMPERATURE -~

REACTOR FUEL MEAN TEMPERA

REACTOR FUEL INLET TEMPERATURE -

LOG & POWER RECORDER

    

1327 1325

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1324 1323 {322 1321 314 1343 4342 1341 1310 4309 1304
TIME OF DAY, NOV. 10, 1954

Fig. 6.8. Typical Reactor Behavior During Power Operation.

FROOCUES M g oW o pWECUTT TN YT v pTR TR T W ITNT R ST TrROmRYT

1303

1302

n

1301 1300 1259

4

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TABLE 6.3. NUCLEAR AND PROCESS DATA OBTAINED DURING
A TYPICAL OPERATION PERIOD (NOV. 10)

 

Operation Performed

 

 

Type of
Type of Data Fuel System  Fuel System Regulating Fuel System  Fuel System
Measurement Obtained* Blower Speed Blower Speed Rod Blower Speed Blower Speed
Increasing Increasing  Withdrawn  Decreasing Increasing
Time 1300 1310 1312 1321 1323
Log N power P, Mw 0.0566 0.99 1.58 1.56 0.57
Py Mw 1.07 1.70 1.94 0.57 2.17
5P, Mw 1.013 0.71 0.36 0.99 1.60
i, sec 28 41 9 60 81
OP/0t, Mw/ sec 0.0363 0.0173 0.04 0.016 0.198
T sec 10 75 33 63 54
Fuel outlet T, 1 °F 1405 1485 1527 1546 - 1491
femperature T, » °F 1474 1527 1543 1491 1566
8T, °F 69 42 16 —55 75
Ot, sec 45 56 24 75 58
8T /61, °F/sec 1.54 0.75 0.67 -0.73 1.29
Fuel inlet T; v °F 1335 1269 1231 1239 1314
temperatore T, o °F 1256 1227 1240 1314 1208
8T, °F ~79 ~42 9 75 -106
St, sec 58 53 42 75 88
6T /1, °F/sec -1.36 ~-0.79 0.21 1.00 ~1.20
Fuel mean T °F 1370 1377 1379 1388 1402
femperature T, o °F 1365 1377 1391 1402 1387
8T ., °F -5 0 12 14 -15
St, sec 52 33 75 73
5Tm/5z, °F/sec —~0.096 0 0.36 0.19 —-0.20
Reactor fuel Ar,, °F 70 216 296 287 177
AT AT,, °F 218 300 303 177 358
XAT), °F 148 84 7 -110 181
Ot, sec 52 54.5 33 75 73
S(AT)/Bt, °F/sec 2.87 1.54 0.21 ~1.47 2.49
Regulating rod dy, in. | 7.60
movement dy, in. 10.37

75

T T e T R £ e epegrm,

T T Y W

 

 
e o o

 

 

TABLE 6.3 (continued)

 

Operation Performed

 

 

 

Type of
Type of Data Fuel System  Fuel System Regulating Fuel System  Fuel System
Measurement Obtained* Blower Speed Blower Speed Rod Blower Speed Blower Speed
Increasing Increasing Withdrawn Decreasing Increasing
Regulating rod Sa’], in. 2,77
."’°"er_“e”* 01, sec 9
Od/At, in./sec 0.32
Ak/k, % 0.0914
(Ak/k)/St, %/sec 0.011
Fuel system helium §4e TPM 0 500 1000 0
blower speed s_, tpm 500 1000 0 1500
Os, rpm 500 500 -1000 1500
Ot, sec 14 14 60 42
Os/8t, rpm/sec 36 36 ~17 36
*The following symbeols are used:
Quantity Subseript
P power 1 initial cendition
t time 2 final condition
T period o outlet condition
T temperature i ‘inlet condition
d position m mean
£ multiplication factor
s speed
O change in quantity

A difference between two quantities

The 6's are the times required to go from the initial condition to the final condition at the greatest observed rate of
change. The temperatures quoted have all been corrected by the method described in Appendix K.

rods}). The experiment consisted of tworuns. In
the first run the reactor was taken directly from
100 kw to slightly over 2 Mw. In the second run
the reactor was taken to power from a subcritical
condition.

The progress of these experiments, as recorded
by the thermocouple recorders on the inlet and

outlet sides of the six individual fuel tubes,is

shown in Fig. 6.9. Reading from right to left, the
reactor was at 200 kw power at 2300 on November
9 at the beginning of the experiment. The fuel
system helium blower speed was increased to 1700
rpm slowly, starting at 2307. Ten minutes later
the sodium system coolant blowers were turned on.
As the power rose to about 2.5 Mw, a temperature
difference of some 320°F appeared across the re-
actor tubes. This is shown in Fig. 6.9 by the
parting of the inlet and outlet temperature indi-
cations of the reactor fuel tubes. This temperature

76

difference remained constant until the blower speed
was reduced at 2340, At 2345 the inlet and outlet
tube temperatures were nearly the same again at
200 kw power. A sharp drop in the tube tempera-
tures at this point corresponded to the insertion
of the shim rods.

 The sharp increase in the temperature differential
across the tubes at 2353 was the result of a steep
rise in power when the blower speed was increased
from 0 to 1700 rpm. At 0005 on November 10, the
regulating rod was first entirely inserted (from
7 in. withdrawn) and then fully withdrawn. This
motion was followed by a drop and then a sharp
rise in the fuel tube temperatures. This action
marked the end of Run 1.

During Run 2 the reactor was brought subcritical
by turning off the fuel system helium blower and
inserting the regulating rod, starting at 0016. The
sodium system helium blowers had been operating

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TIME OF DAY

Fig. 6.9. Behavior of Reactor Fue!l Tube Inlet and Qutlet Temperature During Experiment on Control by Temperature Coefficient,

 

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some of the time during Run 1 to extract power from
the sodium. As soon as the reactor became sub-
critical the operation of the sodium system blowers,
which were then removing about 400 kw of power
from the sodium, made the reactor go critical again
and stabalize at the 400 kw level within 5 min. Itis
interesting to note that because the mean sodium
temperature was some 65°F lower than the mean fuel
temperature, the fuel temperatures dropped 65°F
during this time from a mean temperature of 1395 to
1330°F, even though the nuclear power was in-
At 0026 the fuel system blower was
turned on again and the nuclear power temporarily
rose to 3 Mw.*

creasing.

Two minutes after the fuel system
blower was turned on, the fuel heat exchanger
outlet temperature went below 1150°F and the
auvtomatic heat-exchanger-temperature interlock re-
duced the blower speed until the heat exchanger
outlet (reactor inlet) temperature had risen above
1150°F. At 0030 the reactor power had leveled
out at 1 Mw. It had thus been successfully demon-
strated that the power demand would bring the
reactor critical; therefore, at 0033, the fuel and
sodium system blowers were turned off and the
experiment brought to an end.

Effect of One Dollar of Reactivity (Exp. H-8)

One of the objectives of the ARE was the obser-
vation of effects of introducing excess Ak into the
reactor during high-power operation. The intro-
duction of the excess Ak could be most easily
accomplished by movement of the regulating rod,
which was worth 0.4% excess k& (one dollar of
reactivity). One experiment consisted merely of
withdrawing the regulating rod from an entirely
inserted position, and recording or noting the
effects and, then, after equilibrium had been estab-
lished, inserting the rod to its original position.
Since the regulating rod had a travel rate of
0.32 in./sec, reactivity could be introduced at the
rate of 0.011% Ak/k-sec. Fig. 6.10 shows the
history of the experiment.

At 1110 on November 10, with the reactor at an
initial power of 2.2 Mw, the regulating rod was
withdrawn its full 12 in. of travel. The reactor
went on an observed 42-sec period’ until the
nuclear power was 3.9 Mw 35-sec later. The re-

 

"This period was measured from the Log N recorder
trace. The period recorded on the period meter was
partially a time rate of change of period since Ak/k was
not constant but was increasing during this time. A
discussion of this is given in Appendix S,

78

actor fuel temperatures rose an average of 45°F in
49 sec. Once the rod withdrawal had been com-
pleted, the fuel inlet and outlet temperatures
leveled off at new values. The extracted power
rose a few per cent when the rod was withdrawn,
mainly because, with the higher fuel mean tempera-
ture, the rate of cooling and, hence, the rate of
heat removal in the heat exchanger was greater.
Since the extracted power rose only slightly during
this time, the nuclear power slowly drifted down-
ward from its peak of 3.9 Mw to 2.9 Mw, as can be
seen in Fig. 6.10. Note that the mean temperature
continued its upward trend. Three minutes later
the rod was completely inserted.
power decreased from 2.9 to 1.4 Mw on a 47-sec’
period and then slowly drifted back to 2.2 Mw, the
starting power. The fuel temperatures then re-
turned to values slightly higher than their original

The nuciear

values. A characteristic behavior of the outlet
temperature was again noted during this experiment,
namely, that movement of a rod affected the outlet
temperature most and the inlet temperature the
least. Additional data on this experiment are given
in Table 6.4.

Again, with this experiment, a time lag was
noted; the inlet temperature lagged behind the out-
let temperature by about the fuel transit time of
47 sec. A reasonable explanation for this lag is
that the time required for the fuel in the reactor at
the start of the experiment to affect the inlet line
thermocouple is the transit time through the system.

Because of this difference in response of the
thermocouples on the inlet and outlet fuel lines
when the rod was withdrawn, the temperature
differential across the reactor rose from 341°F to
a peak value of 395°F in about 1 min and then
dropped to a more or less constant value of 355°F,
which was 14°F higher than at the start. This
means that the extracted power rose slightly, about
8%. Examination of the tracings of the fuel and
sodium heat exchanger cooling water recorders for
this time indicated that the fuel power extraction
went up 6% and the sodium power extraction roughly
2%. The rise in power extraction was due mainly
to the higher mean temperatures and thus the higher
cooling rates attained. The extracted power rose
to about 2.4 Mw at the new equilibrium. The dif-
ference (500 kw) between this value and 2.9 Mw,
as shown by the Log N recorder, went to heating
the reactor. This was evidenced not only by the
continuing increase in reactor mean fuel tempera-
ture but also by the rising travel in inlet and outlet

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REACTOR FUEL
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—=— TIME OF DAY, NOV, 10, 1954

Fig. 6,10, Kinetic Behavior of Reactor Upon Introduction of a Dollar of Reactivity.

79

T

 

R TSI ETORTET ) p R s

T R

T T ———

 

T

 
TABLE 6.4, NUCLEAR AND PROCESS DATA OBTAINED DURING EXPERIMENT H.8 FOR

 

el L

 

A o B L B BRI KL e moemo b v

 

 

DETERMINING THE EFFECT OF ONE DOLLAR OF REACTIVITY (NOV. 10)
_ Operation Performed

Type of Data !

Type of Measurement Obstained* Regulating Regulating

; Rod Withdrawn Rod Inserted o
| S
f Time 1110 1114
i ‘
] l.og N power Py Mw 2.34 2.92 . ';
Py Mw - 3.94 1.41 k

&
OP, Mw 1.60 1.51 ko

i

Ot, sec 35 35 ;

p/5t, Mw/ sec 0.046 0.043 E

7, sec 42 47 E

: Fuel outlet temperature To 1 °F 1573 1620 F
1 T ., °F 1623 1572 v

1 0,2 :
i-i 8T , °F 50 48 ;
o &

5{, sec 41 41

8T0/51, °F/sec 1,22 -1.17 g

Fuel inlet temperature T: 4 °F 1217 1258 :

T.., °F 1257 1221 « }

7,2 E

oT, °F 40 ~37

o, sec . 57 56 :

-}

5Tz./3t, OF/sec 0.70 —0.66 :

Fuel mean temperature T °F 1395 | 1439 E

’ ° 3»

Tm,2' F 1440 1396 E
8T, , °F 45 ~43 b

B, sec 49 49 i
8T /81, °F/sec 0.92 ~0.88 B

Reactor fuel AT AT], °F 356 362 ‘F

AT, °F 366 351 %

¥

XA, °F 10 1 §

8t, sec 49 49 %

SAT)/8t, °F/sec 0.20 ~0.21

Regulating rod movement dy, in. 2.0 14,0 * t

dyy in. 14.0 2.0 %

od, in. 12.0 -12.0 s

*See footnote to Table 6.3.

80

   
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TABLE 6,4 (continued)

 

Type of Data

Operation Performed

 

 

Type of Measurement Obtained* Regulating Regulating
Rod Withdrawn Rod Inserted
Regulating rod movement 01, sec 37 37
8d/ 81, in./ sec 0.32 ~0.32
Ak/k, % 0.4 0.4
(AR/R)/Ot, %/ sec 0.011 ~0.11
Fuel system helium blower speed Sy P 1750 1750
Sor rPM 1750 1750
Os, rpm 0 0
Ot, sec - ~
Os/ 61, rpm/ sec 0 0

*See footnote to Table 6.3.

fuel temperatures just before the rod was inserted

at 1114,

Power Cycling of Reactor

The last day of operation of the ARE, November
12, was devoted to maximum power runs and to
demonstrations of the reactor operation to visitors.
The complete history of the reactor operation during
that time is shown in Fig. 6.11. Chart A shows
the trace of the sodium temperature differential
across the reactor and chart B shows the trace of
the outlet temperature of the water from the fuel
heat exchangers. Clearly seen in chart A are the
three maximum power runs at 10:00 AM, 12:30 and
7:45 M. Both charts show the cycling of the
extracted power that the system underwent, but
the cycles are especially well defined in chart B,

which shows the heat exchanger water temperatures

varying from a low of 65°F to a high of 140°F,
corresponding to total extracted powers of from
around 100 kw to over 2 Mw. In the 12 hr between
8:00 AM and 8:00 PM, 24 complete cycles were
recorded. In the afternoon hours between 1:00 and
4:00 PM, when the most visitors were present, the
reactor was. cycled an average of 3 times per hour.
Of interest to the visitors were the tracings of the
reactor fuel tube AT recorders. The photograph
reproduced as Fig. 6.12, which was taken during
one of the cycles, shows each of the six fuel tube
AT recorders simultaneously tracing out the same
power cycle pattern. The picture was taken about

12:30 PM during one of the maximum power runs.
The recorders show AT’s of around 250°F. The
actual fuel tube differences at this time were more
nearly 355°F.

Additional information on the power cycling may
be obtained from Fig. 6.13 which shows the three
temperature cycles recorded in the hour between
1:10 and 2:10 PM on the last day. Shown in this
figure are traces from the six fuel tube AT re-
corders, a time condensation® of the micromicro-
ammeter recorder (which traced the power), one of
the heat exchanger outlet temperature recorders,
and the individual fuel tube temperature recorder.

Referring to the micromicroammeter trace and
reading from right to left starting at 1317, the
events are elaborated for the first cycle. The other
two cycles are similar.

At 1317 the fuel system helium blower was turned
off, und the reactor power was allowed to decrease
from a level of 1.8 Mw. Eight minutes later the
power had fallen to about 130 kw, and the blower
was turned on to 1700 rpm. The power rose quickly
to 2.3 Mw, at which point a low-heat-exchanger-
temperature blower reverse occurred; that is,
when the helium cooling lowered the heat exchanger
outlet fuel temperature to 1150°F (as noted above),
a relay was automatically actuated that decreased
the helium blower speed until such time as the

 

aThis condensation was necessary because the origi-
nal chert ran at a speed of 80 in./hr, while charts used
subsequently ran at a speed of only 4 or 5 in./hr.

81

R h ey o e e o

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e

 

 
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C T T SRR QP T Y P W B T e me ey T o e TR T w rr oy

 

 

 

FoopeEe e s P
Wk,

Ak, .

LGl

ki,

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e

oy e

4

4

 

 

 

 

Fig. 6.12, Reactor Fuel Tube AT’s During High-Power Operation,

outlet temperature rose to above 1150°F., During
this time the power fell to 1.1 Mw. At 1328 «
“permit'’ signal was given and the power was
again increased. When 2 Mw was reached the re-
actor was leveled off on extracted power and for
10 min the effect of regulating rod movement was
demonstrated to visitors.

The maximum nuclear power attained during the
first cycle was 2.8 Mw? at 1340. During the third
cycle at 1410 a nuclear power of 3.4 Mw was
reached. '

The six fuel tube AT recorders (Fig. 6.12) fol-
lowed exactly the same pattern as the micromicro-
ammeter recorder. |t should be noted, however,
that the actual AT’'s were of the order of 100
degrees higher than those shown on the charts.
At the high power peaks of the third cycle the fuel
outlet line temperatures, as read in the basement,
were indicated to be about 1625°F. The inlet

temperatures ranged from an actual (vs observed)

high of 1340°F to a low of 1150°F during the low-
heat-exchanger-temperature blower reverse,

The change in tube AT's as a result of rod move-
ment was due partly to the phenomenon of the time
lag between the inlet and outlet thermocouples.
Had there been no lag the effect of rod motion on
these cauples would have been small.

The fuel heat exchanger outlet temperatures
showed the temperature cycles in reverse. The
first high peak corresponds to the condition at
1300°F with the blower off. As the blower was
turned on the temperature decreased to 1150°F,
at which point, as noted, the blower reverse oc-
curred with the resultant decrease in blower speed.
The heat exchanger outlet temperature then rose
to 1250°F before the blower speed was allowed to
increase again.

The data from these and other recorder charts
during this time, together with calculated rates
of change, are given in Table 4.5.

83

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TIME OF DAY

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1410 1400 1350 1340 1330 1320 1310 g Z 1415 4405 1355 1345 1335 1325 4315
TIME OF DAY z ‘ l] & TIME OF DAY
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CONDENSATION OF MICROMICROAMMETER 1000
READINGS FROM 80in./hr TO 4in./hr 1408 1400 1352 344 1336 1328 1320 {342
TIME OF DAY
50
INDIVIDUAL FUEL TUBES INLET AND OUTLET TEMPERATURE
0
1410 1400 1350 1340 1330 1320 1310
TIME OF DAY

Fig. 6.13. Reactor Characteristics During Power Cycling.

 

85

 
   

 

 

TABLE 6.5. CHARACTERISTIC NUCLEAR AND PROCES:S. DATA DURING POWER CYCLING OF REACTOR (NOV, 12)

 

 

 

 

 

 

 

 

Power Cycle No. 1 Power Cycle No, 2 Power Cycle No. 3
Type of Type of Data Blower®  Blower®  Blower?  Blower® | Blower?  Blower®  Blower®  Blower®  Blower®  Blower?
Measurement Obtained® Speed Speed Speed Speed ’ Speed Speed Speed Speed Speed Speed (Red
Increasing Decreasing Increasing Decreasing} Increasing Decreasing Increasing lIncreasing Decreasing Increasing Withdrawn
Time 1324 1326 1327 1340 ‘ 1346 1348 1349 1405 1406 1407 1408
Log N power Py, Mw 0.283 2.26 1.04 2.01 | 0.462 2.42 1.62 0.387 2.06 0.962 2.01
P o Mw 2.26 1.04 2,03 0.48 ‘, 2.42 1.62 2.09 2.06 0.962 2.01 3.21
8P, Mw 1.98 ~1.22 0.99 ~1.53 \ 1.96 ~0.80 0.47 1.67 ~1.10 1.05 1.20
81, sec 35 69 37 275 38 37 42 35 70 40 23
8P/t, Mw/ sec 0.057 —0.18 0.027 _0.0056 | 0.052 —0.022 0.011 0.045 ~0.016 0.026 0.052
T sec 18 -89 54 ! 24 59 110 19 60 52 46
Fuel outlet T, °F 1402 1526 1457 1551 1446 1559 1540 1426 1538 1485 1532
temperature T, 5 °F 1526 1452 1529 1446 1559 1540 1558 1538 1485 1532 1595
8T , °F 124 ~74 72 -105 13 -19 18 112 —53 47 63
81, sec 48 76 50 321 49 37 46 44 55 41 39
8T /81, °F/sec 2.58 ~0.97 1.44 ~0.33 2.31 ~0.51 0.39 2.55 —0.96 1.15 1.62
Fuel inlet T; 1 °F 1340 1208 1271 1212 1347 1209 1247 1333 1207 1262 1207
remperature 7, o °F 1208 1271 1210 1346 | 1209 1247 1218 1207 1267 1207 1254
8T, °F ~132 63 ~61 134 . —138 38 ~29 ~126 55 ~55 47
Ot, sec 65 56 49 356 |53 35 46 49 52 49 62
8T /81, °F/sec ~2.03 1.13 ~1.24 0.38 ; ~2.60 1.09 -0.63 —2.58 1.06 1.12 0.76
Fuel mean T, v °F 1371 1367 1364 1381 1397 1384 1394 1380 1373 1374 1370
remperatre T2 °F 1367 1362 1369 1396 1384 1393 1388 1372 1373 1370 1424
8r_, °F —4 -5 5 15 ~13 9 -6 -8 0 —4 54
t, sec 56 66 50 338 51 36 46 47 45 50
8T /81, °F/sec ~0.04 ~0.076 0.10 0.04 . —0.25 0.25 ~0.13 -0.17 0 —0.09 1.08
Reactor fuel AT, °F 62 318 186 339 L 99 350 293 93 331 223 325
AT AT, °F 318 181 319 00 | 350 293 340 331 223 325 341
SAT), °F 256 ~137 133 —239 i 251 -57 47 1238 —108 102 16
5t, sec 57 66 50 338 -8 36 46 46.5 54 51 55
OAT)/ O, °F/sec  4.49 —2.08 2.66 ~0.71 ! 4.92 —~1.58 1.02 5.11 - -2.00 2.00 0.29

 

 

%See footnote to Table 6.3.
bDuring power cycling no data available on blower speeds. J-
©|_ow-heat-exchanger-temperature blower reverse coused blower speed to decrease. No data available on blower speeds.

dNo data recorded on which rod was moved. Estimated Ak/k introduced was 0.33%. This would correspond to a group movement of all three shim rods for a 10-sec period.

86

 

+ e TR T R T N Ty R T T T T T R T T Bk

   
. me

Reactor Transients

One of the characteristics of the reactor that was
most important from the operator’s point of view
was the behavior under transient conditions, be-
cause it is only through transient effects that the
operator acquires a ‘‘feel’’ for the way the reactor
responds to his control under varying conditions.
The various ways in which transients may be
introduced into the reactor are the following:
operation of fuel system helium blower,
operation of sodium system helium blowers,
operation of rod cooling system helium blowers,
changing the fuel flow rate,
changing the sodium flow rate,
movement of shim rods,

.

NO VAW

movement of regulating rods.

A discussion of the effects on the ARE of some
of these operations is given in succeeding para-
graphs. Effects 3 and 5 were very small although
observable. Effect 4 was not observed at high
power because of the danger of freezing fuel in the
pump, although it can be calculated from the inhour
curves (Fig. 5.11) and the known rate of change
of fuel pump speed, Table 6.6 lists calculated
rates of interjection of Ak/k into the reactor by
the various means listed above,

It has already been pointed out that the ARE
reactor and system were very sluggish in responding
to demands at high power. Furthermore, it was
observed that the response at low power (less than
100 kw) was much different from its response at
higher powers {at 1 Mw or greater). One way to
examine the behavior of the reactor is to plot sys-

tem characteristics against the initial power, since
this represents the behavior from a given initial
condition of the reactor.

One of the properties investigated in the above
manner was the reactor period. Some of the ob-
served reactor periods are plotted as a function of

the initial power in Fig..6.14. Since the reactor

could be put on any period from infinite to a given
minimum, the points in Fig. 6.14 are scattered.
Most of the longer periods represent periods ob-
served during such times as the initial rise to
power, which was approached with caution. As
the operators became more familiar with the oper-
ation at power, deliberate attempts were made to
see how fast a period the reactor would attain due
to blower operation or rod movement. As a result,
definite lower limits were established beyond
which reactor periods could not be induced by the
controls available to the operator. The solid line
in Fig. 6.14 represents the lower limit for periods
during blower operation and the dashed line is that
corresponding to shim and regulating rod move-
ment.” The smallest period observed during high-
power operation was the 10-sec period represented
by the rise in the Log N chamber at 1259, Fig.
6.8. A few smaller periods were observed in the
very low-power regime just above critical. The
two lines shown in the figure indicate that the
lower the power the greater the transient that can

 

9l\‘ the reactor had been taken to power with the shim
rods inserted to greater depths, smaller periods would
have been observed for shim rod motion tfiun dare repre-
sented by the dashed line, since the reactivity value
of the rods would have been greater for those insertions,

TABLE 6.6, CALCULATED RATES OF INCREASE OF Ak/k IN ARE FROM YARIOUS OPERATIONS

 

Range (%AkR/E)/ sec

 

Operation

Fuel system helium blower speed increase 0 to 1500 rpm 0.0035
1000 to 1500 rpm 0.0011

Regulating rod movement Whole rod insertion or withdrawal 0.011

Shim rod movement, single rod At 4 in. insertion 0.0039
At 8 in. insertion 0.0067

Shim rod movement, group operation At 4 in. insertion 0.012
At 8 in. insertion 0.019

Fuel flow rate 0 to 46 gpm 0.013

Sodium flow rate 0 to 150 gpm 1x1078

 

87

§
b
P
i
s

 

n

 

 
ey

 

 

ORNL-LR-DWG 6560

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

200 1
® PERIODS RESULTING FROM FUEL SYSTEM ’
HELIUM BLOWER OPERATION
© PERIODS RESULTING FROM ROD MOVEMEN
100
2 L
8 S N —— b & 7 J —_—
'CE i
E 50 R | *,fif.j,, R
FASTEST PERIODS :
e o ___ !FROM BLOWER OPERATION: >/ L n
g .~
O
g SR RN DR N /w A
" _ / 7"~ FASTEST PERIODS
» FROM SHIM ROD AND
| / , - REGULATING ROD MOVEMENT
0] 0 !
20 - o PR - e — i e -
® j /’
A
pre
- /
0 ] ~
10 ~
0.05 [oX) 0.2 05 1.0 2 5 10

INITIAL EXTRACTED POWER {Mw)

Fig. 6.14. Induced Reactor Periods as a Function of Initial Extracted Power.

be introduced into the reactor, i.e., the higher the
power the harder it becomes to introduce a transient
condition. Extrapolation of the solid line to a
power of 1 kw indicates that at this power reactor
periods of 1 sec or less could have been intro-
duced by blower operation.

It should be pointed out that periods could be
observed which were not true periods but were a
combination of the observed period and its time
derivative. Whenever a sudden reactivity change
occurred, such as when a rod was moved slightly,
the period meter would register a transient peak
period that was very small compared with the true
period. The beginning peak in the period curves
of Fig. 5.4 are of this nature. Whenever a con-
tinving change of Ak/k was taking place, the ob-
served period was not a true period but, again, a
combination of the true period and its time deriva-
tive. In Fig. 6.10 the period meter did not register
a constant period when the regulating rod was
moved but, rather, a continually changing period.
Also, it is to be noted that the Log N power re-
corder did not register straight slopes during the
power changes but, instead, constantly changing

88

ones, which again indicated that the period was
changing (cf., App. S).

The time behavior of the fuel temperatures as a
result of blower operation was obtained by plotting
inlet and outlet fuel temperature time rates of
change against the initial power. These plots are
shown in Fig. 6.15. Larger rates of change of fuel
temperatures were observed at lower powers, and
they corresponded to the smaller periods observed.
The greatest rates of change of the fuel tempera-
tures were observed at an initial power of 280 kw,
at which time the inlet temperature changed at a
rate of ~2.75°F/sec and the outlet temperature
at a rate of +2.6°F/sec. The only reason higher
rates of change were not observed at powers
lower than this was that in the low-power regime
the kinetic behavior was not well known and there-
fore a conservative approach to transient con-
ditions was used.

The rates of decrease of temperatures when the
biower was turned off and the reactor was allowed
to come to a lower power are also plotted in Fig.
6.15. The trend of rates of change of the tempera-
tures seems to be reversed.

»

e EEETT P

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ORNL-LR-DWG 6561

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

4
3 \\
~ GREATEST QUTLET TEMPERATURE RATE OF INCREASE
.
© oo
O~
» e GREATEST OUTLET TEMPERATURE RATE OF DECREASE
— Bl --.--_-.-
o — —
{, O - S ——
¥ Tt 00— —
s o e
u 0 e
©® VT T T T T T T LT g oo
‘<2[ C) / _________ cegpemmenmnTTTTT
5 1l ] b L z
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= ® e v ¢ - T
3 A .. ® ° ¢ L\ —_—
s e ——
o A P A
-y T mmheme-e- ;;_ e e
& A A - e bo--
L A //
e - ,<\ GREATEST INLET TEMPERATURE RATE OF INCREASE
& ~ GREATEST INLET TEMPERATURE RATE OF DECREASE
-2 A g
L /
L
/‘ O FUEL OUTLET TEMPERATURE RATE, HELIUM BLOWER SPEED INCREASED
=3 P A FUEL INLET TEMPERATURE RATE, HELIUM BLOWER SPEED INCREASED T~ — "]
s ® FUFL MEAN TEMPERATURE RATE, HELIUM BLOWER SPEED INCREASED
& FUEL CUTLET TEMPERATURE RATE, HELIUM BLOWER SPEED DECREASED
A FUEL INLET TEMPERATURE RATE, HELIUM BILOWER SPEED DECREASED
-g |
o 0.5 1.0 1.5 2.0 2.5

INITIAL EXTRACTED POWER (Mw)

Fig. 6.15. Time Behavior of Reactor Fuel Temperatures as a Function of Initial Extracted Power.

Some very general observations on the transient
behavior of the ARE reactor are summarized below.

Reactor Periods. |n general, the lower the initial
power the faster was the transient introduced by
any type of operation that resulted in upsetting the
equilibrium between nuclear and extracted power.
The smaller periods were associated with lower
initial powers.

Oscillation of Reactor Power. The character-
istic behavior of the nuclear power during an estab-
lished transient was to ‘‘over-shoot’’ and then
oscillate around a mean value before settling down
to that power. The period of the osciilation ap-
peared to be of the order of 2 min and lasted for
about 2 cycles. '

Movement of Rods. The reactor was always more
prompt to respond to rod motion than it was to
blower operation. Also the reactor periods observed
at a given power were smaller due to rod movement
than those due to blower operation.

Nuclear Power., Whenever blower speed was
changed the nuclear power rose {with oscillations

as noted above) to a higher power level than the
corresponding extracted power demand. When a
rod movement occurred, the nuclear power changed
considerably, which resulted in higher or lower
reactor over-all temperatures, and then leveled out
to a new balance with the extracted power at about
the original power level.

Extracted Power. Extracted power was essen-
tially a function only of the demand and changed
due to a change in operation of one of the various
heat exchanger blowers. During rod movement at
a given demand power, the extracted power re-
mained essentially constant. The unbalance be-
tween extracted and nuclear power resvited in a
change of the mean reactor temperature.

Reactor OQutlet Fuel Temperature. The outlet
fuel temperature always rose and fell in the same
direction as the nuclear power level. For blower
operation the rise in the outlet temperature was
always less than the fall of the inlet temperature.
The opposite was true for regulating rod movement.

Reactor Inlet Fuel Temperature. The inlet fuel

89

 

= T TR

g
j

 

 

 
i temperature rose and fell in the opposite direction
] to the change in nuclear power during blower oper-
ation and in the same direction as the nuclear
power during rod movements, The inlet tempera-
ture change was greater than the outlet temperature
change for blower operation. The opposite was
true for shim rod operation.

Reactor Mean Temperature. The reactor mean
temperature followed the trend of the reactor inlet
temperatures in all cases, except that the changes
and rates of change were much smaller.

Time Lags. The system was very sluggish be-
cause of the long tansit time (47 sec) of the fuel.
[n addition to the sluggishness of the system, there
g appeared to be time lags between the responses of
: various temperature indicating instruments for the
same action. These lags were of the order of 2
min for low-power operation {less than 100 kw) and
of the order of T min for full-power operation in the
megawatt range. The topic of time lags is dis-
cussed in greater detail in a following section.

 

Calculated Power Change Resulting from a
Reguiating Rod Movement

A formula was developed which would estimate
a power change in the reactor from a given change
in the reactor outlet temperature and a regulating
rod movement. This was possible primarily be-
cause of the relationships existing between P,
AT, .and Ak/k. The development of the relation-
ship is described below.

 

AP = 2(0.11) (46)

AT -
o

It

0.72 x 9.8 x 10™3

 

When equilibrium exists between extracted power
and nuclear power, the power level is given by

P = kq AT = le:q(TO -T),

1

where
k = a constant (specific heat of the fuel),
. g = tuel flow (gpm),
AT = the difference between inlet and outlet
- fuel temperatures,
T, = outlet fuel temperature,
T, = inlet fuel temperature.

%

 

Also the mean reactor fuel temperature is given by

1
T =7(T0+Ti)'

m

and
AT = To - T

From this it is seen that %
AT = 2T0 - 2Tm ,

and

P = 2kgq (To - Tm) . .

Now consider a change AP in the power:
AP = 2kq (AT, - AT ) ,

but the temperature coefficient reactivity a is

 

Ak/k
o =
AT
so that
1
AP = 2kq [ATO - —(Ak/k)]
a
Since
Ak/k = [(Ak//e)/in.]RR x Adpyo
where ‘
Adpp = regulating rod movement,
[(Ak/R)/in] , , = 0.033% in. for the regulating

rod, .

by putting in the experimental values,'? it is seen
that

(0.00033) Ad | x 10™3 Mw

(10.12 AT + 47.75 Ad) x 1073 Mw .

The change in power, as calculated from this
formula, checked closely the observed power change
for several different cases which were examined.
The calculated data are given in Table 6.7.

Reactor Temperature Differential as a Function
of Helium Blower Speed
A relationship between the fuel system helium

blower speed and the resulting fuel AT was ob-
tained from the data token during ARE operation.

 

107he factor 0.72 is used to correct the temperatures
in the manner described in Appendix K.

. ®

e e

g T

Frowpewt b e

CEoavrr

W T e Nmemmer T

woE T ey

-l I ERT TY T ETEEER TOT AT Coppe cEE AP F R wemmPReT “rmwer P

 
R e

Akl

&

e

WK, v

TABLE 6.7. SOME CALCULATED AND OBSERVED POWER CHANGES FROM REGULATING ROD MOTION

 

Observed Qutlet

Regulating Rod

Calculated Power Observed Power

 

Date Time Temperature Change ATO Movement Change Change
°F (in.) (Mw) (Mw)
Nov. 10 1110* 55 12 1.59 1.60
1114* 53 12 1.55 1.51
1321 13 2.8 0.37 0.36

 

*Data recorded during Exp. H-8.

Some of the experimental fuel AT's are plotted in
Fig. 6.16 as a function of their corresponding
blower speeds. These points could be represented
by a parabola of the form

AT? = Ks ,
where
K = a constant of proportionality,
s = the blower speed.

The curve that fit the experimental points best was
found to be

AT = (0.755)V/2 |

where AT is expressed in hundreds of degrees F
and s is given in hundreds of rpm.

The extracted power can now be expressed in
terms of blower speed:

P = kqAT = kq(0.75s)1/2
= [0.1] x 46 x (0.75s)'/2] x 102 x 10~3
= 0.506 (0.755)'/2 Mw .

The factor 102 is used because AT is expressed
in the empirical formula in hundreds of degrees.

o~

The 10-% factor expresses power in megawatts.
The results obtained with the formula checked
with the observed values of power in numerous
cases. As an example, during experiment H-3, for
a blower speed of 1590 rpm, an extracted power
of 1.75 Mw was observed. The formula gives

P = 0.5060.75 (15.9)1"/2 = 1.75 Mw .

The Phenomenon of the Time Lag

 One of the most noteworthy phenomena encoun-
tered during the experiment was the time lag that
seemed to be an intrinsic part of the system be-
havior, It was remarkably demonstrated in Fig. 6.8
when the blower was turned on. [t was 0.75 min

ORNL—LR—DWG 6562
400

350

AT = (075572

300
—~ 250
w
K

5 200
AT IN HUNDREDS OF DEGREES F

% 150 i S IN HUNDREDS OF rpm
w

100

50

 

o 400 800 1200 1600 2000
FUEL SYSTEM HELIUM BLOWER SPEED (rpm)

Fig. 6.16. Reactor AT as a Function of Fuel

Helium Blower Speed.

later that the inlet temperature records showed a
response, about 1.5 min before the mean tempera-
ture began dropping, and almost 2 min before the

outlet temperature showed a corresponding rise.
The reasons for the lags are obscure, but in an

ottempt to analyze the phenomenon several things
must be considered: the extensiveness of the sys-
tem, the reactor geometry, i.e., location of inlet
and outlet fuel passages with respect to reactor
core, heat transfer properties and possible phe-
nemona within reactor, location, design, and attach-
ment of the thermocouples within the reactor, and
other unknown factors. There is a good possibility
that the time lag phenomenon was connected with

the temperature discrepancies noted in Appendix K.

Errors in marking the charts could possibly
account for as much as 0.5 min, but certainly this
cannot wholly account for the lag which was noted

91

 

 
 

 

consistently throughout the progress of the experi-
ment. There is no doubt the system was vefy
sluggish and slow to respend. This sluggishness
was undoubtedly due, in large part, to the length
of the fuel system. The transit time in the system
was about 47 sec at full pump speed, as observed
during the critical experiments. There does not
seem to be any convenient mechanism for ex-
plaining longer time lags. A partial explanation
is advanced in Appendix O, where geometrical
considerations show that the instantaneous temper-
ature change of the fuel was greater near the center
of the reactor than at the points of the thermo-
couple locations. Thus the nuclear power was
observed to change before the inlet and outlet
thermocouples indicated changes.

An attempt was also made to analyze the heat
transfer -conditions within the reactor. Since the
fuel and sodium flows were in opposite directions
within the reactor, the fuel inlet line thermocouples
were located near sodium outlet lines and the fuel
outlet line thermocouples were located near sodium
inlet lines. The temperature differences existing
between the sodium (and moderator) and the fuel
near the inlet and outlet thermocouples were of the
order of 100°F or more and could possibly have
influenced the readings of the thermocouples. This
phenomena was invesfigm‘ed,H and no delay ap-
proaching 2 min could be ascribed to the heat
transfer characteristics of the system. The greatest
time lag which could be found in the heat transfer
mechanism was of the order of a few seconds. The
time lags of the reactor AT and mean temperature
thermocouples were also investigated because
these thermocouples had been insulated from the
metal of the system. The time lags found were of
the order of 15 sec or less, and thus they were
about the same as those for uninsulated thermo-
couples.  An experimental check showed that

~ helium from the rod-cooling system had no appreci-

able effect on the readings of fuel temperatures
when the blower speed was changed. Therefore,
the phenomena of lag and low temperature readings
could not be attributed to this cause.

It may well be that what may be described as
““thermal inertia’’' played a part in the time lag.
Since the reactor as a whole did not respond to a
temperature change very fast, even though the
fuel did, the thermocouple readings could have
been affected to the extent that they received heat

 

”H. F. Poppendiek, Reactor Experimental Engi-
neering,

92

not only from the fuel tubes but from other parts
of the reactor. Better design and focation of the
thermocouples might have prevented such time lags.

Reactivity Effects of Transients in the
Sodium System

The introduction of transient conditions into the

sodium system was expected to have only small
effects on reactor behavior. This proved to be the
case in the few instances for which data are avail-
able. The two operations of the sodium system
which were observed to have effects on reactivity
were changing the rate of sodium flow and changing
the rate of sodium cooling.

The first effect was observed as a part of one of
the low-power experiments (Exp. L-7), in which,
with the reactor on servo at l-w power, the fuel
flow rate was varied and the change in the regu-
lating rod position was noted. The concluding
portion of the experiment consisted in stopping the
sodium coolant flow and observing the change in
regulating rod position with the fuel flow at its
normal value of 46 gpm. During this time the regu-
lating rod changed from 7.85 to 7.69 in., a move-
ment of 0.16 in., which corresponded to a Ak/k of
only 5.3 x 1073, This experiment was not repeated
at high power, and therefore no data are available
on reactor power and temperature changes as a
function of sodium flow.

The second effect, the effect on the reactor sys-
tem of changing the sodium cooling rate, was ob-
served in Exp. H-12 during high-power operation.
In this experiment the sodium system helium
blowers were turmed off and the following power
and temperature changes were observed. At 2110
on November 11 the reactor was at 2. 1-Mw extracted
power; of this amount, approximately 1.6 Mw was
being extracted from the fuel and 0.5 Mw from the
sodium. At 2113 the sodium system helium blowers
were turned off, and the power dropped, in the next
90 sec, to 1.96 Mw at a rate of 1.5 kw/sec, and,
in the next 9% min, it dropped at a much slower
rate of 280 w/sec and leveled off at 1.80 Mw. The
total decrease in power was 300 kw. It is to be
noted that since the extracted power from the
sodium at the beginning of the experiment was
500 kw, about 200 kw of heat was still being lost
by radiation and other means through the barrier
door openings and other parts of the circuit, even
with the helium circulation stopped. When the
helium blower was turned on again at 2124 the

reactor responded with an initial rise of 120 kw to

 

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1.92 Mw at the rate of 1.33 kw/sec and then a very
slow rise over a period of nearly 18 min to 2.06 Mw
at a rate only 0.1 as great.

Any over-all reactor temperature effect from
sodium transients would have been reflected in a
change in the mean fuel temperature, and therefore
the mean fuel temperature was observed. When
the sodium blower was turned off at 2113 the re-
actor mean fuel temperature dropped from 1313 to
1310°F in 3 min, corresponding to the initial power
decrease noted. There was no drop in temperature
which matched the very slow rate of decrease in
power after the initial decrease. Upon starting the
sodium blowers again at 2124 the mean temperature
rose 13°F in 5 min to a new value of 1323°F, 10°F
higher than at the start of the experiment. The
increase over the initial temperature was partly
attributable to the thermal inertia of the reactor.
The reactor did not lose heat rapidly, but, when the
blowers were again turned on, the nuclear power
production increased by 300 kw and raised the
temperature level to above that at the initial con-
dition.

It may be concluded from these results that
neither siopping the sodium flow or the sodium
cooling had much effect on the reactor behavior.
It is interesting to observe that the mean reactor
fuel temperature went down with the decrease in
nuclear power, which resulted from stopping the
sodium blowers. |t might be suspected that the
reactor mean temperature would have risen because
of lack of cooling in the moderator.

Reucii\)iiy Following a Scram

The behavior of a reactor after a scram is de-
termined primarily by delayed neutrons and photo-
neutrons, |2 which in the ARE were produced in
the beryllium oxide moderator. These processes
keep the flux of the reactor from decaying immedi-
ately. '

It was observed that after every scram the re-
actor decay took place in a series of descending
oscillations, the time between oscillations being
equal to the fuel transit time of 47 sec. Figure
6.17 shows a trace of one of the fission chambers
after the scram at the conclusion of Run 7 of Exp.
L-5, one of the low-power runs. The scram took
place at the exireme right-hand edge of the trace.
Six prominent pips and many lesser ones can be
seen as the decay progressed.

 

}25. Glasstone and M. C. Edlund, Elements of Nuclear
Reactor Theory, Yan Nostrand, p 88-89 (1952).

This behavior can be explained on the basis that
the slug of fuel passing through the reactor at the
instant before the scram carried the last group of
delayed neutron emitters produced at power. This
phenomenon was particularly pronounced at the
time the photograph of Fig. 6.17 was taken because
it was at that time that an experiment'® was under
way in which the reactor was allowed to rise from
1-w power to a predetermined power level of from
50 to 100 w on a constant period before it was
scrammed. At the instant of scram the power of.the
reactor was 100 w, whereas one transit time previ-
ously (47 sec) the power was only about 14 w. As
a result, immediately after the scram the delayed
neutrons were particularly intense in the slug of
fuel which was in the reactor at the time of the
scram. The transit time of 47 sec is comparable
to the 56-sec group of delayed neutrons.'® As
this slug traversed the system and went back
through the reactor, the delayed neutron emitters
gave rise to neutron multiplication within the re-
actor and, at the same time, the associated gamma
rays fell on the beryllium of the moderator and
gave rise to more neutrons by the reaction

Be’ +y—>Be8 + n .

The total effect of both types of reactions was
to give a strong multiplication every 47 sec. During
the first pip shown in Fig. 6.17, the flux rose by
a factor of 2; with succeeding pips, the multipli-
cation became less. The average mean decay time
of the first four pips was 72 sec. For the two
longest lived groups of delayed neutrons the mean
lives are 32 and 80 sec.'? Therefore, the attenu-
ation of the pips closely followed the theoretical
delayed neutron decay. A remarkable feature of
Fig. 6.17 is that the pips could be distinguished
for about 12 min. Since after the first 3 or 4 min
the delayed neutron emitters were gone, the re-
maining effect was due solely to photoneutrons.

The scram behavior at high power was very
similar, as observed with the use of the safety
chambers. The fission chambers could not be
used at high power, and therefore the first few
seconds after a scram could not be observed with
them.

 

1314 experiment L-5 the regulating rod was calibrated
by the periods induced by rod motion {cf., chap. 4).

Mqussfone and Edlund, op. cit., p 65.

93

-

 

 

k.

 
 

 

94

COUNTING RATE (arbitrery scale)

C.R.M. NO.1; RUN 7, EXP. LS
2212-30; NOV. 7, 1954

THIS FIGURE IS A COMPOSITE OF TWO CHARTS SHOWN
= HERE PIECED TOGETHER TO GIVE A GRAPHIC REPRE-
= SENTATION OF THE DECREASE IN COUNTING RATE IN
THE CIRCULATING SLUGS FOLLOWING A SCRAM, THERE
— |$, HOWEVER, SOME ERROR IN BOTH ABSCISSA AND
- ORDINATE ACROSS THE JOIN,

—~s———— TIME

Fig. 6.17. Circulating Slugs After a Scram,

ORNL-~LR-DWG 3855

 

 

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FINAL OPERATION AND SHUTDOWN

The last scheduled experiment conducted on the
reactor was the measurement of the xenon buildup
following the 25-hr run at 2,12 Mw., The 10-hr
period of operation at one-tenth full power was
concluded at 0835 on November 12. During the
following ]]1/2‘|1r period from 0835 to 2004, when
the reactor was shut down the final time, the oper-
ation of the reactor was demonstrated for Air Force
and ANP personnel who were gathered for the
quarterly ORNL-ANP Information Meeting. The
demonstrations inciuded repeated cycling of the
foad by turning the blowers on and off, and group
movement of the three shim rods to change the
reactor mean temperature. The information ob-
tained during this time on” the dynamic behavior of
the reactor is described above under the subtitle
“’Reactor Kinetics."”'

Also during the final 11%-hr period of operation,
two complete surveys of system temperatures were
taken with the reactor at maximum power. It was

during one of these runs that the equilibrium fuel

outlet temperature of 1580°F and the total reactor
power of 2.45 Mw was attained (cf., App. L).

Since all operational objectives of the experiment
had been attained and it was estimated on the
basis of time and the assumed reactor power that
the desired integrated power of 100 Mwhr would be
attained by 2000 on Friday, November 12, it was
decided to terminate the experiment at that time.
Colonel Clyde D. Gasser, Chief, Nuclear Powered
Aircraft Branch of WADC, who was then visiting
the Laboratory, was invited to officiate at the

termination of the experiment. At 8:04 PM on
November 12, with Colonel Gasser at the controls,
the reactor was scrammed for the last time and the
operation of the Aircraft Reactor Experiment was
brought to a close. A photograph taken in the con-
trol room at that time is shown as Fig. 6.18.

At the time the experiment was terminated it was
believed that the total integrated power was more
than the 100 Mwhr which had been prescribed as
a nominal experimental objective, but subsequent
graphical integration of the Log N charts, Ap-
pendix R, revealed that the actual total integrated
power was about 96 Mwhr. At the time the reactor
was scrammed the sodium system had been in
operation (circulating sodium) for 635.2 hr and the
fluoride system for 462.2 hr. Of this total fluoride
circulating time 220.7 hr were obtained with the
reactor critical and 73.8 hr after the reactor was
first brought to power.

Although the nuclear operation was concluded
on Friday evening, the fuel and sodium were per-
mitted to circulate until the following morning, at
which time they were dumped into their respective
dump tanks. The final phase of the experiment,
including the dumping operation, subsequent analy-
sis and recovery of the fuel, and examination of
the systems and components for corrosion, wear,
radiation damage, etc., are to be discussed in a
subsequent report. Since much of the data cannot
conveniently be obtained until the radiocactivity
has decayed, the last report will not be forthcoming
immediately.

95

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7. RECOMMENDATIONS

No comprehensive report is complete without
conclusions, discussion, and recommendations. In
this report the conclusions are presented in the
summary at the beginning, and the discussion is
incorporated in the text and, especially, in the
Appendixes. Furthermore, certain alterations and
modifications that would have been desirable in
the conduct of the experiment are implicit in the
discussions throughout the report. However, the
changes, if effected for any similar future ex-
periments, could lead to substantial improvements
in design, instrumentation, operation, nd interpre-
tation. Accordingly, a list of recommendations
based on the ARE experience is presented; no
significance is inferred by the order of listing.

1. The operating crew should maintain the same
shift schedule as the craft labor — electricians,
instrument mechanics, pipe fitters, etc. — assigned
to the operating crew, and all members of one crew
should have their off-day at the same time. In
particular, there should be four operating crews
that maintain the same rotating schedule as the
rest of the plant.

2. The cause of the obvious discrepancy between
the temperatures read by the line thermocouples
close to the reactor and those further removed
from the reactor should be ascertained. No physi-
cal phenomena, with the possible exception of
radiation, have been proposed, to date, which could
account for the observed difference.

3. In order to measure the extracted reactor
power from the secondary or tertiary heat transfer
mediums, these systems should be adequately
instrumented for flow rates and temperatures.
Such instrumentation might also make possible
a thermodynamic analysis of the performance of
the various heat exchangers. _

4, Therrinocourpkles on the outer surface of fuel
and sodium tubes in the heat exchangers should be
installed so that they read wall temperatures rather
than an intermediate gas temperature.

5. Of the two thermocouples which were required
to measure the reactor mean temperature and the
two required to measure the temperature gradient,
one was electrically insulated from the pipe wall.
The insulation effected a time lag between the
wall and thermocouple temperatures of about
15 sec. Thermocouple installations for obtaining
time-dependent data should be made so that the
wall temperature can be read without a time lag.

6. To heat small {up to 3/8 in. OD) lines and
associated valves, tanks, etc. to uniform high
temperatures (~1400°F) requires a surprising
degree of precision in the installation of both
heaters -and insulation. Furthermore, where calrod
heaters are employed, they should be installed on
opposite sides of a line, and the heaters and line
should be jointly wrapped with heat shielding
before insulation is applied.

7. To control accurately the temperature of any
system, thermocouples should be installed at
any discontinuity of either the heaters (i.e.,
between  adjacent heaters) or the system (i.e., at
the junction of two lines, etc.), and, with the
exception of obviously identical installations,
each heater should have its own control.

8. Where gas lines are subject to plugging due
to the condensation of vapor from the liquid in the
system to which the lines are connected, it is
necessary either to heat the lines to above the
freezing point of the vapor or to employ a vapor
trap. While a vapor trap proved satisfactory in
preventing the fuel off-gas line from plugging, it
was not possible to heat the sodium off-gas line
sufficiently (because of temperature limitation of
the valves) to prevent the gradual formation of a
restriction. Gas valves that can be operated at
higher temperatures and are compatible with
sodium are needed.

9. The double-walled piping added a degree of
complexity to the system that was far out of pro-
portion to the benefits derived from the helium
annulus. Parts of the annulus were at subatmos-
pheric pressure, and no leak tests were made on
the fuel system once it was filled with the fluoride
mixture. The helium flow was not needed for
distributing heat in the sodium system, and it was
of uncertain value in the fuel system. Conse-
quently, future systems should not include such
an annulus.

10. The use of any type of connection other
than an inert-arc-welded joint in any but the most
temporary fuel or sodium line should be avoided.

11. All joints, connections, and fittings in the
off-gas system should be welded, and, in general,
they should be assembled with the same meticulous
care that characterized the fabrication of the fuel
and sodium systems in order to minimize the pos-
sibility of the unintentional release of fission
gases.

97

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12. As a secondary defense in the event of the
release of activity in the reactor cell (i.e., pit),
the cell should be leaktight. The leak-tightness of
the numerous bulkheads out of such a cell should
be carefully checked.

13. The air intake to the control room was lo-
cated on the roof of the building, and thus adverse
meteorological conditions readily introduced off-
gas activity into the control room. With an airtight
control room equipped with its own air supply (or
a remotely located filtered intake), the control room
operations could continue without concern for the
inhalation of gaseous activity.

14. The off-gas monitrons should be shielded
from direct radiation. Furthermore, although these
monitrons were nof needed during the experiment,

~ they are known to build up background activity

which would eventually mask that of the gas they
are to measure. The development of monitrons in

‘which activity will not aceumulate is recommended.

15. The radiation level and the airborne activity
throughout the building were measured by monitrons
and constant air monitors, respectively. The data
from both instruments should be continuously
recorded. Furthermore, the output of the air moni-
tors should be modified by the addition of a dif-
ferentiating circuit so that the recorded data would
give better measures of the airborne activity.

16. The helium ducts leaked to the extent that
it was possible to attain only a fraction of the
desired helium concentration therein. They were
of the conventional bolted-flange design and should
have been modified to permit seal welding of all
joints; they should be subjected to stringent leak
tests.

17. The helium consumption for the last three
and one-half months of the ARE experiment was
2/3 million standard cubic feet (over 3000 standard
cylinders). The average consumption rate during
the last two weeks of the experiment was about
8.5 ¢fm, and peak consumption rates of 25 c¢fm were
recorded. This extraordinarily high helium con-
sumption could be reduced by the use of dry air
or nitrogen for much of the pneumatic instrumen-
tation in which helium was used.

18. Various valves in both the gas and liquid
systems leaked across the valve seats. The valve
development program should be emphasized until
valves are obtained that can be depended upon as
reliable components of high-temperature (~1200° F)

‘sodium, fluoride, or gas systems. Furthermore,

either limit switches that would operate at higher

98

temperatures to indicate valve open or closed
should be developed, or the existing limit switches
should be located in cooler regions.

19. The use of frangible disks to isolate the
standby fuel pump did not enhance the feasibifity
of continuing the experiment in the event of the
failure of the main fuel pump during high-power
operaticn. The frangible disks are objected to in
that they require a nonreversible operation. In
general, leak-tight valves should be developed
that may be used rather than the frangible disks.

20. Although the ARE, as designed, was to
incorporate numerous ‘‘freeze sections,’’ only two
were included in the system as finally constructed,
and the operability of these two was so question-
able thdt their use was not contemplated during
the course of the experiment. Such sections should
either be eliminated from consideration in future
systems, or a reliable freeze section should be
developed.

21. Much useful information would be obtained
if provisions could be made for sampling each
liquid system throughout the operation. This was
possible on the ARE only prior to the high-power
experiments.

22. The variable inductance-type flow and level
indicators should be converted to the null-balance
type of instruments in order to eliminate the
temperature dependence of the pickup coil signal.
Furthermore, these coils should be located in a
region at much less than 1000°F in order to in-
crease coil life.

23. Although numerous spark plug probes were
satisfactorily employed to measure levels in the
various tanks, the intermittent shorts that were
experienced with several sodium probes might
have been avoided if clearances between the
probe wire and the stand pipe in which the probes
were located had been greater.

24. The use of mercury alarm switches on
vibratory equipment where there is little leeway
between the operating and alarm condition will
give frequent false alarms and should therefore be
avoided.

25. Weight instruments are of questionable
accuracy in a system in which the tank being
weighed is connected through numerous pipes to
a fixed system, especially when these pipes are
covered with heaters and insulation and are subject
to thermal expansion,

26. The flame photometer is an extremely sensi-
tive instrument for the detection of sodium and

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NaK. However, in employing this instrument to
detect the presence of sodium (or NaK) in gas, the
sampling line should be heated to about 300°F,

27. The magnet faces in the shim rods should
be designed so that dirt particles cannot become
trcpped thereon and thus require higher holding
currents. '

28. The two control points, i.e., the upstairs
control room and the basement heater and instrument
panels, should have been more convenient to one
another in order to effect the greatest efficiency
of operation.

29. The communications system in the building
was inadequate. Except for the auxiliary FM
system, which was frequently inoperative, there
was only one phone in the control room, which
was on the main station of the PA system. How-
ever, the PA system was not capable of audibly
supporting two conversations at the same time.
Accordingly, the capacity of the PA system should
be increased so that as many as 4 or 5 conver-
sations may be simultaneously effected. Also,
more outlets should be provided both in the control
room and at other work areas in the building.

30. A megawatt-hour meter should be instalied
(possibly on the log N or the AT recorder) in order
to provide a continuous measure of the integrated
power as the experiment progresses.

31. In order to analyze various related time-

dependent data (as required, for example, in the
determination of the various temperature coef-
ficients), it is necessary that the recorder charts
be marked at the start of the experiment. While
the charts can be marked by hand, this was a
source of error which became very important in the
analysis of data from fast transients. Accordingly,
all such charts should be periodically and simul-
taneously marked by an automatic stamper. Further-
more, each chart should be stamped with a dis-
tinctive mark that would positively identify it.

32. To analyze the kinetic behavior of a reactor
system it is necessary to have a recorder chart of
the time behavior of all equipment which can
introduce transients in the reactor, as well as
charts of the process data and conventional nuclear
data. Therefore the helium blower speeds and
shim rod positions should have been recorded
continually.

33. The recorded data and the data sheets
comprise a fairly comprehensive picture of the
experiment. However, even if all the dota are
recorded they are of value only to the extent that
meaningful interpretations can be made. While
automatically marking the data charts will be
helpful, it will still be necessary to rely on log
book information which should be recorded in
great detail, possibly to the extent of making this
the only responsibility of a shift ‘‘historian.”’

99-106

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Appendix B
SUMMARY OF DESIGN AND OPERATIONAL DATA

This appendix tabulates both the design and operational data pertinent to the aircraft reactor experi-
ment. Most of the design data were extracted from two design memoranda,'+2 although some values had to
be revised because of subsequent modifications in the design. In addition to the design data, the experi-
mental values of the various system parameters are included. All values obtained experimentally are
shown in italics and, for comparative purposes, are tabulated together with the corresponding design
values. There are, of course, numerous design numbers for which it was not possible to obtain experi-
mental numbers. The flow diagram of the experiment is presented in Fig. B.1. The values of temperature,
pressure, and flow given on this drawing are design values — not experimental values.

 

w. B. Cottrell, ARE Design Data, ORNL CF-53-12-9 (Dec. 1, 1953).
2y, B. Cottrell, ARE Design Data Supplement, ORNL CF-54-3-65 (March 2, 1954).

DESCRIPTION

1. The Reactor Experiment

 

Type of reactor
Neutron energy

Circulating fuel, solid moderator
Thermal and epithermal

Power (maximum} 2.5 Mw
Purpose Experimental
Design lifetime 1000 hr

Fuel

NaF-ZrF ,-UF, (53.09-40.73-6.18 mole %)

Moderator BeO

Reflector BeQ

Primary coolant The circulating fuel
Reflector coolant Sodium

Structural material Inconel

Test stand Concrete pits in Building 7503
Shield 7"/2 ft of concrete
Heat flow Fuel to helium to water

107

 
2. Physical Dimensions of Reactor {in.)

Cold Hot

 

 

(70°F) (1300°F)
Core height : 35.60 35.80
Core diameter 32.94 33.30 i
Side reflector height 35.60 35.80 .k
Side reflector inside diameter 32,94 33.30 X
Side reflector outside diameter 47.50 48.03 i
Top reflector thickness 4,00 4.25 ¥
Top reflector diameter 32.94 33.30 T
Bottom reflector thickness 4.93 4.98 E
Bottom reflector diometer 32.94 33.30 ;
Pressure shell inside diameter 48.00 | 48.55 i
Pressure shell wall thickness 2.00 2.02 E
Pressure shell inside height 44,50 45.00 N
Pressure shell head thickness 4.00 4.04
| Fuel elements 66 paralle| Inconel tubes containing £
3 the circulating fuel. The tubes :
! were connected in six parallel cir- ‘
cuits each having 11 tubes in

series. Each tube was 1,235 in.
0.D., with a 60-mil wall,

3. Volumes of Reactor Constituents (Cold)

a. Inside Pressure Shell

Volume (ft3) N

 

 

BeD Fuel Na Inconel Rods? Total

Core 14.55 1.33 0.80 0.37° 0.50 17.55
Side reflector 17.68 0 0.91 0.10¢ 0.267 18.95
Annulus outside reflector 0 0 0.58 0.19 0 0.77
Top reflector 0 0.21 1.62 0.08% 0.06 1.97
Bottom reflector 0 0.24 1.67 0.44° 0.07 2.42
Space above reflector and annulus 0 0 2.14 0.04¢ 0.034 2.21
Space below reflector and annulus 0 0 2.17 0.52¢ 0.04¢ 2.73

Total 32.23 1.78 9.8%9 1.74 0.96 46.60

Total volume inside inner rod sleeve.
blncludes the |Inconel-¢lad stainless steel rod sleeves.

€Includes all three sleeves around fission chambers

TP T e R gy BT P TR T TS roghgn g

4 ncludes volume of insulating material around inner fission chamber sleeve.
b. Operating Fuel System
Volume (ft3)

 

Cold Hot A
Core 1.33 1137
External system (to minimum pump level) 3.48 3.60
Pump (available above minimum level) ~ 1.65 ~ 1,70 -
Total 6.46 m

108

 
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B R B58 857
P B74
BLEED LINE T,
»le O FLOW IN gpm (F) rFueL ok RELIEF VALVE
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REFLECTOR COOLANT —==F FREEZE VALVE
rgvm o e !m i ’ PRESSURE IN psig
: 5 | FROM b SUPPLY TRAILER ] HELIUM () PUSH-BUTTON CONTROL
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“ - O -
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VALVE NORMALLY OPEN Zp_ ?OT:;?afplfaFoPERAnNG CONDITIONS
= Q. U
(g»—@ REMOTE MANUAL ANGLE VALVE
) —»4— VALVE NORMALLY CLOSED APPROXIMATE VOLUME OF MAIN SYSTEM:
- ¥ MANUAL VALVE INTERNAL 4,3 £1°
—»<— THROTTLING VALVE INITIAL~EXTERNAL 3.5 ft°
¥+ SOLENOID VALVE ENRICHING FLUID _ £.7 1> MAX.
—Jd—  CHECK VALVE TOTAL 6.5 13 MAX.

Fig. B.1. ARE Flow Di .
ig ow Diagram 109

 

 

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¢. Miscellaneous Fuel System

Volume
v (f+3)
Maximum capacity ffi"'afid flush tanks 14.5
Bottom (waste) in fill and flush tanks 0.3
Dump line to fill and flush tank Ne. 2 0.8
Dump line to fill and flush tank No. 1 Farpag - 075
By-pass leg - 0.4
Heat exchanger leg from pump to fill line to Tank No. 2 1.6
Reactor leg from fill line to pump (minimum operating level) 3.0
Pump capacity, minimum to maximum level ’ 1.7
Carrier available 15.5
Concentrate available 1.3
d. Sodium System
Volume
(f3)
Inside pressure shell 10
Qutside pressure shell ~10
Total 20
MATERIALS
Amounts of Critical Materials
BeO blocks (assuming p = 2.75 g/cm?) 5490 ib
BeO slabs (assuming p = 2.75 a/cm3) 48 Ib
Amount of uranium requested 253 b of U235
Uranium enrichment 93.4% Y233
Uranium in core : 301040 Ib
Uranium inventory in experiment 126 to 177 b

111

 

 

 
 

 

 

i

i E‘fi“i‘

2. Composition of Reactor Constituents

a. Fluoride Fuel Mixture

Fuel? Fuel Carrier W Fuel Concentrate

NaF

mole 53.09 50 66.7 N

wt % . 9:@ 34 20.1 21.3 ;
ZrF4

mole % 40.73 50 0 ~

wt % 62.12 79.9 0 )
UF 4

mole % 6.18 0 33.3

wt % 17.54 0 78.7
fmpurities (ppm)

Ni <56 25¢ 474

Fe 56 35¢ 844

Cr 4455 10¢ ~20%

“Fuel composition for high-power operation.
bEinal analysis before high-power operation.
“Average of all 14 batches of carrier.

4E stimate based on a few preliminary analyses.

b. Inconel?

. Amount i Amount i Amount :
Constituent (wt %) Constituent (wt %) Constituent (wi %)
Ni 78.5 Cob 0.2 Zrb 0.1 .
Cr 14.0 Alb 0.2 C 0.08
Fe 6.5 Ti® 0.2 Mo Trace
Mn 0.25 Ta® 0.5 Ag, B, Ba Trace
Si 0.25 Wwe 0.5 Be, Ca, Cd Trace
Cu 0.2 Zn® 0.2 Y, Sn, Mg Trace
“B. B. Betty and W. A. Mudge, Mechanical Engineering, February 1945,
bAccuracy, $£100%. Y-12 Isotope Analysis Methods L aboratory (spectrographic analysis). Y-12 Area Report Y- F20-14,
¢, Beryllium Oxide*
Impurities Amount Impurities Amount Impurities Amount
P {(ppm) P (ppm) P (ppm)
Si 1050 Ca 780 Na 330
Al 213 Fe 114 Mg 50
Pb 45 Zn <30 K <25
Ni <20 Cr 10 Li <5 s
Mn <5 B 2.8 Ag <1
Co <1

*W. K. Ergen, Activation of Impurities in BeO, Y-12 Area Report Y-F20-14 (Moy 1, 1951).

12

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e e

{

 
- ..xj..n.\'

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L

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T T e e e i

i ,Jfl_ L ek h R

R L

S

i, |

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et

 

e

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e, L G e BB ia L e

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Data fro‘m ANP Physical Properties Group.
bPreIiminary values for the liquid 600 te 800°C.

“Data from The Properties of Beryllium Oxide, BMIT-18 (Dec. 15, 1949).

dPorosify of BeQ, from ANP Ceramics Group, 23% at p=2.27, 0% at p = 2.83,

®Data from Metals Handbook, 1948 Ed., The American Society for Metals.

I Data from Liquid Metals Handbook, NAVEXOS P-733 (June 1952).
8B. 0. Newman, Physical Properties of Heat Transfer Fluids, GI-401 (Nov. 10, 1947).

d. Helium
Oxygen contamination <10 ppm
e. Sodium
Oxygen contamination <0.025 wt %
3. Physical Properties of Reactor Materials
Melting Thermal Vi i Heat Densi
Point  Conductivity |sco§|ty Capacity e/n5|1:;y
(°C)  (Btu/hr-f1.F) (cp (Btu/1b-°F) (9/cm”)
Fuel Carrier:*
NaF-ZrF, (NaZrF) 510 2.5 +5% 8.0 at 600°C  0.30% p =3.79 - 0.00093T
50-50 mole % 5.3 at 700°C 600 < T < 800°C
3.7 at 800°C
Fuel Concentrate:“
NaF-UF , (Na,UF ) 635 0.5 (estimated) 10.25 at 700°C  0.21% p =5.598 - 0.00119T
66.7-33.3 mole % 7.0 ot 800°C 600 < T < 800°C
5.1 at 900°C
Fuel:4
NaF-ZrF ,-UF, 530 1.3 (estimated) 8.5 at 600°C 0.24% p = 3.98 - 0.00093T
53.09-40.73-6.18 5.7 at 700°C 600 < T < 800°C
mole % 4.2 at 800°C
BeQ° 2570 16.7 at 1500°F 0.46 at 1100°F  from 2.27¢ to 2.83
19.1 at 1300°F 0.48 ot 1300°F  at 20°C
22.5 ot 1100°F 0.50 at 1500°F
Inconel® 1395 8.7 from 20 to 0.101 8.51 at 20°C
100°C 77 < T <212°F
10.8 at 400°C
13.1 at 800°C
Sodium/ 98 43.8 ot 300°C 0,38 at 250°C 0.30 0.85 at 400°C
38.6 at 500°C  0.27 at 400°C 0.82 at 500°C
0.18 at 700°C 0.78 at 700°C
Heliumé <-272.2 0.100 at 200°F 0.0267 at 200°C  1.248 1.79 x 1074 at 0°C
0.119 ot 400°F 0.0323 at 400°C 0 < T <300°C  1.30 x 107 at 100°C
0.136 at 600°F 0.0382 at 600°C 1.03 x 1074 at 200°C
Insulation
Superex (Si02) 0.18 0.38
Sponge felt (MgSiOs) 0.13 0.48

113

T

 
 

 

 

 

REACTOR PHYSICS

- 1. General
Neutron energy
Thermal fissions, %
Neutron flux, n/em?.sec
Gamma flux, y/cm?.sec
2. Power

Maximum power, Mw
Design power, Mw
Power density (core av at max power) w/cm®
Maximum specific power (av kw/kg of U233 at 1.5 Mw)
. Power ratio (max/av)
Axially
Radially in core
Radially in fuel tube
“Maximum power density in fuel, w/cm3
‘Maximum power density in moderator, w/cm?

3. Neutron Flux in Core (n/cm2Z.sec)

Thermal (max)
Thermal (av)
Fast (max)
Fast (av)

, Intermediate

4, Leakage Flux (per fission) |

Reflector
Fast
Intermediate

Thermal

Ends
Fast
Intermediate

Thermal

5. Fuel

Enrichment, % U235

Critical mass, Ib of U235 in clean core

U235 in core (ive., critical mass plus excess reactivity) Ib
U235 in system, Ib

U235 consumption at maximum power, g/day

6. Neutron Flux in Reflector

Maximum flux at fission chamber holes, nv/w
Counting rate of fission chambers, counts/sec.w

114;

Thermal and epithermal

60 - .
~1014 ?
~100] s
| 10! ?
» i:
1.5 (2.5) -
1.0 :
3 (5)
94 (153)
1.5: 1 ;
1.2:1 .
1.7:1 P
110 (180) |

1.3 (2.2)

“"r*'-m*"ww'fl P oo MT G S peRc COE TV mE e WM T o g T T

1.5 x 1013 . _
0.7 x 1013 :
3.5 x 1013
1.5 x 1013 P
2.0 x 1013
0.0011
0.066
0.191
0.013
0.102
0.0002
93.4
(32.75) :
30t040  (36) | b
126 to 177 (138) E
1.5 (2.5) :
1.5 x 106 ’

2 x 10°
R i B

!i 7. Reactivity Coefficients

, Effect

4 Thermal base

i

d

R Uranium mass (at & = 1)

i .

4 ,

: Sodium density

4 Moderator density

’f Inconel density

: Fuel temperature

.

3 Reactor temperature

vg-! .

. Sodium temperature

4 Moderator temperature

4 .

i . 1. Material

4

T Barytes concrete block in

i the reactor pit only

s Poured Portland concrete

i Barytes concrete block
2. Composition

i

4 Cement

! Concrete

4 Barytes

3 ‘

4 3. Relaxation Lengths (em)

; , ,

_. 1-Mev gamma rays

i

4 2.5-Mev gamma rays

b 7.0-Mev gamma rays

1 Fast neutrons (1 to 5 Mev)

4

j. *

i

4

4

5

*‘,’

 

Symbol Value
Ak/k —-0.011 from 68 to 1283°F
—0.009 from 1283 to 1672°F
Ak/k | |
0.25 (0.236)
(AM/M)
AkR/E
— ~0.05 from 90 to 100% p
(Ap/p)
Ak/E
— 0.5 from 95 to 100% p
(Ap/p)
Ak/k _
e - =0.17 from 100 to 140% p
(Ap/p)
(Ak/R)/°F (~9.8 x 1077)
(Ak/k)/°F ~5x1073 (-6.1x1077)
(AR/R)/°F (-5.88 x 1077)
(AR/E)/°F (1.1 % 107%)
SHIELDING
Thickness (in.) Density
(from source outward) (g/cm?)
12 3.3
18 2.3

Barytes Concrete

60 3.3

CaCO, +Si0, + Al 0,
Cement and aggregate of gravel
Cement and aggregate of chSO4

Portland Cement

4.14 4,62
6.72 6.86
9.46 14.0
8.2 11.0

115

 

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|

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il i R

 
 

 

 

4, Source Intensities

No. per Fission

Core gammas

Prompt 2.0
Fission-product decay 2.0
Capture gammas 0.94

Gamma flux outside of reactor thermal insulation per watt
2.0-Mev gammas
7.0-Mev gammas

Neutron flux outside of reactor thermal insulation per watt
Thermal to 250 kev
Fast to 250 kev

5. Activity in External Fuel Circuit per Watt

Time Qut of 0.52-Mev Gamma
Reactor Activity
(sec) (photons/cm3.sec)
0 8.1 x 106
5 6.6 x 108
10 5.9 x 10¢
20 5.1 x 10¢
30 4,7 x 106
40 4.5 x 108

REACTOR CONTROL

1. Control Elements

Source

Regulating rod

Shim rods

Temperature coefficient, (Ak/k)/°F

2. Source

Location

Type

Strength, curies

Neutron intensity, n/sec

3. Regulating Rod

Location Core axis
Diameter, in. : 2

Travel, in. 12

Cooling Helium

Total Ak/k, % 0.40 (0.40)
Maximum (Ak/k)/sec (slowspeed), % 0.010 (0.011)

Speed, in./sec

116

Energy (Mev)

2.5
2,5
7.0

3 x 104 y/cm?.sec
0.8 x 104 y/cm?.sec

3 % 10% n/cm2.sec
1.5 x 104 n/cm2.sec

1-Mev Delayed-Neutron
Activity
(n/cm3.sec)

8.5 x 103
2.3 x 103
1.4 x 103
0.74 x 108
0.50 x 103
0.35 x 103

) =d

~=5x 1075 (~6.1x107)

Core axis

Po-Be

15 (7)

3.5 x 107 (1.6 x 107)

0.3 or 3 (0.32) (fast speed not used)

 

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Backlash, in.
Servo motor
Actuation

Regulation

There were seven regulating rods made up whi
used in the experiment was the one which mos

Rod number 1 2
Calculated Ak/k 0.13 0.18
Measured AE/E

Weight per inch of rod, ib 0.042 0.06

4. Shim Rods (3)

Location

Diameter, in.

Travel, in.

Cooling

Material

Magnet release time, sec

Maximum withdrawal speed, in./sec
Total Ak/k per rod, %

Maximum (% Ak/k)/sec per rod
Motor

5. Nuclear |nstrumentation

0.007

Diehl 1A, 115 v, 60 cycle, 3400 rpm, reversible
Manual

Temperature error-signal (not used)

Flux signal, £ 1 2% N i S1.3°F mean T

ch gave a calculated Ak/k from 0.13 to 1.35%. The rod
t nearly gave a measured Ak/k of 0.4%.

3 4 5 6 7
0.27 0.39 0.56 0.90 1.35
(~0.25) (0.40)
] 0.091 0.132 0.21 0.32 0.50

One at each 120 deg on 7.5-in.-radius circle
2
36
Helium
B C
4
<1072
0.036 (0.046)
5.0 (5.8)
0.005, av over rod (0.0039 at 4-in. insertion)
Janette 1.2 amp, 115 v, 60 cycle, 1725 rpm, reversible

a. Fission Chambers (2)
Function Counting-rate signal
Sensitivity 0,14 (counts/sec) per (n/cm?.sec)
Range 1017, i.e., 104 in instrument, 107 in position and shielding

b. Parallel-Circular-Plate lonization Chambers (3)

Function (2 chambers)
(2 chambers)

Sensitivity

Range

Safety level (scram signal)
Regulating rod temperature servo
50 pa at 1019 n/cm?.sec

~ 103, i.e., from 5 x 1072 t0 1.5 Ng

c. Compensated lonization Chambers (2)

Function {1 chamber)
(1 chamber)

Sensitivity

Range

Micromicro ammeter
Regulating rod flux servo

50 pa at 1019 n/cm?.sec

~ 108, i.e., from 1076 to 3 Np

117

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T T PR T O PR T T I

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6. Scrams and Annunciators E
‘a. Automatic Rod Insertion and Annunciation &
‘ Nucl P Annunciat é
p ‘ ) uclear rocess nnunciator L
Cause Set Point - Reverse® ;
e ~ Scram Seram No. g
; Neutron level 1.2 and 1.5 NFb X 30 .
‘i Period 1 sec X 30 b
3 Period 5 sec X 32 .
4 Reactor exit fuel temperature > 1550°F x 25 ,
4 Heat exchanger exit fuel temperature < 1100°F x 25 =
4 &
{ Fuel flow <10 gpm X 25 A
; : - .
i Power Off X 30 ¥
| Seram switch Scram position X 30 i
:
| : b
4 %Shim rods automatically driven in. , .
. PNeutron level set so that the fast scram annunciater, No. 30, would annunciate at 1.2 Ng (normal flux) although |
the safety rods would not be dropped until the neutron level reached 1.5 Ng. There was not another neutron level b
annunciator at 1.5 Ng. ' &
b. Nuclear Annunciators
Set Point Annunciator No.
Safety circuit Electronic trouble 28 #
Count rate meter Off-scale 26 .
Servo Off-scale 31 |
Rod-cooling helium Off 29 .
Neutron level 1.2 Ng 27 :
!
¢
] i
e - E .
- '
£
1 §

118

 
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c. Fuel System Annunciators

Pump power {main pump)

Low heat exchanger fuel flow

High heat exchanger water temperature
Low heat exchanger fuel temperature
High fuel level (main pump)

Low fuel level (main pump)

High pump pressure (main pump)

High fuel reactor pressure

High fuel level (standby pump)

Low fuel level (standby pump)

High pump pressure (standby pump)
Pump power (standby pump)

High temperature differential across reactor tubes
High fuel temperature

Low heat exchanger fuel temperature
High heat exchanger tube temperature
Low heat exchanger tube temperature?
Pump lubricant (main pump)

Pump coolant (main pump)

Pump lubricant (standby pump)

Pump coolant {standby pump)

Set Point

>50 amp
<20 gpm
>160°F
<1150°F

Maximum pumping level
Minimum pumping level

>5 psi

>50 psi (>41 psi)

Maximum pumping level
Minimum pumping level

>5 psi
>50 amps

>235°F (> 400°F)

>1500°F
<1100°F
> 1500°F
<T150°F

<4 gpm (<2 gpm)
<2 gpm (<2 gpm)
<4 gpm (<2 gpm)
<2 gpm {<2 gpm)

de Sodium System Annunciators

Pump power (main pump)

LLow sodium flow (back loop)

High heat exchanger water temperature
Low heat exchanger sodium temperature
High sodium level {main pump)

Low sodium level (main pump)

High pump pressure (main pump)

Pump power (standby pump)

Low sodium flow (front loop)

High heat exchanger water temperature
Low heat exchanger sodium temperature
High sodium level (standby pump)

Low sodium level (standby pump)

High pump pressure (standby pump)
High sodium temperature differential across reactor
Low sodium reactor pressure

Pump lubricant (main pump)

Pump coolant {main pump)

Pump lubricant {standby pump)

Pump coolant (standby pump)

Set Point

> 50 amp
<100 gpm
>160°F
<1100°F

Maximum pumping level
Minimum pumping level

>65 psi
>50 amp
<100 gpm
>160°F
<T100°F

Maximum pumping level
Minimum pumping level

> 65 psi
>60°F
<60 psig
<4 gpm
<2 gpm
<4 gpm
<2 gpm

(>48 psig)

(<2 gpm
(<2 gpm
(<2 gpm
(<2 gpm

)
)
)
)

Annunciator

No.

33
34
35
36
37
38
39
41
45
46
47
48
49
51
52
53
54
65
66
69
70

Annunciator

No.

9
10
11
12
13
14
15
17
18
19
20
2]
22
23
24
42
67
68
71
72

“The helium blower was interlocked so that when the fuel temperature decreased below this temperature the biower
speed was reduced to zero until the fuel temperature exceeded 1150°F.

119

T gy e vy

=

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s ARk 4

 

Sodium system helium blower
Pit oxygen

Pit humidity

Helium supply low

Space cooler water temperature
Low water flow (any of 5 systems)
Rod cooling water temperature
Low reservoir level

Stack closed

Pit activity

Monitrons

Vent gas monitor

Vent header vacuum

LLow nitrogen supply

Low air supply

DC-to-AC motor-generator set
AC-to-DC motor-generator set

e. Miscellaneous Annunciators

Set Point

Sodium system blower on before fuel system blower
<90% He (Disconnected)

> 10% relative humidity

<500 psi

>160°F

> 10% below design

> 160°F

<16 psi

<5 mph wind velocity or high ion chamber reading
0.8 pc/em®

>12 me/he (>7.5 mr/br)

0.8 uc/cm®

>29 in. Hg vacuum

<300 psi

<40 psi

Off

Off

SYSTEM OPERATING CONDITIONS3

1. Reactor

Annunciator

No.

W NN WA —

55

 

HE o e o el

T "

*

T T "

e ey

k.
+
&
¥

 

P . - o

YR e

#

 

*After 462 hr of circulating time, the last 221 hr were attained after the reactor first bacame critical, and 74 of
these after the reactor first operated above 1 Mw.

 

CRPERE S, VR e wemE poocte

3The design values were taken from drawing A-3-0, ‘'Primary Heat Disposal System,’* Flow Sheet, in which the
physical properties of the fuel were assumed to be p = 3.27 g/em3, =7 to 13 cp, and cp =0.23 Btu/I1b:°F. The
operating values were taken from the 25-hr Xenon run (Exp. H-8). :

Fuel inlet temperature, °F 1315 (1209)
Fuel outlet temperature, °F 1480 (1522)
4 Mean fuel temperature, °F 1400 (1365) .

. Sodium inlet temperature, °F 1105 (122¢6)
Sodium outlet temperature, °F 1235 (1335)

._ Fuel flow through reactor (total), gpm 68 (46)

Fuel flow rate in fuel tubes (11.3 gpm), fps 4 (3) ;
Sodium flow through reactor, gpm 224 (150) E
Heat removed from fuel, kw 1270 (1520) k
'7 Heat removed from sodium, kw 650 (577) &
Fuel dwell time in reactor, sec 8.3 :
Fuel cycle time .
For 50% of fuel, sec 33.6 (47) £
4 For rest of fuel, sec 46.4 (47) .
i Maximum fuel tube temperature, °F 1493 |
E Maximum moderator temperature, °F 1530
E Sodium circulating time, hr 1000 (635)
Fuel circulating time,* hr 1000 (462)

1

120

 

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2, Fuel System

 

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“The water entered each heat exchanger at ambient temperature and was then dumped.

2There were two helium-to-water heat exchangers in parallel; total flow, 130 gpm.

a. Fuel Loop
F
Temperature Pressure Flow
(°F) (psig) (gpm)
Reactor outlet 1450 (1522) 46 40
Heat exchanger inlet 1450 (1522) 34 20¢ E
Heat exchanger outlet 1150 (1209} 12 20 f
Pump inlet 1150 (1209) 2 (0.3) 40  (4¢6)
Pump outlet 1150  (1209) 54 40 ‘
Reactor inlet 1150  (1209) 50 (39) 40 4
E.
%There were two fuel-to-helium heat exchangers in parallel.
b, Helium Loop
Temperature® Pressure Flow
(°F) (in. H,0) (cfm)
Fuel-to-helium heat exchanger No. 1 inlet 180 2.2 7,300
Fuel-to-helium heat exchanger No. 1 outlet 620 1.5 12,300
Helium-to-water heat exchanger No. 1 inlet 620 1.5 12,300
Helium-to-water heat exchanger No. 1 outlet 180 1.2 7,300
Fuel-to-helium heat exchanger No. 2 inlet 180 1.1 7,300
Fuel-to-helium heat exchanger No. 2 outlet 620 0.4 12,300
Helium-to-water heat exchanger No. 2 inlet 620 0.4 12,300
Helium-to-water heat exchanger No. 2 outlet 180 0.1 7,300
Blower outlet 180 2.3 7,300 |
“The helium passed through two temperature cycles in each loop. ;
c. Water Loop _,_
Temperature® Pressure Flow? t:'
(°F) (psig) (gpm) ;
b
Helium-to-water heat exchanger inlet 70 (61) 10 65 (103) é_
Helium-to-water heat exchanger outlet 135 (124) 8 65 1
E
!

3. Sodium System

a. Sodium Loop

 

Temperature Pressure Flow
(°F) (psig) (gpm)
Reactor outlet 1235 (1335) 58 224 (152) 1
Sodium-to-helium heat exchanger inlet 1235 (1335) 52 112 fi
Sodium-to-helium heat exchanger outlet 1105 (1226) 51 112
Pump inlet 1105 (1226) 48 (36) 224
Pump outlet 1105 (1226) 65 224
Reactor inlet 1105 (1226) 60 (49) 224

121

 
G Ly b B

 

 

b. Helium Loop

Temperature
(°F)
Sodium-to-helium heat exchanger inlet 170
Sodium-to-helium heat exchanger outlet 1020
Helium-to-water heat exchanger inlet 1020
Helium-to-water heat exchanger outlet 170
Blower outlet 170

¢, Water Loop

Temperature
(°F)
Helium-to-water heat exchanger inlet 70 (s61)
Helium-to-water heat exchanger outlet 100 (114)

4, Rod Cooling System

a. Helium Loop

Temperature
(°F)
Rod assembly outlet 240
Helium-to-water heat exchanger inlet 240
Helium-to-water heat exchanger outlet 110
Blower 110
Rod assembly inlet 110

4There were three heat exchangers in parallel.

Pressure

(in. HZO)

P ONN A

Pressure
(psig)

10
8

Pressure

(in. H,0)

Flow
(cfm)

2000
4700
4700
2000
2000

Flow
(gpm)

77 (38.3)
77

Flow
(cfm)

1270
4234
333

1000%

1000

bCopocity of each of the two parallel blowers; however, the second blower was in standby condition,

b. Water Loop

Temperature
(°F)
Water-to-helium heat exchanger inlet 70 (61)
Water-to-helium heat exchanger outlet 100 (63)

Pressure
(psig)

10
8

“The water entered edch heat exchanger at ambient temperature and was then dumped.

bTotal flow for three parallel heat exchangers.

5. Water System

Equipment Flow per Unit
(gpm)
Space coolers 7
Reflector coolant system 77
Rod cooling system 6
Fuel coolant system 65
Pump cooling systems 3

Total

122

No. of
Units

W N

56
154
18
130
12

370

Flow?
(gpm)

18 (17.6)
18

Flow
(gpm)

(56)
(77.6)
(17.6)
(206)
(13)

(370.2)

 

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1. Welding Specifications®

Process

Base metal

Position

Filler metal

Preparation of base metal
Cleaning fluid

Clearance in butt joints
Clearance in lap joints
Welding current
Electrode

Shielding gas blanket
Gas blanket beneath weld
Welding passes

Welders qualifications

MISCELLANEOQUS

Inert-gas, shielded-arc, d-c weld

Inconel

Horizontal rolled, fixed vertical, or horizontal

‘/16 to %. in. Inconel rod (Inco No. 62 satisfactory)
Machined, cleaned, and unstressed
Trichloroethylene

%, to % in. with 100 deg bevel

Flush at weld

d-c, electrode negative, 38 to 80 amp

]/16 to 3}32 in. tungsten, 90 deg point

Argon, 99.8% pure; 35 cth

Helium, 99.5% pure; 4 to 12 cfh

Ttoh

Welded or passed QB No, 1% within 45 days

e o e

— . b

2For details, see Procedure Specifications, PS-1, ORNL Metallurgy Division, September 1952,
bOperation Qualification Test Specifications, QTS-1, ORNL Metallurgy Division, September 1952.

2. Stress Anadlysis

a. Moments and Stress in the Pressure Shell Ends

Maximum radial stress (per psi pressure) 85 psi
Minimum radial stress (per psi pressure) ~80 psi
Maximum radial moment (per psi pressure) 80 in.-Ib
Minimum radial moment (per psi pressure) ~55 in.-ib
Maximum tangential stress (per psi pressure) 70 psi
Minimum tangential stress (per psi pressure) - 18 psi
Maximum tangential moment {per psi pressure) 180 in.-1b
Minimum tangential moment (per psi pressure) ~10 in.-lb

b, Stress in the Pressure Shell Vessel

Maximum circumferential stress (per psi pressure) 14 psi
Minimum circumferential stress (per psi pressure) ~23 psi
Maximum longitudinal stress (per psi pressure) 13 psi
Minimum longitudinal stress (per psi pressure) ~75 psi

T T T T I R g S T

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123

v

 
 

c. Summary of Pipe Stresses® F
. . ' Maximum Cold Maximum Hot 5 g
i Line Pipe Size Prestress Prestress Temperature Preslsure b
No. (in.) (Ib) (Ib) (°F) (psig) i
‘ Fuel Piping : z
111 2 3,850 2,170 1500 39 T
112 1, 11,230 6,450 1500 - 34 :
113 1%, 24,300 13,900 1400 ;34 :
114 1 15,400 8,850 1500 34 .
115 ] 38,000 19,600 1375 n ;
116 1%, 28,200 14,500 1325 60 -
117¢ 2 11 ‘
118 2 28,000 12,800 1500 i !
119 A 22,000 11,300 1325 60 :
| 120 2 15,350 7,900 1325 60 -
3 Sodium Piping 1 :
? 303 2), 5,260 3,690 1050 58
304 17, and 2 4,280 3,000 1050 58 :
305 1%, and 2 4,300 3,020 1050 58 :
306 1%,2,3 13,250 9,310 1050 51 )

307 2 4,170 2,935 1050 Es

308 1%,2,3 13,250 9,310 1050 ]

309 2 2,210 1,550 1050 65

310 2}, 4,430 3,110 1050 65 )
313 2 28,600 16,400 1325 51

2All piping ASI Schedule 40,
bNof all temperatures and pressures given were reactor design point values.

This line consisted primarily of pipe connections and was not stress cnalyzed,

3. Fluoride Pretreatment

Ti T
Treatment Purpose (LT)G em(;:)e(;;\fure
\
Hydrofluorination Remove water 2 RT to 700
Hydrogenation Reduce oxides and sulfates 2 700 to 800
Hydrofluorination Fluorinate oxides and sulfates 4 800
Hydrogenation Remove HF, NiF2 and FeF, 24 to 30

 

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124

 

 
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Appendix C
CONTROL SYSTEM DESCRIPTION AND OPERATION'

F. P. Green, Instrumentation and Controls Division

CONTROL SYSTEM DESIGN

The control system designed for the ARE had to
be able to maintain the reactor at any given power
and, when necessary, had to alsc be able to over-
come quickly the excess reactivity provided for
handling fuel depletion, fission-product poisons,
and power level increases. The motor speeds
and gear ratios were set so that safety-rod with-
drawal could not add Ak/k at a rate greater than
0.015%/sec. The rods were designed to contain
15% Ak/k, and the control system was designed to
limit the fast rate of regulating rod withdrawal to
be used for high-power operation to an equivalent
Ak/k of 0.1%/sec, with a maximum Ak/k of 0.40%,
which is one doliar, for steady-state fuel circu-
fation. For the critical experiment and low-power
operation, the regulating rod withdrawal rate was
limited to an equivalent Ak/k of 0.01%/sec.

A fundamental distinction was made between
insertion and withdrawal of the shim rods. Either
the operator or an automatic signal could insert
any number of rods at any time, whereas withdrawal
was always subject to both operator and interlock
permission. Withdrawal was limited, moreover, to
what was needed or safe under given circumstances,

Quantiiative, rather than qualitative, indication
of control parameters was used where possible.
Where feasible, only one procedure of manual con-
trol was mechanically permitted to minimize time-
consuming arbitrary decisions on the operator’s
part.

The control philosophy of paramount importance
in the general operating dynamics was the use of
only absorber rods to control the nuclear process
and the use of only helium coolant to control the
power generation.

Most of the control and safety features of the
ARE were similar to those of other reactors. Con-
siderable care was exercised in the design of
instrument components, control rods, and control
circuits to make them as nearly fail-safe as was
practical.  An instrumentation block diagram of
the reactor control system is shown in Fig. C.1,

 

Yhis appendix was originally issued as ORNL CF-53-
5-238, ARE Control System Design Criteria, F. P. Green
(May 18, 1953), and was revised for this report on
March 7, 1955,

in which the reactor is shown to be subject to
effects of the control rods and to certain auxiliary
facilities. The reactor, in turn, affected certain
instruments which produced information that was
transmitted to the operator and to the control sys-
tem. The operator and the control system also
received information from indicators of rod position
and motion. By means of the operator’s actions
and instrument signals, the control system trans-
mitted appropriate signals to the motors and
civtches. The actuators iocated over the reactor
pit in the actuator housing, in turn, affected the
reactor by corresponding rod motions which closed
the control loop. In addition to the above indi-
cations that were at the operator’s disposal on
the console, there were many annunciators physi-
cally located along the top of the vertical board
in the control room. The group of eight annunci-
ators labeled ""Nuclear Instruments’’ were provided
to indicate that conditions were improper for
raising the operating power level or that improper
operating procedures had caused safe limits to be
exceeded.

INSTRUMENTS DESCRIPTION

The reactor instruments discussed in the fol-
lowing have been arbitrarily limited to those di-
rectly concerned with the measurement of neutron
level and with fluid temperatures and flows associ-
ated with the reactor. The parallel-circular-plate
and compensated ion chambers, the fission cham-
bers, and BF, counter were of ORNL design,
and, except for the latter two, were identical to
those used in the MTR and LITR.? Since the
fission chambers were located in a high-tempera-
ture region within the reactor reflector, a special
chamber had to be developed that could be helium
cooled. The compensated ion chambers and the
log N and period meters had a useful range of
approximately 10%, and thus they indicated fluxes
of 10'% n/em?.sec (maximum) and 10% n/cm?.sec
(minimum}. The BF; counter had a maximum count-
ing rate of about 10° n/sec and the fission cham-
bers and count rate meters were capable of covering

 

25, H. Haenaver, E. R. Mann, and J. J. Stone, An Off-
On Servo for the ARE, ORNL CF-52-11-228 (Nov. 25,
1952).

125

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k.

 

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T RN R Yoy R e R

T

 
 

 

a range of 10'! if they were withdrawn from the
reflector as the flux increased. Since the count
rate meter has a long integrating time and the
resultant time delays would make it an unsatis-
factory instrument for automatic control, it was
used as an adjunct to manual control. |

The instantaneous positions of the four control
rods were reported to the operator by selsyns. In
addition, the positions of the rods with respect to
certain fixed mechanical limits were detected by
means of lever-operated microswitches. The limit-
switch signals tied in with the control system and
with signal lights on the control console. The
upper limit switches on the three shim rods oper-
ated when the rods were withdrawn to about 36 in.
from the fully inserted position. Their exact
location was adjusted to provide optimum sensi-
tivity of shim rod action. The lower limit switches
served to cut the rod drive circuits when the
magnet heads reached the magnet keepers.

The seat switches operated indicating lights on
the console, which served only to indicate the
proximity of the pneumatic shock absorber piston
to the lower spring shock absorber. The lights
told the operator the immediate effectiveness of a
““scrammed’’ or dropped rod and, hence, that the
rod had not jammed on its fall into the reactor
core. '

The regulating rod was equipped with two travel
limit switches whose actions tied in intimately

with the manual and automatic regimes of the
~ servo-control system, which is described in o
following section.

Thermocouples were located at many points on
the reactor and the heat exchangers to monitor the
fuel temperatures, since temperature extremes or
low flow rates could have resulted in serious
damage to the reactor. The temperatures were
recorded, and electrical contacts in the recorders
interlocked with the scram circuit to provide shut-
down if the reactor inlet temperature dropped below
1100°F or the outlet temperature exceeded 1550°F.
The operator was also warned by an annunciator
alarm of low helium pressure in the rod-coocling
system because loss of cooling would have in-
creased the danger of a rod jamming due to me-
chanical deformation.

CONTROLS DESCRIPTION

Motion of the shim rods or the regulating rod was
obtained from two energy sources, gravitational
and electromechanical. Rod motion by safety

126

action was obtained by gravitational force upon
release of the shim rods from their holders as a
result of de-energizing the electromagnetic clutches.

The three shim rods, each with its magnet, were
subject to a number of different possible sources
of safety signal. The necessary interconnecting
circuitry was centered about an ‘‘auction’’ or
"“sigma’’ bus. Instrument signals were fed to this
bus by sigma amplifiers. The sigma bus may be
said to go along with the ‘*highest bidder’’ among
the grid potentials of the sigma amplifiers. This
important auction effect allowed all rods to be
dropped by any one safety signal. [f all three
amplifiers and the three magnets had been identi-
cal in adjustment and operation, all three rods
would have dropped simultaneously as the sigma
bus potential rose or fell past some critical value.
Usually, only one of the rods dropped initially,
and the others were released by interlocked relay
action.

Two kinds of scrams were possible: ‘‘fast’
scrams caused by amplifier action and ‘‘slow’’
scrams caused by interruption of magnet power as
the result of the action of relays. In a fast scram
the shim rods were dropped when the reactor level
reached or exceeded a specified flux level, as
determined by the safety chambers. The signal
from the safety chambers was amplified by the
sigma amplifiers and subsequently caused the
magnet current to become low enough to drop the
rods. The slow scram was the actual interruption
of power to the magnet amplifiers by relay action.
This relay action could be caused by temperatures
in excess of safe limits or stoppage of fuel flow.

The shim rod drives were 3-wire, 115-v, instan-
taneously reversible, capacitor-run motors rated
at 1800 rpm. The shim-rod head, or actuator as-
sembly, was a worm gear driven through a gear
reducer at slightly less than 2?’/4 in./min.  The
sequence of shim rod scram, rod pick-up, and with-
drawal required a minimum of 25 min.

Both individual and group activation of shim rods
were selected in preference to group control alone
to permit increased (vernier) flexibility in ap-
proaching criticality. Group shim rod action would
have made difficult the control of flux shading
throughout the reactor, particularly on rod insertion,
since, upon cutting the motor power, the friction
characteristics of the three drives would have
caused the rods to coast to different levels.

The regulating rod drive consisted of two Diehl
Manufacturing Company, 200-w, instantly reversible,

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FOR START UP PLATE CHAMBER PLATE CHAMBER PLATE CHAMBER CHAMBER : CHAMBER r CHAMBER CHAMBER SCRAMS CHAMBER CHAMBER CHAMBER CHAMBER
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DRIVE DRIVE
RECORDER REGCORDER MECHANISM MECHANISM |
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AMPLIFIER AMPLIFIER
A~1 LINEAR SERVO FLux TROUBLE
AMPLIFIER FROM MICROMICROAMMETER l BATTERY SERVO MONITOR
MICROMICROAMMETER TEMPERATURE SIGMA SIGMA PERIOD DEMAND
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T TR T I T T e T - o

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two-phase servomotors that operated from reversing
contactors which received their actuation from a
mixer-pilot relay located in the serve amplifier.
Only one of the motors operated at a time, de-
pending on whether the temperature or flux servo
was coupled into the control loop. For the flux
servo operation, an additional speed reducer was
used so that the speed of the rod was one tenth
that available for temperature servo operation.

Initiation of the flux servo ‘'auto’ regime was
contingent on reactor power great enough to give
a micromicroammeter reading of more than 20
and a servo-amplifier error signal small enough so
that no control rod correction was called for. With
the servo on auto the regulating rod upper or lower
limit switches actuated an annunciator. Although
not used in the experiment, initiation of the tem-
perature servo auto regime was contingent on
reactor power greater than a minimum controllable
power (>2% N _) and a servo-amplifier error signal
small enough that the servo would not call for rod
motion.

The rod-cooling system consisted of a closed
loop for circulating. helium through the annuli
around the shim rods, the regulating rod, and the
fission chambers and then through three parallel
helium-to-water heat exchangers. The helium was
circulated by two 15-hp a-c motor-driven positive-
displacement blowers. The motors were controlled
from the reactor operating console.

The fuel and reflector coolant temperatures were
also controlled from the conscle. The helium
coolant removed heat from fuel and sodium system
heat exchangers and passed it to an open-cycle
helium-to-water heat exchanger, from which the
water was dumped. The helium in the fuel system
was circulated by an electric-motor-driven fan that
was controlled from the console. The helium in
the sodium system was circulated by two hydraulic-
motor-driven blowers for which only the on and off
controls were located on the console.

In order that the reactor operation would be as
free as possible from the effects of transient
disturbances or outages on the purchased-power
lines, a separate electrical power supply was
available for the control system. The supply
consisted of a 250-v, 225-amp, storage-battery bank
of 2-hr capacity. The bank was charged during
normal power source operation by a 125-kw motor-
generator set. This emergency d-¢ supply fed
emergency lights and a 25-kw motor-generator set
to supply instrumentation power (120/208 v, single

phase alternating current) during purchased-power

- outages.  An auto-transfer switch provided an

automatic switchover transient of several cycles
upon loss of the purchased-power source. The
auto-transfer switch automatically returned the
system to normal power 5 min after resumption of
purchased-power supply. The system could be
returned to normal power manually at any time after
resumption of the purchased-power supply.

CONSOLE AND CONTROL BOARD DESCRIPTION

The control console was made up of two large
panels, one to the right and one to the left of the
operator’s chair, and eight subpanels directly in
front of the operator, Fig. C.2. Over the top of the
console, the operator also had a full view of the
vertical board of instruments which indicated and
recorded nuclear parameters and pertinent process
temperatures. The rotary General Electric Company
type SB-1 switches used were of the center-idle,
two- and three-position variety. They were wired
so that clockwise rotation produced an ‘‘increase’’
in the controlled parameter; such operation has
been described as potentially dangerous. However,
the scram switch, an exception, produced a scram
in either direction of rotation.

The top row of the left-hand panel of the console
contained the source-drive control (Fig. C.2). In
the second row the first two switches were the
controls for the two-speed rod-cooling helium
blowers, and the third switch was the on-off con-
trol for the electric motors which were the prime
movers for the hydraulic-motor-driven blowers. In
the bottom row were the group control for the three
shim rods and the annunciator acknowledge and
reset pushbuttons.

The center eight subpanels were identical in size
and shape and contained the six selsyns which
indicated the rod and fission chamber positions
and the two servo control assemblies. The temper-
ature calibrated potentiometer located near the
panel center was used to set the fuel mean temper-
ature for automatic, servo-controlled operation.
The subpanel on the extreme left contained the
flux vernier (not shown) used to set flux demand
into the flux servo system for controlled operation.
The dial was calibrated to correspond to a setting
of from 20 to 100 on the micromicroammeter Brown
recorder. Associated with all controls were the
necessary limits-of-travel indicating lights, along
with the on-off and speed-range indicators.

The right-hand panel contained the switches most

129

 

TR Y

 
 

 

closely associated with the fuel system and the
energy removal controls. In the upper row were
the switch for the fuel-loop helium-blower prime
mover, which was a 50-hp electric motor, the scram
switch, and the by-pass switch (not shown on
Fig. C.2). The by-pass switch was provided to
permit the fuel-loop prime mover to be run, even

* though the reactor power was low, to remove the

afterheat that was expected to be present after a
shutdown. The bottom row on the right-hand panel
contained the fuel-system-helium blower speed
control, the selector switch for barrier-door control,
and the barrier-door drive control.

The four panels in the center of the vertical
board contained the 12 recorders which pertained
to the reactor. They were, left to right, top row:
reactor inlet temperature, reactor AT, reactor out-
let temperature, reacfor mean femperafljre; center
row: count rate No. 1, pile period, safety level
No. 1, and micromicroammeter. In the bottom row
there were: count rate No. 2, log N, safety level
No. 2, and control rod position.

The range of flux through which the reactor
passed from the insertion of the source to full-
power operation was so great, more than 10]3, that

several classes of instruments were used,
The normal flux (Np) was arbitrarily desig-
nated as 1, ond then the first, and lowest, range
encountered was the source range, from somewhat
less than 10='3 to 10~'); the second was the
counter range, from 10! to 10~%; the third was

the period range, from 10~% to 3.3 times 1073, and’

the last was the power range, from 3.3 x 10=3 to 1.
At greater than 10~3, in the period range, the servo
system could be put into auto operation if desired.

CONTROL OPERATIONS

In the operation of the ARE, operator initiative
was overridden by two categories of rod-inserting
action: scram, the dropping of safety rods; and
reverse, the simultaneous continuous insertion of
all three shim rods, which was operative only
during startup. The fast scram was in a class by
itself, being the ultimate safety protection of the
reactor, and could never be vetoed by the operator.
The various occasions for automatic rod insertion
and annunciation for nuclear trouble are listed in
Table C.1.

There were five interlocks between nuclear re-
actor control, fuel and moderator coolant pumping,

TABLE C.1. CAUSES OF AUTOMATIC ROD INSERTION AND ANNUNCIATION

 

 

 

 

Fast Scram Slow Scram Reverse Annunciation

 

Neutron level (1.5 NF)

Neufr_on level (1.2 NF)

leseoc period

5esec period

Fuel temperature (>1550°F)
Loss of fuel flow (power >10 kw)
Loss of control bus voltage

Loss of purchased rpower

Manual screm

Reac‘to'rrpower (AT > 400°F)
Fuel temperature (<1100°F)

Fuel helium blowers on without reflector helium blowers on
Servo c;ffjrahge | |
Rod coolant helium off

Count rate meter off scale

Safety circuit trouble

 

130

 

TR T

 
       

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131

 

 

 

 

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and helium-cooling operation. (1) When operating
at power (above 10 kw), loss of fuel flow would
have initiated a slow scram. (2) Loss of fuel flow
would have, likewise, removed a permissive on
operation of the fuel-loop helium-blower prime
mover. (3) Excessive deviation of fuel temperature
from set point would have produced a scram for
either too high or too low a temperature. Low fuel
temperature would also cause the helium pilot
motor to run the magnetic clutch control for the
helium blowers to the zero position and thereby
stop the blowers and thus stop fuel cooling. (4)
Moderator coolant flow was interlocked with the
moderator coolant helium blowers to prevent helium
circulation when the sodium flow was low or
stopped. A slug of cold sodium would have entered
the reactor when circulation was restarted if this
precaution had not been taken. (5) Permissives on
increasing helium flow were fuel temperature in

 

1. Two fission chamber drives

a. Manually actuated.

range, prime movers operating, reactor in power
range, and power greater than 10 kw. The operator
was notified by the ‘‘permit’’ light when these
conditions were satisfied. A pilot decrease was
not interruptible by the operator until the cause of
the automatic action was alleviated.

The elementary control diagram is shown in Fig.
C.3. The vdrious relay and limit switches referred
to in Fig. C.3 are described in Tables C.2 and
C.3, respectively. These tables and Fig. C.3 show
the procedural features of the control system as it
applied to operator initiative. This system was
divided into channels, each consisting of one or
more relays actuated by the operator subsequent
to properly setting up interlock permissives. The
interlocks often depended upon the aspects of
relays in other channels. Primarily, the channels
corresponded to a mechanical unit involving a rod
control, a process loop control, or an instrument
system control. The channels were:

b.

Travel limit switches, panel-light indicated.

2. Servo control system

Auto initiated manually by permission of power greater than 2% N, for temperature servo,

a
and power great enough to give a reading of more than 20 on the micromicrosmmeter recorder
for any scale range, for flux servo, and a démand error close to zero (neither light burning)
for either servo system.

b. Auto regime sealed-in light indicated if initiated with proper conditions and power remained
greater than that specified in 2a above.

¢. Manual regime regained by manually dropping out seal or by manipulation of control rod
switch.

d. Reverse interrupted rod withdrawal by any means and caused rod insertion.

e. Manual contro!l rod overrode automatic regime and dropped out seal.

/. Permits on manual rod control required that either the reactor to be at a power level greater

than ]0"5 NF or one count rate meter be ‘*on scale.”’

3. Slow scram

a.

b.

A series-parallel relay system was used to initiate a seram. This system was used because
a large number of series contacts would have been inducive to false drop-outs and hence
false scrams. The three most urgent criteria calling for scram had to be fail-safe and were
therefore placed in the series-connected section.

R-16 was inactive in the operating regime; R-17 and R-18 were normally actuated and were

responsive to manual scram, scram reset, fuel temperature extremes, and R-16.

4. Reverse, R-23 and R-24, plus an indicator light

d.

Operated ail three rods if the by-pass switch was in ‘*normal’’ and

(1) the group insert switch was closed, or

(2) a slow scram occurred (equivalent to a scram follower), or

(3) the servo system was in the manual regime; the reactor power was less than 10 kw; and
a period of less than 5 seconds occurred.

On initial startup, No. a(3) initiated the reverse at any power leve! (due to jumper-shorted

133

 

R

 

* TR T

 
 

é i
: {
] contact) until a negative temperature coefficient of reactivity was proved. :
b. The b&-puss switch prevented reverse (but not slow scram) vyhen in ‘*by-pass’® position.
5. Shim rod wifhdrdwal b
: d. Permissives were: ‘ ' . :
(1)} no reverse in progress, ' b
b (2) reading on-scale on at least one count rate meter or neutron level greater than 10-3
N | s
(3) no manual insertion in progress, :
(4) rod heads not resting on upper travel hmlts *
6. Shlm rod insert 2 i
a. Permitted if heads not on lower travel limits, z
b. Actuated individually and manually. 1
c. Actuated together by any action operating reverse. }
i
] 7. Fuel-system helium-blower prime movers ¢
a. The prime movers could be started by 5-13 if ;
: (1) clutch relays were actuated (rods '*hung''), or if by-pass switch $-5 was on by -pass, ;
. (2) helium blbwer speed lower limits were closed, and ¥
g ‘ (3) fuel flow was not below 20 gpm. ' ¥
:: b. The prime movers were automatically turned off if a shim rod was dropped or fuel flow }
dropped below 20 gpm. ! ’
c. Relay R-38 started the prime movers by activating the starter relay RF-1 located in the :

basement,

8. Helium pilot controls (power-loading system)
a. Permit indicators on power increase were lit if
(1) the reactor level was greater than 10 kw or the by-pass switch was set on by-pass,
" (2) the prime movers were running, and
(3) the fuel temperature was above 1100°F.
b. Power increase from the coolant system could be demanded if
(1) the permit lights were on and
(2) pilot switch, S-14, was thrown to *‘increase.’’
¢. If the power decreased and the controls were not already on shut-off limits, the helium
circulation would be shut off.
(1) manually by using switch $5-14, or
(2) automatically by whichever system suffered from low fuel temperature, from prime mover
cut-out, or from low power (<10 kw), since by-pass switch contact $5-3 was closed
during normal critical and power operation. Low fuel temperature interlocks automati-
cally decreased the corresponding helium pilot controls if the fuel dropped below the

critical temperature at either of the two heat exchangers.

9. The by-pass switch function was somewhat obscure since it occurred in the prime movers, pilot
controls, and reverse circuits. Its inclusion resulted from the possibility of requiring additional
fuel cooling following a shutdown and after an extensive running period at power level. A
large gamma heat source, primarily in the moderator (because of its very large heat capacity),
calied the a:fl'erliedt, remained in the reactor immediately following power operation.

By using both hands, the operator could throw and hold S5 on by-pass while he restarted the
prime movers and subsequently increased the helium pilot controls to such a point that the =

 

_ temperature leveled off. S . i
1 The by-pass circuits permitted testing of cooling system control circuits and presetting of
:i ' : cooling system control limits while the reactor was shut down and also permitted the vital .

subcritical test for fuel temperature coefficient of reacfivity.a

 

] 3Wm. B. Cottrell and J. H. Buck, ARE Hazards Summary Report, ORNL-1407 (Nov. 20, 1952).

134

 

 
S, o

iy, akide o

The use of the manual reverse cut-out, 55-4, operated by throwing the by-pass switch, was

divorced from the above considerations and devoted only to test operation of the shim rod

motors at shutdown, at low power, or with the magnet amplifiers turned off.

10. Source drive

a. Manually actuated.

b. Travel limit switches, panel-light indicated.

REACTOR OPERATION

The preliminary operating plan for the ARE was
described in ORNL-1844.% In the check operations,
considerable time was spent in loading the inert
fuel carrier and in the subsequent shakedown run.
Critical loading, subcritical measurement of fuel
temperature coefficient, regulating rod calibration,
zero-power operation at 1200 to 1300°F, and power
operation are described in Appendix D.

In the starting of a freshly loaded reactor the
operator is most concerned about a reliable neutron
signal. Attention must therefore be given to as-
suring that the fission chambers are in their most
sensitive positions, and that they are giving a
readable signal on the scalers or count rate meters.
[t is imperative that the source be inserted for the
earliest indication of subecritical multiplication to
be seen during the first startup. Use of the source
in subsequent experiments becomes less important
as the history of the reactor operation builds up.

For a normal, intentional shutdown the operator
had the choi¢e of a variety of procedures. For
every scram, however, the interlocks of the system
were such that the helium flow in the fuel-to-helium
heat exchangers was cut off by opening the electri-
cal circuits of the motors. This action was neces-
sary to prevent the fuel from freezing in the heat
exchangers. The various slow scrams included
the following: (1) maximum reactor outlet temper-
ature, (2) minimum reactor inlet temperature, (3) a
1-sec period, (4) N greater than 3.0 Mw, (5) fuel
flow below 20 gpm. A 5-sec period inserted the
shim rods, which decreased Ak/k at a rate of
0.015%/sgc, and operated only during low-level
startup.

In addition to the process signals which initiated
slow scrams, there were several other potential
process malfunctions for which it might have been
desirable to scram the reactor. However, the
action to be taken in these situations was left up
to the decision of the operator. There were two
reasons for not putting these process malfunctions

 

ADesign and Installation of the Aircraft Reactor Ex-
periment, ORNL-1844 (to be published).

 

on automatic scram. First, sensory signals usually
lag the event to such an extent that a fast scram
cannot provide better protection than a delayed
scram. Second, reaction to the signals requires
limited judgment. An annunciator alarm signal
was provided to draw the operator’s attention to
the possible difficulty for each case that would
have required limited judgment. The cases were
the following: (1) pronounced changes in differential
temperatures between reactor outlet tubes, (2)
lowering level in any surge tank, (3) rising level
in any surge tank, (4) alarm from a pit radiation
monitor or monitron, (5) high oxygen concentration
or humidity in the pits.

The operation of placing the ‘‘position’’-type
regulating rod serve system on automatic control
in any part of the power range above 2% N, for
temperature servo, and a micromicroammeter re-
corder reading of greater than 20 on any scale, for
flux servo, was arranged to require that the reactor
to be on a stable, infinite period. In the normal
course of events, the rod should have been manually
set at mid-range. Six d-c voitage signals were fed
into the servo amplifier, three from the servo ion
chamber, inlet temperature recorder, and outlet
temperature recorder, one from the voltage supply
adjusted by the ‘‘temperature demand’’ potentiome-
ter, and two from the flux demand and micromicro-
ammeter recorders. The operator therefore had a
unique temperature setting for each level of power
operation that produced *‘zero error’’ in the ampli-
fier output, and hence no ‘‘error’’-light indications
and no demand for rod movement. It was at this
setting that the servo ‘‘auto’’ regime could be
initiated. A small change in reactor mean temper-
ature or flux level could be effected in the auto
regime by resetting the temperature demand or flux
demand potentiometers, but the power level could
not be changed.

No clear line could be drawn between normal
operation and cperation under difficulty. A few of
the situations in which the operator would have
expected to feel more than average need for atten-
tion to instrument signals and control actions were
loss or interruption of power, automatic shutdown

135

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P T T e T R IRI EpT TY T Tw s e 1 S B e T g T T T T e m,m - T Tttt o vme o e j—

 

 
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TABLE C.2. LIST OF RELAY SWITCHES

 

 

Contact Instrument Set Point
RS-9 Count rate recorder No, 1 Closed above 1 count/sec
RS-10 Count rate recorder No. 1 Opened off scale

RS-11 Count rate recorder No. 2 Ciosed above 1 count/sec
RS.12 Count rate recorder No, 2 Opened off scale

RS.13 Front sodium flow Closed above 60 gpm
RS-14 Back sodium flow Closed above 60 gpm
RS-15 Log N Closed when N > 2% N
RS-16 Fuel heat exchanger No. 2 outlet Closed above 1150°F
RS-17 Fuel heat exchanger No. 2 outlet Closed above 1100°F
RS-18 Fuel flow Closed above 20 gpm
RS-19 Not used

RS-20 Log N Closed when N > 10™° N
RS-21 Log N Closed when N > 107> N
RS-22 Log N Closed when N > 6.67 x 1073 N ,
RS$-23a Safety level No. 1 Opened above 120

RS-23b6 Safety level No. 2 Opened above 120

RS-24 Fue! flow Opened above 20 gpm
RS.25 Not used

RS-26 Reactor outlet temperature (mixed manifold) Opened above 1550°F
RS.27 Not used

RS.28 Not used

RS-29 Pile period Closed at less than 5 sec
RS<30 Micromicroammeter Closed above 20 ppa
RS-31 Count rate recorder No, 1 Closed above 1 count/sec
RS-32 Count rate recorder No, 2 Closed above 1 count/sec
RS-33 Fuel heat exchanger No. 1 outlet Closed above 1150°F
RS-34 Fue! heat exchanger No, 1 outlet Closed above 1100°F
RS-35 Pile period Opened at less than 5 sec
RS-36 AT across reactor tube No. 1 Closed above 450°F
RS-37 AT across reactor tube No. 2 Closed above 450°F
RS-38 AT across reactor tube No. 3 Closed above 450°F
RS.39 AT across reactor tube No. 4 Closed above 450°F
RS-40 AT across reactor tube No. 5 Closed above 450°F

RS-41 AT across reactor tube No. 6 Closed above 450°F

RS.42 Not used

RS-43 Not used

139

 

—p—

T s D

 
TABLE C.2 (continued)

 

 

 

 

i

 

+

1

Contact Instrument Set Point
RS-44 Not used
RS.45 ACet0.DC M-G set Closed when set was charging batteries
RS-46 Not used
RS-47 Not used
RS-48 Not used
RS-49 Not used
RS-50 Fuel flow Closed above 20 gpm
RS.51 Fuel flow Closed above 20 gpm
RS-52 Reactor outlet temperature (mixed manifold) Opened above 1500°F
RS-53 Wind velocity recorder Wind above 5 mites/hr or vent gas monitors
RS-54 Fue! heat exchanger No. 1 outlet Closed above 1150°F
RS-55 Fuel heat exchanger No. 1 outlet Closed above 1100°F
RS«56 Fue! heat exchanger No. 2 outlet Closed above 1150°F
RS.57 Fuel heat exchanger No. 2 outlet Closed above 1100°F
RS-58 Main fue! pump low-water-flow alarm Closed above 4 gpm
RS-59 Fuel heat exchanger No. 2 outlet Closed above 1100°F
RS-60 Secondary fuel pump lowewater-flow alarm Closed above 4 gpm
RS-61 Back sodium heat exchanger outlet Closed above 1100°F
RS-62 Front sodium flow Closed above 100 gpm
RS-63 Back sodium flow Closed above 100 gpm
RS-64 Helium concentration recorder Closed at less than 90% He
RS«65 Pit humidity indicator Opened at > 10% relative humidity
RS-66 Helium supply header pressure Opened at <500 psig
RS~67 Water reservoir level Opened at <16 psig (<18,000 gal)
RS«68 Main sodium pump tank level Opened at low level
RS«69 Main sodium pump tank level Opened at high level
RS-70 Standby sodium pump tank level Opened at low level
RS-71 Standby sodium pump tank level Opened at high level
RS.72 Main fuel pump tank level Opened at low level
RS-73 Main fuel pump tank level Opened at high level
RS-74 Standby fuel pump tank level Opened at low level
RS.75 Standby fuel pump tank level Opened at high level
RS-76 Water controller for front sodium heat exchanger Opened above 160°F
RS.77 Water controller for back sodium heat exchanger Opened above 160°F
RS.78 Water controller for fuel heat exchanger Opened above 160°F -
RS.79 Standby sodium pump mator ammeter Opened above 50 amp
140

 

rrd rapt i 'r T v em rERemErr e e S mnye®r e w e P Omerremye B e T IR T e g Y T TR S e T o p o ST e ey et e et e e W e By

 
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TABLE C.2 (continued)

 

 

Contract Instrument Set Point
RS-80 Water controller for space coolers Opened above 160°F
RS-81 Water controller for rod-cooling heat exchanger Opened above 160°F
RS5-82 Individual heat exchanger tube temperature Opened above 1500°F
RS-83 Individual heat exchanger tube temperature Closed above 1150°F
RS-84 Main sodium pump motor ammeter Opened above 50 amp
RS-85 AT across reactor tube No, 1 Opened above 400°F
RS-86 Main fuel pump motor ammeter Opened above 50 amp
RS-87 AT across reactor tube No. 2 Opened above 400°F
RS-88 Standby fuel pump motor ammeter Opened above 50 amp
RS-89 AT across reactor tube No. 3 Opened above 400°F
RS-90 Standby sodium pump lowe-watersflow alarm Closed above 4 gpm
RS-91 AT across reactor tube No. 4 Opened above 400°F
RS.92 Main sodium pump low-water-flow alarm Closed above 4 gpm
R3-93 AT across reactor tube No. 5 Opened above 400°F
RS-94 Main fuel pump loweoil-flow alarm Closed above 2 gpm
RS-95 AT across reactor tube No, 6 Opened above 400°F
RS-96 Standby fuel pump lowsocil-flow alarm Closed above 2 gpm
RS-97 Standby sodium pump lowe-oil-flow alarm Closed above 2 gpm
RS-98 Main sodium pump loweoil-flow alarm Closed above 2 gpm
RS.99 Scdium outlet pressure at reactor Closed above 48 psig
RS-100 Fuel inlet pressure at reactor Opened above 48 psig
RS-101 Main sodium pump tank pressure Opened above 47 psig
RS-102 Standby sodium pump tank pressure Opened above 47 psig
R$-103 Standby fuel pump tank pressure Opened above 5 psig
RS-104 Main fuel pump tank pressure Opened above 5 psig
RS-105 Rod cooling helium pressure Closed above 13 psig
RS.106 Fuel loop water flow Closed above 110 gpm
 RS-107 Water flow to fuel and sodium pumps Closed above 10 gpm
RS-108 Rod cooling water flow Closed above 15 gpm
RS-109 Front sodium loop water flow Closed above 65 gpm
RS-110 Back sodium loop water flow Closed above 65 gpm
RS-111 Space cooler water Flow Closed above 45 gpm
RS-112 Vent header vacuum Opened below 29 in. Hg
RS-113 Reserve nitrogen header supply pressure Opened at <300 psig
RS-114 Rod cooling helium Opened at <40 psig
RS-115 DCsto-AC M-G set Closed when set was running

141

S R e S e A

 

- T = A pRm—— s

 
 

 

TABLE C.2 (continued)

 

 

 

 

 

Contact Instrument Set Point
RS-116 AC-to-DC M:G set Closed when set was running
RS-”? Heat exchanger pit radiation monitor Opened at high level
RS-118 Heat exchanger pit radiation monitor Opened at high level
RS-119 Vent gas monitor Opened at high level
RS-120 Vent gas monitor Opened at high leve!
RS-121 Sodium AT at reactor Opened at <60°F
TABLE C.3. LIST OF LIMIT SWITCHES
Contact Device Set Point
LS.1 Fission chamber No. 1 lower limit Closed on limit
L S.2 " Fission chamber No, 1 upper limit ‘Closed on limit
L S-3 Fission chamber No. 2 lower limit Closed on limit
LS-4 Fission chamber No. 2 upper limit Closed on limit
LS-5 Regulating rod lower limit Closed on limit
LS-6 Regulating rod upper limit Closed on limit
LS.7 Shim rod No., 1 upper limit Closed on limit
LS.8 Shim rod No. 2 upper limit Closed on limit
L S-9 Shim rod No. 3 upper limit Closed on limit
LS-10 Shim rod No. | lower |imit Closed on limit
LS-11 Shim rod No. 2 lower limit Closed on limit
LS.12 Shim rod No. 3 lower {imit Closed on limit
LS-13 | Shim rod No. 1 clutch Closed when rod was attached
LS-14 Shim rod No. 2 clutch Closed when rod was attached
LS.15 Shim rod No. 3 clutch Closed when rod was attached
LS.16 Shim rod No. 1 seat Closed on seat
LS-17 Shim rod No. 2 seat Closed on seat
LS-18 Shim rod No, 3 seat Closed on seat

 

action, and buildup of xenon poison after a
shutdown,

As far as loss of power is concerned, the operator
would have been left figuratively, and nearly
literally, in the dark were the instrument bus and
control bus to go dead simultaneously. In this
very unlikely occurrence, all the instrument lights
and all relay-indicating lights would have been
out, leaving the operator without visual knowledge
of whether the scram (automatic on loss of power)

142

was effective. If he had reasonable faith in the
law of gravity, he would probably have remained at
his post until power was restored. There would
have been no normal scram indication if the relay
cabinet main fuse were to have opened. However,
since the amplifier power cabinet would have
operated from emergency power, there was a check
on release of the shim rods in the fact that the
magnet currents would have been interrupted.
Whenever the reactor was subject to automatic

*

 

 

 
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shutdown action, the operator’s concern was di-
vided between the causative effects and the
chances of returning the reactor to normal oper-
ation. The circuits prevented the operator from
interfering with shutdown actions as long as the
causative conditions remained. Hence the oper-
ator's reaction to anything less drastic than a
scram was to clear the responsible condition.

When the reactor was scrammed, the chances of
getting back into normal operation were of course
dependent upon the time history of previous oper-
ation as well as upon the need for remedying the
trouble. However, any length of complete shutdown
could have been tolerated after prolonged operation
at normal flux, since xenon poison buildup was
low.

143

e

 

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b
}

 
i, Bega

 

Appendix D
NUCLEAR OPERATING PROCEDURES!

The importance and critical nature of the program, as well as the short time scheduled for the

operation of the reactor and system, necessitated a tight experimental program and precise

operating procedures. The nuclear operating procedures were, of course, prescribed in advance

of the experiment, It is of considerable interest to note that the actual experimental program

followed the anticipated program very closely tbrougbbut. The only significant deviations were

the inclusion of some additional experiments as time permitted. The following appendix is a

verbatim copy of the procedures by which the reactor was operated, Some minotr discrepancies

in opetating conditions {i.e., 34-gpm fuel flow vs 46.gpm actual flow) and procedures (i.e., use

of the original enrichment system, which was replaced) will be noted.

It is anticipated that the reactor and its associ-

ated circuits will be operated for about 250 hr
with fused salts in the system prior to the intro-
duction of fuel. All process instrumentation and
components (non-nuclear) will be checked out
during this period. The mechanical operation of
the nuclear equipment will be checked out at
temperature. The safety rods will be raised and
dropped approximately 50 times during this period.

Several tests preliminary to the fuel loading
will have been run. The rated flow of the fuel
carrier will be 34 gpm at a mean temperature of
1300°F,

1. The fuel loading system will be operated with
carrier. The fuel storage tank will be installed
and carrier forced into the transfer tank and then
into the reactor system. The strain gage weighing
devices will be checked.

2. The helium blowers will be operated and the
resultant drop in mean temperature observed. The
operating crew will practice in handling the helium
flow so as to drop the mean temperature at a rate
of 10°F/min and at a rate of 25°F/min. Care will
be exercised not to drop the temperature of the
fuel carrier to below 1150°F as it leaves the heat
exchangers. As the latter temperature approaches
that value, the helium flow will be reduced and
the system brought back to its mean operating
temperature of 1300°F.

3. With the helium blowers off and starting with
fuel carrier at rated flow of 34 gpm and mean
temperature of 1300°F, the flow rate will be
decreased in steps to about one-third rated value.
All temperatures will be observed to note any
spurious changes caused by the decreased flow.

 

 

]This appendix was originally issued as ORNL
CF-54-7-144, ARE Operating Procedures, Part I,
Nuclear Operation, J. L. Meem (July 27, 1954),

144

ADDITION OF FUEL CONCENTRATE?

The fuel concentrate will be added in batches
to an ‘‘eversafe’’ container which will contain

115 kg of U235, The density of the fuel concen-
trate may be represented by the equation:

p (g/em3) = 5.51 - 0.0013 T (°C)

At a temperature of 1300°F, the density is
4.59 g/cm?, or 287 Ib/ft>3. In each pound of the
fuel concentrate there is 0.556 1b of U235,

One quart of the concentrate weighing 9.59 Ib
and containing 5.33 Ib of U235 will be forced into
the transfer tank and weighed. With the safety
rods completely withdrawn and the neutron source
and fission chambers inserted, the fuel concen-
trate will be forced into the fuel circuit, At this
time the carrier flow will be 34 gpm and the
concentrate and carrier temperatures will be
1300°F. The scram level on the safety chambers
will have been set at about 10 kw and the period
scram at 1 sec. A BF, counter will have been
installed temporarily in place of the neutron ion
chamber for the temperature servo control, and the
regulating rod will be on slow-speed drive.

The system volume up to the minimum operating
level in the fuel pump is calculated to be 4.64 f3.
Assuming a critical mass of 30 Ib in the 1.3 f13 of
reactor core, there will be approximately 124 1b of
U235 jn the fuel circuit when the reactor first goes
critical with the shim rods complietely withdrawn.
Accordingly, it is anticipated that approximately
0.78 13 or 23.4 quarts of fuel concentrate must
be added for initial criticality. Twelve quarts will
be added in succession with the shim rods com-
pletely withdrawn and the fission chambers fully

 

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was not employed, although the principles and tech-
niques of enrichment outiined here were followed.

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inserted. After each quart of concentrate is added,
counts will be taken on both fission chambers and
the BF, counter. The reciprocal counting rate
will be plotted vs. the fuel concentration to ob-
serve the approach to criticality,

After 50% of the fuel concentrate (12 gquarts) has
been added to the system, a sample of the mixed
fuel and carrier will be withdrawn for chemical
analysis.

SUBCRITICAL EXPERIMENTS

During the addition of the first 12 quarts of fuel
concentrate, the safety rods have been completely
withdrawn from the reactor. From here on, the shim
rods will be inserted approximately 25% while a
quart of fuel concentrate is being added. The shim
rods will then be withdrawn and a count taken on
the fission chambers as before. The shim rods

will be inserted and withdrawn in this fashion for
each succeeding fuel addition. The reciprocal of
the counting rate vs. fuel concentration will still
be plotted to indicate the increase in criticality.

When approximately 90% of the critical mass has
been added, the fuel addition will be stopped and
several subcritical experiments performed. For
these experiments it is desired that the £ of the
reactor be about 0.97 to0 0.98, In Fig. D.1is shown
the relationship between Ak/k and AM/M as a
function of the critical mass M. Using this curve,
an estimate of the point at which fuel addition
must be stopped can be made.

For the first experiment, the fuel temperature
will be decreased at a rate of 10°F /min by starting
the helium flow. Assuming a temperature coef-
ficient of =5 x 10~ {(Ak/%)/°F, the resultant A%/
should be about 0.25% after 5 min. ¥ AM/M is

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

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Fig. D.1. Reactivity-Mass Ratio as a Function of Critical Mass.

145

 

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approximately 4 Ak/k {Fig. D.1), the count rate
should increase by an amount corresponding to the
addition of 1.0% more fuel. If the change in count
rate is too small to be definite, the experiment
will be repeated at a rate of decrease of 25°F /min
until a definite increase in count rate is observed,
or until the temperature has been decreased 100°F,
if, after this decrease in temperature, no increase
in count rate is observed, it will be assumed that
the temperature coefficient is negligibly small and
the experiment will proceed. As has always been

0.008

the philosophy on the ARE, if the temperature
coefficient is observed to be positive, the experi-
ment will be concluded and the fuel circuit drained.

For the second experiment, the reactor tempera-
ture will be returned to its original value of
1300°F, and with the reactor containing 90% of
the critical mass, the fuel flow rate will be gradu-
ally decreased. If the available delayed neutron
fraction is 0.47% Ak at full flow and 0.75% Ak
when the flow is stopped (Fig. D.2), a Ak of
0.28% should appear. The count rate should in-

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FUEL FLOW RATE {(gpm)

Fig. D.2. Reactivity from Delayed Neutrons as a Function of Fuel Flow.

146

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crease by an amount equivalent to the AM corre-
sponding to additional Ak in the delayed neutrons.
A rough experimental check on the calculated
value of the delayed neutron fraction will thus
be available.

During the period of subcritical operation, the
count rate will be carefully observed for sudden
changes that could be caused by fuel segregation.
No such difficulty is anticipated, but if such an
effect is observed, the reactor will be shut down
until the reason for such behavior is ascertained.

INITIAL CRITICALITY

Upon completion of the subcritical experiments,
the fuel flow and reactor temperature will be
returned to the initial conditions of 34 gpm® and
1300°F.  Counts will be taken on the fission
chambers with the shim rods 25% inserted and
completely withdrawn. The shim rods will then
be 50% inserted and a quart of fuel concentrate
added. The rods will be withdrawn to 25% and
a count taken and then completely withdrawn and
a count taken. From here on, two curves of re-
ciprocal counting rate vs. fuel concentration will
be plotted, one at 25% rod insertion and the original
curve with the rods completely withdrawn. If the
reactor contained 90% of the critical mass during
the subcritical experiments, about two l-quart
additions of fuel concentrate will bring the reactor
critical with the rods completely withdrawn. The
last fuel additions will be made very slowly. When
criticality has been definitely reached, the reactor
will be shut down and a fuel sample taken for
chemical analysis.

After the sample has been taken, the reactor will
again be brought barely critical (a few hundredths
of a watt) and held at constant power by watching
the count rate meters. The gamma-ray dosage will
be measured with a ‘“‘Cutie-pie’’ at all pertinent
points throughout the pits and recorded for future
reference on radiation damage and shielding.

ROD CALIBRATION vs FUEL ADDITION

The reactor will be brought critical at a very
low power and the shim rods adjusted so that the
regulating rod is 5 in. above center. The regu-
lating rod is estimated to be worth about 0.04%
Ak/in. One quart of fuel is approximately 4%
AM or 1% Ak (Fig. D.1). Therefore, ]/25 of a quart
should be worth about 1 in, of regulating rod.

 

3Actuai, 46 gpm.

Figures D.3 and D.4 show a calibration of a
regulating rod taken on a mockup of the ARE at
the Critical Experiment Facility. While it would
be fortuitous if the regulating rod in the actual
ARE gave the same calibration, it is expected
that the general shape of the curves will be the
same,

With the shim rods in a fixed position, the
regulating rod will be fully inserted and approxi-
mately ]/25 of a quart of fuel concentrate will be
added slowly to the system. The reactor will
be brought critical on the regulating rod and its
new position noted. This procedure will be
repeated until the regulating rod has been cali-
brated from 5 in, above to 5 in. below its mid-
position,

The above experiment will have been run with
the shim rods almost completely withdrawn, The
shims will now be inserted about 25% and ]/2 quart
of fuel added (approximately 0.5% Ak), The shim
rods will be withdrawn until the reactor goes
critical.  This procedure will be continued to
give a rough calibration of shim rod position vs
fuel addition over one-fourth of the rod. When
the reactor is critical with the shims approximately
15% inserted, the shims will be adjusted so that
the regulating rod is 5 in. above center, and a
second calibration of the regulating rod vs fuel
addition will be run as above,

Calibration of the shim rods vs fuel addition
will then be continued until the reactor is critical
with the rods about 25% inserted. At this time,
a third and final calibration of the regulating
rod vs fuel addition will be run. The reactor will
then be shut down and a fuel sample taken,

LOW-POWER EXPERIMENTS

The nominal power of the reactor is obtained
as follows: It is estimated that in the reflector
at the mid-plane of the reactor the fission pro-
ducing flux is 2 x 10% nv/w, and the average flux
over the length of the fission chamber when fully
inserted is 1.5 x 10% nv/w. Preliminary tests
on the fission chambers show a counting efficiency
of approximately 0.14 counts/sec.nv. The re-
lationship between counting rate and power is
therefore approximately 2 x 10° counts/sec = 1 w.
The nominal power of the reactor will be based
on this relationship during the preceding experi-
ments,

Until this time the reactor has been operated
on the fission chambers alone at a nominal power

147

 

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ROD POSITION {in.}

Fig. D.3. Regulating Rod Calibration (Rod B).

of about 0.01 w (about 2000 counts/sec). The
reactor power will now be increased to about
1 to 10 w, at which time the neutron ionization
chambers should begin to give readings on the
log N recorder, the micromicroammeter, and the
period recorder. The reactor will be leveled out
and the flux servo turned on. One of the fission
chambers will be completely withdrawn, which
should reduce its counting rate by several orders
of magnitude, A careful comparison of the new
count rate vs the old count rate will be made.

The reactor will be held at constant power for
exactly 1 hr and then scrammed. A fuel sample
will be drawn off and sent to the Bulk Shielding

Facility for determining its activity. This activity

will be compared with that of a previous sample
(Fig. D.5), which was exposed in a known neutron

148

flux in the Bulk Shielding Reactor. From the
comparison, the fission rate or absolute power
of the ARE will be determined for this run (cf.,
app. H). All neutron level instruments will be
calibrated accordingly.

Rod Calibration vs Period. A family of curves
of reactivity vs reactor period (inhour curves)
with the rate of flow of the fuel as a parameter
is shown in Fig. D.6.

With the fuel flow rate at 34 gpm, the servo will
be turned off and the reactor placed on infinite
period manually with the regulating rod at mid-
position. From the rod calibration vs fuel plot
and the theoretical inhour curve, the regulating
rod withdrawal for a 30-sec period will be esti-
mated (present estimate is about 1 in,)., The rod
will be withdrawn accordingly and the neutron

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ROD POSITION {in.)

Fig. D.4. Regulating Rod Sensitivity (Rod B).

level allowed to rise about two orders of magni-
tude, whereupon the regulating rod will be fully
inserted and the induced gamma rays allowed to
decay for about 20 min., The reactor will then
be brought back to its original power, as de-
termined by the fission chambers. If the position
of the regulating rod does not return to its original
value because of photoneutrons, the reactor power
will again be reduced by insertion of the regulating
rod until the photoneutron effect becomes negli-
gible. The above procedures will be repeated for
20-sec and 10-sec periods. The 10-sec period
should correspond roughly to a 2-in. withdrawal
of the regulating rod. Runs will then be made at

correspondingly longer periods until the period

for a Y%-in. withdrawal of the rod is obtained
(approximately 100 sec). At this time a repeat
run on the 30-sec period will be made to ascertain
whether photoneutron buildup is causing appreci-
able error in the measurements.

The shims will then be adjusted so that the

regulating rod is 1in. below center at infinite
period. The rod will be withdrawn 1 in. and the
period recorded. This procedure will be repeated
at successive starting positions of the regulating
rod 1in. apart from 5 in. below center to 5 in.
above., A check on the initial 30-sec run with
the rod withdrawn from the mid-position will be
made periodically to ensure that photoneutron
buildup is not interfering with the measurements.
From the standpoints of safety and experimental
convenience, the regulating rod should be worth
between 0.3% and 0.5% Ak for its full 12-in.
travel. If these experiments show that the value
of the rod is not in this range, a new rod will be
installed ot this time and the period calibration
repeated. |f convenient, a fuel sample will be
taken at this time. _
Measurement of the Delayed Neutron Fraction.
The delayed neutron fraction for a stationary fuel
reactor is about 0.73%. For the ARE ot a design
flow of 34 gpm, the fraction is calculated to be

149

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COUNTING RATE

ORNL-LR-DWG 1946

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FUEL CAPSULE, FUEL 44

CARRIER CAPSULE, FUEL 45

0 2 4 6 8 10

 

i2 14 1A i8 20 22 24

TIME AFTER SHUTDOWN ({ hours)

Fig. D.5. Decay Curves for Fluoride Activated in the Bulk Shielding Reactor.

0.47%, and for smaller flow rates, the corre-
sponding reactivity may be found from Fig, D.2.

Starting with the reactor stabilized at 34 gpm
and the flux servo on, the flow rate will gradually
be reduced. The calibrated regulating rod should
be inserted so as to indicate the same reactivity
as shown in Fig. D.2. This will give experimental
verification of the previously calculated inhour
curves, Fig. D.6.

Preliminary Measurements of Temperature Coef-
ficient, With the reactor held at 10 w by the flux
servo, the helium flow will be turned on so as to
drop the fuel temperature at a rate of 10°F/min.
As the mean reactor temperature drops, the servo
will insert the regulating rod so as to maintain
constant flux, and from the rod calibration, the
corresponding Ak can be obtained. Before the
rod has reached the limit of its travel, the helium

150

flow will be shut and the fuel allowed to return
to its original temperature. From the recordings
of rod position and mean temperature, a plot of
Ak vs mean temperature can be obtained. The
initial slope of this curve will correspond to the
fuel temperature coefficient. Because of the weak
signal received by the flux servo, this measure-
ment will be only approximate and is to be
repeated at higher power.

Depending upon how the reactor responds, the
procedure can be repeated by dropping the fuel
temperature at rates of up to 25°F /min, Care will
be taken not to drop the fuel temperature so low
as to set off the low-temperature scram.

APPROACH TO POWER

At the conclusion of the above experiments and
before the fuel storage tank is removed from the

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0 10 20 30 40 50 60 70 80 90 100 10 120

REACTOR PERIOD (sec}

Fig. D.6. Reactivity as a Function of Reactor Period for Several Fuel Flow Rates.

pits, sufficient fuel concentrate will be added to
the system so that 4% excess reactivity is avail-
able., This excess reactivity will be absorbed
with the shim rods. The shim rods will have a
rough calibration by this time, and it is expected
that they will be worth about 12% in excess
reactivity, Therefore, they will need to be in-
serted about one-third of their length.  After
addition of the final amount of fuel concentrate
the liquid level in the pump will be checked to
ensure that at least 0.3 ft3 of volume remains for
expansion of the fuel. This expansion volume
will allow the fuel to expand isothermally from
1300 to 1600°F, which is well above the high-
temperature scram level. The final sample of fuel
will be taken for chemical analysis.

- The reactor will now be shut down so that the
concrete block shields can be put over the pits
and the pits flooded with helium. The fuel storage
tank will first be removed and a final check on

all process equipment made. The BF, counter
will be removed and the temperature servo chamber
installed. The regulating rod will be put on fast
drive. A check list is being prepared of all items
to be reviewed before the pits are sealed.

[f the temperature coefficient is of sufficient
magnitude to control the reactor, the flux servo
will be left in, However, if the magnitude of the
temperature coefficient is marginal, the flux servo
will be removed at this time and the temperature
servo connected.

After the pits have been sealed, the reactor will
be brought to a power of 1 kw and allowed to
stabilize. Up until this time all neutron ion
chambers have been fully inserted. The ion
chambers will be withdrawn slowly, one by one,
while the power level is being constantly moni-
tored with the fission chambers. The ion chambers
will be set for a maximum reading of around 5 Mw,

151

 

 
 

 

The safety chambers will be set to scram at the
same level, -

The reactor is now ready for full power oper-
ation. The rod will be withdrawn and the reactor
put on about a 50-sec period, Somewhere in the
region from 10 to 100 kw a noticeable increase
in reactor period and an increase in reactor temper-
atures should be observed. The helium flow will
be started slowly and gradually increased until
a nominal power of 100 to 200 kw is reached
(AT from 25 to 50°F).

At this time, the reactor will be allowed to
stabilize for about ¥ hr and all readings recorded.
If the temperature coefficient is of insufficient
magnitude to stabilize the reactor, as determined
by previous measurements, the control of the
reactor will be turned over to the temperature
servo. ' ‘

The power as determined from the heat exfraction
by the helium will now be measured. The heat
capacity of the fuel is 0.23 Btu/Ib.-°F and the
density of the fuel is represented by

' (g/cm?) = 4.04 — 0.0011 T (°C) .

At a mean temperature of 1300°F, the average
density is 3.27 g/cm3. The power can be ex-
pressed as

p (kw) = 0.11 x AT (°F) x flow (gpm) .

Therefore at 34 gpm the power extracted by the
fuel is 3.74 kw/°F.

Having obtained o heat balance, the extraction
of heat from the fuel in the heat exchanger will
be increased until a power of about 500 kw
(AT = 134°F) is reached. Again the reactor will
be allowed to stabilize and the extracted power
measured. A third and final heat balance will be
made at a power of about 1 Mw (AT = 268°F).
This is close to the maximum power obtainable
without changing the initial conditions of 1300°F
mean reactor temperature and 34-gpm fuel flow.

During the preceding discussion, no mention has
been made of heat extraction other than in the
fuel circuit. An appreciable quantity of heat will
be removed by the sodium in the reflector coolant
circuit. Before the reactor goes to high power,
the sodium inlet and outlet temperatures will be
near the isothermal temperature of 1300°F. Since
there is some gamma heating in the reflector
region, it will be necessary to lower the inlet
temperature of the sodium by extracting heat with
the reflector coolant heat exchangers. The sodium
flow rate will be held at 224 gpm and the sodium

152

mean temperature at 1300°F. Since the heat
capacity of the sodium is 0.30 Btu/lb.°F and its
specific gravity is 0.78, the power extracted by
the sodium will be 7.7 kw/°F. No more than 10%
of the power generated is expected to go into the
sodium,

EXPERIMENTS AT POWER

Measurement of the Temperature Coefficients,
After the reactor power has been calibrated, the
power will be reduced to about 500 kw and allowed
to stabilize. The flux servo will be turned on,
and by means of the helium demand the mean
temperature will be dropped at an initial rate of
10°F /min. (lf the temperature servo is connected,
it will have to be replaced for this experiment.)
As discussed previously the initial slope of the
plot of Ak vs mean temperature will represent the
fuel temperature coefficient,  Contrary to the
procedure at low power, however, the helium flow
will not be decreased after the temperature has
dropped. The reactor will be allowed to stabitize,
and, after about 30 min, the regulating rod will
have leveled out at a new position, and the reactor
will have assumed a new mean temperature. These
readings will represent the over-all reactor temper-
ature coefficient (fuel plus moderator). If, during
the above experiment, the servo inserts the rod
to its limit, it will be left at that position since
the temperature should stabilize the reactor.

[f the fuel temperature coefficient is quite small,
the experiment can be repeated with an initial
rate of temperature decrease of 25°F/min. Care
will be taken that the reactor does not go on too
fast a period and that the low-temperature scram
fimit is not exceeded.

Maximum Power Extraction. Except for specifi-
cally stated instances, the reactor has been
operated continuously up to this time at a mean
temperature of 1300°F and a fuel flow rate of
34 gpm. Under these conditions, the maximum
obtainable power is 1.1 Mw. At this power, the
reactor outlet temperature and the heat exchanger
inlet temperature are both 1450°F. The heat
exchanger outlet and the reactor inlet are both
at 1150°F. It is to be noted that when the heat
exchanger outlet temperature drops below 1150°F,
an alarm is sounded.

The mean reactor temperature will be elevated
from 1300 to 1325°F by a slight withdrawal of the
shim rods. The mean temperature of both fuel
and sodium will be held at this temperature. The
helium blower speed will now be increased until

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a AT of nearly 350°F appears across the reactor.
With this AT, the reactor outlet and heat exchanger
inlet are at the upper limit of 1500°F, and the
heat exchanger outlet and reactor inlet are at the
lower limit of 1150°F. The power from the fuel
will be 1.3 Mw, At this time the pump speed can
be increased from 34 gpm to about 40 gpm, caution
being taken that the reactor inlet pressure does
not exceed 50 psig. The maximum AT across the
reactor will still be 350°F, and the power ex-
tracted in the fuel circuit will be 1.5 Mw. This
is the maximum power at which the reactor can
be operated.

Power Transients. With the reactor on manual
control and the regulating rod at mid-position, the
helium blower speed will be regulated until the
power in the fuel system is 1 Mw, and the reactor
will be allowed to stabilize. The helium flow
will be suddenly increased to its maximum and
the transient response of all temperature and
nuclear recordings noted. The experiment can be
repeated from successively decreasing initial
powers of 500, 200, 100, 50, 20, and 10 kw. At
some level below 100 kw, the power cannot be
reduced further because the helium blowers will
be completely shut off. When this lower limit of
initial power has been reached, the series of
experiments will be concluded. !f at any time
the reactor period gets too fast or the upper or
fower temperature limits are approached, the ex-
periments will be concluded.

Sudden Changes in Reactivity, The reactor will
be brought to a power of 1 Mw and allowed to
stabilize, From the previous calibration of the
regulating rod, the rod will be placed at a position
such that a complete withdrawal will give 10 cents
of excess reactivity. The rod will suddenly be
withdrawn and the transient response of the inlet
pressure and all nuclear and temperature recorders
noted. The experiment can be repeated with
sudden changes of 25, 50, 75, and 100 cents of
reactivity. If the fuel temperature coefficient has
a value of approximately -5 x 103, as expected,
a sudden change in reactivity of 100 cents should
be safe. However, if the experiments with smaller
reactivity changes indicate that such a large step
will be unsafe, this series of experiments will be
concluded.  Otherwise, the experiment will be
repeated from initial powers of 100 kw, 10 kw, and
successively lower power levels,

Reactor Startup Using the Temperature Coef-
ticient., The reactor will be brought to 1 Mw of

power and the shims adjusted so that the regu-
lating rod is completely withdrawn, The helium
flow will be cut off and the reactor power allowed
to drop to its normal power with no heat extraction
(estimated to be between 10 and 100 kw). The
regulating rod will now be inserted by 0.1% of
reactivity and the reactor allowed to go subcritical
for about 10 min. After this time, helium flow
will gradually be increased. If the temperature
coefficient is =5 x 10=3, a drop in the mean
temperature of the reactor of 20°F will bring the
reactor critical again. As soon as a positive
period is noted, the helium cooling flow will be
held fixed and the reactor allowed to come up to
power and level out of its own accord.

The experiment can be repeated by driving the
reactor subcritical by 0.2, 0.3, and 0.4% with the
regulating rod, Care will be exercised not to
approach the upper or lower temperature limits,

Effect of Xenon Buildup. The change in re-
activity calculated from xenon buildup in the ARE
is shown in Fig. D.7. The reactivity as plotted
is nominal because of uncertainties in the xenon
cross sections and is probably a maximum. The
shape of the curves represents the change in
reactivity if no xenon is lost by off-gassing. The
reactor will be operated for 5 hr at full power,
and then reduced to 10% of full power. If no xenon
is lost, the reactivity should change as shown in
the lower curve of Fig. D.7. The reactor will
then be operated for 25 hr at full power, and
subsequently reduced to 10% power. Again if no
xenon is lost, the reactivity change should be
as shown in the upper curve of Fig. D.7. If some
xenon does off-gas, the shape of the curves will
be changed, and by proper analysis, an estimate
of the amount of off-gassing can be obtained.

At full power, the reactivity change will be
calculated from the change in reactor mean temper-
ature and the temperature coefficient. At 10%
power, the reactor will be put on flux servo and
the movement of the regulating rod will measure
the reactivity change. Since the moderator will
contain considerable heat when the power is
reduced, the mean reactor temperature will slowly
decrease during about the first % hr after the
power is reduced to 10%. This will cause an
insertion of the regulating rod by the flux servo.
Since the xenon buildup will cause a withdrawal
of the regulating rod, a correction must be applied
for the effect of the drop in moderator temperature.,

153

 

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L& Ry

el

To obtain this correction experimentally, a
control experiment will be performed. The reactor
will be operated at full power for ‘/2 to 1 hr and
the moderator allowed to come up to temperature.
During this short time, no appreciable xenon will

be formed. The reactor will then be reduced to
10% power and the flux servo turned on, The
reactivity change from the drop in moderator
temperature will be observed and used as a cor-
rection for the xenon buildup experiments.

155

 

 

.

T N ST

 

- e e e

 
 

 

Appendix E

MATHEMATICAL ANALYSIS OF APPROACH TO CRITICALITY
. W. E. Kinney

When the ARE was brought to critical by suc-
cessive fuel additions, it was observed that the
usual plot of [1 — (1/multiplication constant)] vs
uranium concentrafion increased, at first, very
rapidly, but, when the curve got close to 1, the
rise was very slow. Qualitatively, such behavior
is observed in many reactors, but the ARE ex-
hibited the effect to an unusual degree., In order to
explain this, the ORACLE three-group, three-region
code was modified so that flux shapes at succes-
sive fuel additions could be calculated. Group
constants were obtained by flux weighting with
fluxes from an Eyewash calculation on the ARE,

Figure E.1 shows the space distribution of the
thermal flux for no fuel and for runs 2 through 6.

The effect of the reflector as fission neutrons be-
come more nhumerous can be seen, Figure E,2
compares experimental and calculated startup
curves for the fission chambers which were located
in the reflector as indicated in Fig. E.1. In the
calculation,

where CR is the counting rate of the fission
chamber, o7 is the fission cross section for group

i, and ¢, is the group i flux. For the ARE startup

CR].

 

ORNL—~LR—-DWG 4450

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

5

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& CORE i—a REFLECTOR i |
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e | | | :
4 \ | l |

i [s - | I
0 \o\o\o__o___?,—c[-/"fi\ |

2
xl |
0
0 {0 20 30 40 50 60 70

RADIUS {cm)

Fig. E.1. Thermal Flux vs Radius.

156

 

%
t
i

Y g g

g

e B

g g

A R T @mg‘m'vmw?“?\ TR T e T R ey Brnemmrmr

LA el L ] ol

 
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1.0
0.8
0.6
E ® CHAMBER |
= A CHAMBER 2
I © COMPUTED
0.4
0.2
0
0 2 4 6 8 10 12 14 16 18

U235 CONCENTRATION (Ib/f3)

Fig. E.2. 1 - (1/m) vs U235 Concentration.

where m is the multiplication constant, CR . is the
counting rate on run j, and CR_ is the counting
rate with no fuel. It seems, then, that the un-
expected, rapid initial rise in the counting rate of
the fission chamber and the [1 ~ (1/m)] curve is
due not only to the general rise in flux level but

also to the formation of the thermal-flux maximum
near the fission chambers. Once the shape of
spatial distribution of the thermal-neutron flux is
set up, the fission chambers register only the
general increase in flux level and the count rate
increases slowly,

157

 

'
g
a
¢
i

T

 
ol R

FROTTYCI

A e 1

 

 

 

Appendix F
COLD, CLEAN CRITICAL MASS

The value of the cold, clean critical mass is of
interest in connection with any reactor, since it is
the calculation of this elusive number that takes
so much time during the design of the reactor and
system. The cold, clean critical mass may be
found by extrapolating the curve of uranium (Ib of
U23%) in the core vs Ak/k (%) in the rods to the
point where there is no rod poisoning. On the
other hand, if the mass equivalent of the rod
poisoning is determined and deducted from that
then in the core, another curve may be obtained.
This curve gives corrected mass vs Ak/k (%) in
the rod and may also be extrapolated to 0% Ak/k
in the rods. Both these curves, as shown in
Fig. F.1, extrapolate to the same mass value at
0% Ak/k in the rods, which indicates that the rod
calibration was quite accurate and the data were
reliable. The value of the cold, clean critical
mass extrapolated from Fig. F.1 was 32.75 |b of
U235.

The data from which the curves in Fig. F.1 were
plotted are tabulated in Table F.1. The data used
to determine the cold, clean critical mass were
those obtained during the rod calibration from fuel
addition (Exp. L-2). This experiment and Exp. L-5,
rod calibration from reactor period, then provided
the information on the value of the regulating rod,
The values of the shim rods were taken from
Appendix J, ‘“Calibration of the Shim Rods.*” The
Ak/k values in both regulating and shim rods were
then converted to their mass equivalents by using
the (Ak/k)/(AM/M) ratio of 0.236, as determined
experimentally, Three points listed in the table
and shown in the figure were derived by using the
first regulating rod, which was subsequently re-
placed because it was too ‘‘light.”” The calibra-

158

tion for that rod was not accurate, and, hence, the
three points were disregarded in extrapolating the
curves,

ORNL—LR—DWG 6419

 

36 T

 

I

CURVE J ‘
35 A: U255 N CORE vs. Ak/k IN RODS
B:  Ak/k IN RODS vs. U2% IN CORE CORRECTED TO NO
ROD POISCNING BY USING (A4/k) / (AM/M) =0.236

| | |
O OBTAINED WITH FINAL REGULATING ROD

34 [~ e OBTAINED WITH INITIAL REGULATING ROD, WHICH WAS ™
REPLACED BEFORE BEING WELL CALIBRATED

i 50T A

 

 

 

 

 

33 /j =
—_ e | -
2 Rgers ¢
o N
E N
32 \‘\\
. }\
.

 

30 ]

 

 

 

 

 

 

 

 

29

 

O Q2 04 Q86 08 1.0 1.2
Ak/k IN ALL RODS ()

Fig. F.1. Extrapolation to Cold, Clean Critical
Mass.

 

T e

.

g

> M rprapety B Tt g W o vy g gt ©

B B el -

 

" e g e eSS —

 
pdig,

W RS k.

TABLE F.1. CRITICAL URANIUM CONCENTRATIONS FOR VARIOUS ROD INSERTIONS

 

A

W aaie

.

Lo S A e,

P

o e an

e, dn i, . -

I‘

 

 

(Exp. L-2)
Regulating Shim Rod Insertion AL/E of Ak/F of Shim Rods y2ss U235 Total Ak Total U235 i,
Run Ne, Rod Insertion (in.) Reguiating (%) Concentration in Core in Rods AM/M in Clean Core
(in.) Ne.T No.2 No.3 Rod(% No.1 No.2 No.3 (Ib/#3) (ib) (%) Rods (Ib)

0 0.92 3.3 1.7 0 0.016 0.19 Q.08 O 23.94 32.80 0.286 1.212 3159 A
1 3.35 3.3 1.7 ¢ 0.058 0.19 008 0 23.98 32.85 0.328 1.390 31.46 k
2 0.93 3.3 2.5 ¢ 0.016 0.19 0.13 0 24,12 33.04 0.326 1.381 31.66 k

3 4,25 0 0 3.5 0.140 0 0 0.20 24,17 33.11 0.340 1.441 31.67
4 6.35 0 0 3.5 0.210 0 0 0.20 24.22 33.18 0.410 1.737 31.44 ;
5 7.5 0 0 3.5 0.248 0 0 0.20 24.25 33.22 0.448 1.898 31.32 :

6 8.5 0 0 3.5 0.281 0 0 0.20 24,27 33.25 0.481 2,038 31.21
Level 7.4 2.2 2.5 0 0.244 011 013 O 24,27 33.25 0.484 2,051 31.20 i

Trimmed

7 8.0 2,2 2.5 0 0.264 0.1 013 0 24.29 33.28 0.504 2135 31.14 k.
8 8.52 2.2 2.5 0 0.281 ¢.1t 03 ¢ 24,31 33.30 0.521 2.207 31.09 lr
9 10,92 2.7 2.8 0 0.360 0.14 015 0 24.41 33.44 0.650 2,754 30.69
10 1.25 5 5 ¢ 0.041 033 033 0 24.43 33.47 0701 2,970 30.50 E
11 11.56 4 3 o 0.382 023 016 0 24.44 33.48 0.772 3.271 30.21 :”
b

159

 

 

T T IR R

 
i oo ¥ i, s, i

Appendix G
FLUX AND POWER DISTRIBUTIONS

A. D, Callihan

There were, of course, no measurements of the
flux or power distribution in the reactor during the
actual  experiment, These distributions were
measured, however, on a zero power, critical mock-
up of the experimental reactor over a year earlier.
A detailed description of the critical mockup, in-
cluding the reactor parameters obtained therefrom,
may be found in ORNL-1634,! The neutron flux
distributions obtained from indium and cadmium-
covered indium foil measurements, the fission
neufron flux distributions obtained by the catcher-

 

 

 

1A. D. Callihan and D. Scott, Preliminary Critical
Assembly for the Aircraft Reactor Experiment, ORNL-
1634 (Nov. 28, 1953).

D. Scott

foil method, and the power distributions obtained
from aluminum catcher foils, which are of particular
interest, are presented here,

NEUTRON FLUX DISTRIBUTIONS

The neutron flux distributions were measured
with indium and cadmium-covered indium foils in a
number of runs in which a remotely placed uranium
disk and an aluminum catcher foil were used to
normalize the power from run to run, The results
of the bare indium and cadmium-covered indium
traverses made at a point 12,06 in. from the center
of the reactor are shown in Fig. G.1. The zero of
the abscissa is the bottom of the beryllium oxide

DWG, 24529A

 

70

AT 12.06 in. RADIUS

INDIUM TRAVERSE, LONGITUDINAL

 

60 /1’"
0 /

| —BARE INDIUM ACTIVATION

A

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

0
=
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x 40 / 0.8
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=
& CADMIUM -COVERED
< INDIUM ACTIVATION
>
E 30 \ 0.6
=
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g g
5
/CADMIUM FRAGTION <
20 \ 0.4 &
/
— N\
—" TOP OF BeO =
e O e 5{)
o
10 e N 0.2
BARE INDIUM ACTIVATION MINUS |
CADMIUM-COVERED INDIUM AGTIVATION
INCONEL SUPPORT PLATE
0 =~ | | 0
0 6 12 18 24 30 36

DISTANCE FROM BOTTOM OF REACTOR (inl)

Fig. G.1. Longitudinal Neutron Flux Distribution.

160

PR .

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column where it rests on the l-in.-thick Inconel
support plate, The scattering by the Inconel
probably accounted for the reduction of the cad-
mium fraction at this point, ’

The radial flux traverse at the mid-plane of the
reactor with the regulating rod inserted is given in
Fig. G.,2, The dashed lines on the figure are
activities extrapolated from the data obtained in
the fine-structure measurements made near the
11-in. position. The wide gap in this traverse was
unexplored because of the importance which was
attached to the study of a unit core cell in the
time available., This emphasis has been at least
partially contradicted by the fission flux traverse
described in the following section. The strong
effect of the center regulating rod assembly is
indicated in these curves. The comparatively
small flux and cadmium fraction depressions at

the center of the regulating rod accentuated the
relative inefficiency of this type of rod and guide
tube arrangement for reactor control. In another
measurement of the radial flux, this time with the
regulating rod withdrawn, the traverses were very
similar to that shown in Fig. G.2, but the neutron
flux was slightly greater (~8%).

FISSION-NEUTRON FLUX DISTRIBUTION

A measure of the distribution of neutrons that
caused fissions was obtained by using the catcher-
foil method in which the aluminum foil, together
with some vuranium, was both bare and cadmium
covered. A radial traverse taken at the mid-plane
of the reactor is shown in Fig. G.3. The flux of
low-energy neutrons produced fission peaks near
the reflector and was depressed by the Inconel
tube (regulating rod sleeve) at the center. The

 

 

 

  

 

 

  

 

 

  

 

 

 

 

  

 

 

 

  

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

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4 6 8 10 12 14 16 18 20 22 24

 

DISTANCE FROM AX!S OF REACTOR (in)

Fig. G.2. Radial Neutron Flux Distribution.

161

B RO TR T TR 1

 

 

b
L
k
L
I

 

 
 

 

cadmium fraction followed roughly the same pattern.
Similar half-length longitudinal traverses are shown
in Fig., G.4 for a position 10.09 in. from the center,

POWER DISTRIBUTION

Three longitudinal power distributions in this
reactor assembly were measured by using aluminum
catcher foils placed against the end surfaces of
the short, cast, fuel slugs contained in an Inconel
tube., The size of these foils was such that the
counting rate from each was higher than necessary
for the desired statistics. In the arbitrary units
reported, an activity of 50 represents approximately
105 counts. Had time permitted, it would have

been desirable to repeat the experiment with
smaller foils to improve the resolution, particularly
since the results were very sensitive to foil loca-
tion because of fuel self-shielding. Each catcher
foil was nominally located centrally on the axis
of a fuel slug. One traverse extended from below
the Inconel support plate to above the top of the
beryllium oxide; the other two covered the upper
half of the core. The data obtained in the fuel
tubes at the indicated radii are shown in Fig. G.5.
The data of Fig. G.5 are replotted in Fig. G.6 to
show the radial power distribution at several ele-
vations in the reactor, all normalized at the center.
It is to be noted thatthe 37.86-in. elevation traverse
is 2 in. above the top of the beryllium oxide column.

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

DWG. 24533A
14 l
CATCHER-FOIL FUEL DISK RADIAL AT MID-PLANE
f2 ? i t\
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e L
2 i ‘ 1
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0 L l ‘ ‘ 0
0 3 6 9 24 27

 

DISTANCE FROM AXIS OF REACTOR (in)

Fig. G.3. Radial Fission Neutron Flux Distribution,

162

 

 

 

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ACTIVITY (arbitrary units)

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

DWG 215344

| T |

CATCHER - FOIL FUEL DISK LONGITUDINAL AT 40.09 in. RADIUS
T ll 1.2
A~ ,BARE ACTIVATION
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CADMIUM - COVERED ACTIVATION\
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DISTANCE FROM BOTTOM OF REACTOR (in.)

. Fig. G.4, Longitudinal Fission Neutron Flux Distribution,

163

 

:
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ACTIVITY (ARBITRARY UNITS)

164

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DISTANCE FROM BOTTOM OF REACTOR (in.)
Fig. G.5. Longitudinal Power Traverses,
DWG 2i536A
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DISTANCE FROM AXIS OF REACTOR {in.}
Fig. G.6. Radial Power Traverses.

 

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: Appendix H
POWER DETERMINATION FROM FUEL ACTIVATION'
E. B. Johnson

Perhaps the most basic value obtained from the
operation of the ARE was the power level at which
it operated. This quantity was determined both
from activation of the fuel and from a heat balance.
The determination of reactor power from activation
of the fuel was based on the measurement of the
relative activity of samples of the fuel exposed in
the ARE and that exposed in a known flux in the
Bulk Shielding Reactor (BSR).

The activity of the aggregate of fission products
is not easy to predict because of the large number
of isotopes for which calculations would be re-
quired. For identical conditions of exposure and
decay, that is, for identical flux, exposure time,
and waiting period after exposure, the specific
activity should be the same, however. Further-
more, at low fluxes the activity should be pro-
portional to the flux.

The ARE was operated at a nominal power level
of approximately 1 w for 1 hr, was then shut down,
and a sample of the irradiated fuel was withdrawn
for counting and analysis. Since the activity in
this fuel sample was too low to measure accurately,
the test was repeated at an estimated power of
10 w. The ARE_experimental program then pro-
ceeded as scheduled. Decay curves were obtained
from the fuel samples from the ARE and were then
compared with the decay curve of a similar fuel
sample irradiated in the BSR in a known neutron
flux of about the same magnitude. From the re-
lative activity of the samples, o determination of
the power of the ARE was made.

THEORY

2

It has been shown® that the power (in watts)

produced in an enriched-fuel thermal reactor is

P = 4264 x 107" x nv, x G

where G is the number of grams of U235 in the

volume over which the average thermal-neutron flux
is nv,,, and the constant, 4.264 x 10~ 1! contains

 

1This appendix was originally issued as ORNL
CF-54-7-11, Fuel Activation Method for Power Determi-
nation of the ARE, E. B. Johnson (July 31, 1954), and
was revised for this report on April 22, 1955.

2), 1, Meem, L, B, Holland, and G. M, McCammen,
Determination of the Power of the Bulk Shielding Re-
actor, Part lIl. Measurement of the Energy Released
per Fission, ORNL-1537 (Feb. 15, 1954).

fore (P/G)

the fission cross section, energy release per
fission, and the necessary conversion factors.
Obviously, if the thermal-neutron flux and the
amount of fuel present are known, it is quite simple
to obtain the power. The problem was to obtain
the power without measuring the flux in the ARE.

The quantity P/G obtained from the measurement
in the BSR will be called (P/G)BSR The decay
curve for the two fuel samples irradiated in
the BSR is shown in Fig. D.5 of Appendix D. The
power production in a given sample is proportional
to both the counting rate and the amount of fission-
able material in the sample. Thus,

 

(P/G)ARE (CR/G)ARE
(P/G)BSR (CR/G)BSR
where
CR = counting rate at time ¢ after shutdown,
G = weight of uranium in the sample.

This equation can be rewritten as

CRARE GBSR

= (P/G) X e— X

BSR
CRBSR GARE

(P/G)

 

ARE

The total power of the ARE is then the product of
the power per gram (in the sample)} and the total

amount of uranium in the fuel circuit (Gtot)ARE‘ or

P arE = (P/G)ape * {Gio)are

(P/G) CRARE

= X o—_—_—

BSR

CRgsr

Gesr

X G x «%OJARE
ARE

The unperturbed thermal-neutron flux in which the
samples were exposed in the BSR was 1.895 x 107
neutrons/cm?.sec. However, the self-depression®
of the flux by the uranium sample was 0.80. There-

BSR become.s

(P/G) = 4.264 x 10~ 11 % 1.895 x 107 x 0.80

BSR
= 6.464 x 107% w/g

 

3w, K. Ergen, private communication.

165

ST T T R TSP peT T T

 
 

 

g G e, s ot ol bt Wik s b de BB T s e e R e b S e E S S s e e el e, AL < :

Each of the fuel samples irradiated in the BSR
contained 0.131 g of U233, Thus
CR
ARE
6.464 x 1074 w/g x — ———
CRBSR

0.131

i

PARE

 

It

8.468 x 107> x —

(Gfor)AR E

CARE

X

This equation was then used to determine the
power of the ARE as indicated by fuel activation
during the previously mentioned operation of T hr

at 1- and 10-w nominal power. The instruments
which recorded the neutron level (log N, etc.) were
then calibrated in terms of reactor power.

EXPERIMENTAL PROCEDURE

The BSR is immersed in a pool of water which
serves as moderator, coolant, reflector, and shield.
Therefore, any material which is to be activated
in the BSR must be placed in a watertight container
before immersion in the pool. Capsules to contain
the ARE fuel were made of 25 aluminum with screw-
type caps and were sealed with a rubber gasket.
Each capsule contained approximately 1 g of ma-
terial; one of the capsules is shown in Fig. H.1.

Since it was desirable to irradiate the capsules
in the reactor core, the reactor was loaded with a
partial element, containing only half the usual
number of plates, in the interior of the lattice, as

/\ ORNL-LR-DWG 1947

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Fig. H.1. Apparatus for Fuel Exposure in the Bulk Shielding Reactor,

166

 

>

T

p e ¥

Y e TR

T

L
b

 

TN P

rErmEr T e

| BTNt o m— " ————— 'F’V@"T”'W B — W e mer
ol

i,

PN

Sl .

ek

S

SRRk B

i d.

"

shown in Fig. H.2. A lucite holder was designed
which would fit inside this partial element and
support the capsule to be irradiated at approxi-
mately the vertical centerline of the element. Small
gold foils for measuring the thermal-neutron flux
were mounted in ‘‘drawers’’ adjacent to the cap-
sule, as shown in Fig. H.1. A monitoring foil was
placed on the bottom of the lucite block.

ORNL-LR-DWG 12483

PARTIAL ELEMENT FOR

OO ) )O6
® @) hehe) o)) ®®
@)= ) =)0 ) =)@ @
@O @)=)=)f)=))® @
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E@@@MB%@D@@

PARTIAL ELEMENTS / \ CONTROL ROD

Fig. H.2. Bulk Shielding Reactor Loading
(No. 27).

 

 

 

 

Two samples of each of two ARE-type fuels of
different composition were irradiated. One (No.
44) contained 53.5-40.0-6.5 mole % of NaF-ZrF -
UF, and was the anticipated fuel of the ARE. The
other was No, 45, which was the carrier to which
the uranium-containing fuel concentrate was added
during the critical and low-power experiments; it

contained 50-50 mole % of NuF—ZrF4.

[t was necessary in the BSR to expose the
samples in the aluminum capsules; at the ARE the
fuel was, of course, first activated and then poured
into the capsules. Therefore it was necessary to
determine the contribution of the aluminum capsuie

to the activity observed. Two empty capsules were
exposed (in separate runs) for 1 hr at a nominal
power level of 1 w. At the same time, gold foils
were exposed in the positions adjacent to the
capsule to determine the thermal flux. Similarly,
two carrier-filled capsules and two fuel-filled cap-
sules were irradiated, each separately but each
for 1 hr at 1 w. The capsules each contained 1 g
of material.

The decay curves taken on each capsule were
made with two counters because of the difference
in disintegration rates between the empty and the
fluoride-filled capsules. A scintillation counter
was used to obtain the curves for the empty cap-
sules and for the carrier-filled capsules, and the
high-pressure ion chamber was used to measure
the activity in the fuel-filled capsules. The data
for the carrier-filled capsules were converted, on
the assumption that all the activity was attributable
to sodium, to equivalent countingrates in the high-
pressure ion chamber. The assumption that the
activity was due to sodium was apparently valid,
since the decay curve indicated a half-life of about
15 hr at approximately 6 hr after shutdown.

[t was found from the decay curves that the
aluminum activity was a factor of approximately
10 less than that of the carrier-filled capsules 3 hr
after shutdown and a factor of 450 less than that
of the fuel-filled capsules at the corresponding
time, and was therefore negligible. The activity
in the carrier-filled capsule was about 4% of that
in the fuel-filled capsule 6 hr after shutdown.
There was essentially no difference between the
two different fuel-filled capsules and the two dif-
ferent carrier-filled capsules. For this reason,
only one curve is shown for each in Fig. D.5.

The nuclear power of the Aircraft Reactor Experi-
ment was then set on the basis of the activation
in fuel samples irradiated for 1 hr at 10 w. This
power calibration was the basis of all power levels
indicated in the nuclear log up to the last day of
the experiment, by which time it had become obvi-
ous from the various heat balances that the actual
reactor power was considerably higher (app. L).
A comparison of the two methods of power cali-
bration is given in Appendix N.

167

v e e

 

 

b
F

 
 

 

 

- Appendix |
INHOUR FORMULA FOR A CIRCULATING-FUEL REACTOR WITH SLUG FLOW!
W. K. Ergen
As béin’red out in a previous paper,? the circu-  circulating-fuel reactor has been discussed in some

lating-fuel reactor differs in its dynamic behavior
from a reactor with stationary fuel, because fuel
circulation sweeps some of the delayed-neutron
precursors out of the reacting zone, ond some
delayed neutrons are given off in locations where
they do not contribute to the chain reactions. One
of the consequences of these circumstances is the
fact that the inhour formula usually derived for
stationary-fuel reactors,? requires some modifica-
tion before it becomes applicable to the circulating-
fuel reactor, The inhour formula gives the relation
between an excess muitiplication factor, introduced
into the reactor, and the time constant T of the
resulting rise in reactor power. If the inhour
formula is known, then the easily measured time
constant can be used to determine the excess
multiplication factor, a procedure frequently used
in the quantitative evaluation of the various
arrangements causing excess reactivity, Further-
more, the proper design of control rods and their
drive mechanisms depends on the inhour formula,

Frequently, the experiments evaluating small
excess multiplication factors are carried out at
low reactor power, and the reactor power will then
not cause an increase in the reactor temperature,
This case will be considered here. In this case,
the time dependence of the reactor power P can be
described by the following equation:4

(1) dP/dt

= (/)k,, - PP +‘Bfo°°o(s) P(t - s)ds] .

7 is the average lifetime of the prompt neutrons
and k__ the excess multiplication factor (or excess
reactivity),  The meaning of 8 and D(s) for a

 

TThis appendix was issued earlier as ORNL CF-53-
12-108, The Inbour Formula for a Circulating-Fuel
1’;Iucle_ar Reactor with Slug Flow, W, K, Ergen (Dec. 22,
953).

2william Krasny Ergen, J. Appl. Phys., Yol. 25, No. 6,
702-711 (1954).

SSee, for instance, S. Glasstone and M. C, Edlund,
The Elements of Nuclear Reactor Theory, D. Yan
Nostrand Co., lnc., 1952, p. 294 ff.

4Some authors, for instance Glasstone and Edlund,
loc, cit., write the equations corresponding to (1) in a
slightly different form., The difference consists in terms
of the order k_ (T/T) or k__f, which are negligibly
small. ex ex

168

detail in ref. 2, We approximate in the following
the actual arrangement by a reactor for which the
power distribution and the importance of a neutron
are constant and for which all fuel elements have
the same transit time 6, - ¢ through the outside
loop. In this approximation, 8 D(s) is simply the
probability that a fission neutron, caused by a
power burst at time zero, is a delayed neutron,
given off inside the reactor at a time between s
and s + ds. D(s) is normalized so that

(2) JDls)ds =1 .

If the fuel is stationary, 8D(s) is the familiar curve
obtained by the superposition of 5 exponentials:

2 “As
(3) BD(s) = LBrge 7 .

i=1
The A, are the decay constants of the 5 groups of
delayed neutrons, and the (3, are the probabilities
that a given fission neutron is a delayed neutron
of the 7th group.

For the circulating-fuel reactor we first consider
the fuel which was present in the reactor at time
zero. At any time s, only a fraction of this fuel
will be found in the reactor. This fraction is
denoted by F(s), and by multiplying the right side
of (3) by F(s), we obtain the function 8D(s) for
the circulating-fuel reactor,

Since @, is the total time required by the fuel
to pass through a complete cycle, consisting of
the reactor and the outside loop, it is clear that

ats = nfll

(n =0,1,2,..), F(s) is equal to 1;

at s

m9]+6

(n = 0,1,2...), F(s})is equal to zero.

(We assume 91 2 20 so that the fuel under con-
sideration has not started to re-enter the reactor
when the last of its elements leaves the reacting
zone.) Between s = nf, and s = n0, + 6, F(s)
decreases linearly, and hence has the value

(n0, + 0 ~ s)/6. Ats =nl, - 6(n = 1,2, 3...),

 

W T RTT AR MW g BT R M S TEEET N Yol Sy

LT T s e T R R

PWET TR § WEPER M

Fwy-cEr wemoTRRT

e e -

- T T

 
e

ki

R N

Ak

-

R X ek B,

!

F(s) is zero, but since the fuel under consideration re-enters the reactor between this moment and
s = n@,, F(s) increases linearly: F(s) = (s ~ n@; + 8)/8. For BD(s) we thus obtain:

n@ + 0 —- s

BD(s) = ZBZ is forn@lgsgnel-}-@, n=012...,
i=1
4 S—n6]+6 5 A
) BD(S):”“‘_"E_——EB forn@l-—egsénfil,n=],2,3,...,
i=1
BD(s) = 0 forn6]+9§s§(n+])61—9,72=0,],2,....

Equation (4) is now substituted into Eq. (1), and for P we set P = Poet/T. Then

n6+6n6 + 0 - s As

- 5
/T T i -
.Fpoel = (kex - B) Poet/ + ZIB!AZ Z f e I Poe(t S)/T dS
L=

@ n@ s —nf, + 6
+ E ! "*——“*“——'—‘] e*Ais P elt=s¥VT gg
0 0
n=1 nBI—G

t/T

The common factor Poe cancels out. The substitution o = n@l + @ — s transforms

Jio "% w0, + 0 = ) expl-Dn, + (1/T)s} ds

info

exp{—n@l}h + I/T)]} exp{ oA, + ]/T)] foor exptlA; + (1/T)lol do ,

and the substitutiono = s — n@ + 0 tronsforms

Js,

o _g s —nty +0) expt—[A; + (1/T)] s} ds

into

expi—n0 A, + (1/T)I} expiblr; + (1/T)} foeoexp{—[)\i + (V/T)l ol do

The geometrlc series exp {—n[)\ + (1/T)10,} can now be summed, and the integrals over o evaluated
by elementary methods. After performlng all 'rhese operations, one obtains the following inhour formula:

 

= ;0 .8 - (8,=0
T ]5 pt —= 1 + e - e z](LLZ-8+|)+eM’(] )
(5) by =t B - E - ,
. | = R - e N
(6) p, = A+ (1/7)

B is obtained by integrating Eq. (3) from 0 to . Expressions are obtained which are of the same type
as the ones just dlscussed and whnch can be evaluated by the same methods. The result is

' ~A.8 ~-A.8 ~X.(By =6
| s AB-Tlie o l]()ti0+])+e 179
7 p =Y A Y
=1 (1 —e 1

 

169

 

 

T T ——

-

-

EREoT T TR TR - — Y T
 

 

 

In spite of the formidable appearance of Egs. (5),
(6), and (7), it is easy to find, for any given 0 and
6,, the value of £_ which produces a given time

constant T,

Furthermore, the following reasoning describes
the general features of the equations. Consider
first the dependence of k__on T. I T =, .= A,
and the sum on the right of (5) is equal to (3, k__
is equal to zero, This corresponds to the state in
which the reactor is just critical. |f T becomes
very small, the u. become very large and in the
fraction on the right of (5) the numerator is domi-
nated by ;.0 and the bracket in the denominator by
1. Hence the fraction tends to zero like 0/p;, as
T goes to zero, If 7 is very small, as it is in
practice, T will be small as soon as % , exceeds
3 by a small amount. Then the complicated sum on
the right of (5) is of little importance, and T is
determined by 7/T = k__ - 3, that is, the reactor
period is inversely proportional to the excess of
the reactivity over the reactivity corresponding to
the ‘‘prompt critical’’ condition,

In the stationary-fuel reactor, the k__ which
makes the reactor prompt critical is given by Zf8 ..
With circulating fuel, the reactor is prompt critical
if & = B, which is less than 23, that is, it
takes less excess reactivity to make the circulating-
fuel reactor prompt critical than to do the same
thing to a stationary-fuel reactor, This is physically
evident because the fuel circulation renders some
of the delayed neutrons ineffective. That £ is
less than 3. can also be verified mathematically.

For T intermediate between very small positive
values and +wx, we consider again the analogy to
the stationary-fuel reactor, Here the inhour formula
reads:>

T 5 B
k, =— —_—,
(8) & x T " ig] T+ AT

For every positive & there is one, and only one,
positive time constant T, and vice versa. This is
a consequence of the fact that k__ is a monotonic
decreasing function of T, for if this monotony did
not exist, there could be several positive T values
corresponding to a given value of &__, or vice
versa. In the circulating-fuel reactor the situation
is qualitatively the same; the fractions under the

 

5See Glasstone and Edlund, loc. cite, p» 301, Eq.
10.29.1. See also preceding footnote.

170

sum in Eq. (5) are essentially of the form

x =~ | + 7% w g (x + ]) + e—(a—l)x

 

x2“ _ e-—ax]
(x = “ie]la = 9]/6) 1

and an expression of this form can be shown to be
a monotonic decreasing function of positive x;6
the expression is thus a monotonic increasing
function of T [see Eq. (6)], and as T increases k__
decreases monotonically, according to Eq. (5);
for every positive k__ there is one, and only one,
positive T, and vice versa, This, of course, does
not preclude that there exists for a given k__
several negative T values, in addition to the one
positive value. Negative T correspond, however,
to decaying exponentials, which are of no im-
portance if the rise in power is observed for a
sufficiently long time.

Consider now the behavior of Eq. (5) with varia-
tion of 6, the transit time of the fuel through the
reactor, and of 6., the transit time of the fuel
through the whole loop. If 6 (and hence also 6,)
is very large compared to all 1/A. and 1/p, Eaq.
(5) reduces to Eq. (8), the inhour formula for the
stationary-fuel reactor. The circulation is so slow
that the reactor behaves as if the fuel were sta-
tionary, inasmuch as cll delayed neutrons, even
the ones with the long-lived precursors, are given
off inside the reacting zone, before much fuel
reaches the outside. On the other hand, if 6, and,
hence, also 6, is small compared to all 1/A, that
is, if the transit time of the fuel through the com-
plete loop is small compared to the mean life of
even the short-lived delayed-neutron precursors,
then for T >> #

B
1 + /\I.T

g
©) k=4
ox T 6]

5
i=1
This is the same as the inhour formula for the
stationary-fuel reactor, except that all the fission
yields 3. are decreased by the factor 6/6,. This
is physically easy to understand, since 6/0, is
just the probability that a given delayed neutron
is born inside the reactor, |

Of interest is the intermediate case, in which @

 

bw. K. Ergen, The Behavior of Certain Functions Re-
lated to the Inbour Formula of Circulating Fuel Reactors,

ORNL CF-54-1-1 (Jan. 15, 1954).

 

PYEETTT OE LM A0 g RO T L owmemesetenen. 9

T ME e

S emw oy Ew T

T

e ey

e

e T

-

TEEIETTT PN eEETERE EPSERIPET wrEyrr W

T mET o e

e

-

“
¥

 
bl e R

L - Al

kL.

o, s

vk

o e, B A

£,

b g B S s

i

Ak wba

i

is smaller than the mean life of the long-lived
delayed-neutron precursors and larger than the
mean life of the short-lived precursors. In that
case, the long-delayed neutrons act approximately
according to Eq. (9) and are reduced by the factor
6/6,. On the other hand, the neutrons with the
short-lived precursors behave approximately like
Eq. (8) and are not appreciably reduced. Hence,
a small excess reactivity enables the reactor to

increase its power without ‘‘waiting’’ for the not
very abundant long-deiayed neutrons., The reactor
goes to fairly short time constants with surprisingly
small excess reactivities. However, to make the
reactor prompt critical, that is, to enable it to

exponentiate without even the little-delayed neu- -

trons, takes a substantial excess reactivity be-
cause of the almost undiminished amount of the
latter neutrons,

171

 

 

 
 

 

 

 

Appendix J
CALIBRATION OF THE SHIM RODS

Three essentially independent methods were
used to find the value of the shim rods in terms
of reactivity. The first method involved an
analysis of the counting rate data taken during the
critical experiment, and the second was a cali-
bration of shim rod No. 3 in terms of the regulating
rod., The agreement between these two methods
was very good. As a check on the general shape
of the reactivity curves as a function of rod
position, the rods were also calibrated by using
the fission chambers. The various methods are
discussed in detail below.

CALIBRATION FROM CRITICAL EXPERIMENT
' DATA

After about one-half the critical mass of uranium
had been added to the system, the counting rates
of the two fission chambers and the BF, counter
were taken as a function of shim rod position for
each subsequent fuel addition until criticality was
reached (as described in the main body of this
report, chap. 4). Because the fission chamber
counting rates were subject to the phenomenon
noted in Appendix H, only the BF3 counter data
were used in the rod calibration.

From the BF3 counting rates for each rod po-
sition token after o given fuel injection, the
multiplication M was determined from the relation-
ship

M N
NO ’
where
N = counting rate for a given fuel concen-
tration and shim rod setting,
N, = counting rate before start of enrichment,

For each value of M the value of the multiplication
factor & was determined:

]
E=1-—.
M
Figure 4.9 of Chapter 4 shows k plotted as a
function of the uranium in the system for rod
positions of 20, 25, 30, and 35 in.
The shim rod calibration was obtained from
Fig. 4.9 by first making a cross plot of & against
rod position at the critical mass, as shown in

172

Fig. J.1. Then, with this plot, a value of
(Ak/k)/in. was obtained for every 1-in. movement
of the shim rods from no insertion to 16 in. of
insertion of the rods. A plot of (Ak/k)/in. as a
function of rod position is given in Fig. J.2, and
the data are tabulated in Table J.1.

In order to find the integrated reactivity, or total
worth of the rods, in terms of (Ak/k) as a function
of the number of inches of insertion, the curve
of Fig. J.2 was divided into three sections. Over
each section the curves were fitted to a formula
of the form

o = Ad?2 + Bd + C ,

where o = (Ak/k)/in. The formula and its integral,

which were applied to the three sections of the
curve, are given in Table J.2. For each section,
then, the integrated reactivity was found by inte-
grating over the a curves. The integrated curves
were of the form

Ak 3 , 12 i
p=‘k———~A’d + Bd + C'd

The three sections into which the curve of Fig. J.2

ORNL—LR—DWG 6420

s————INSERTION OF SHIM ROD (in.}

 

 

 

 

i5 10 5 i 0
1.0f | : :
1.00 ‘ =
| //
099 //
0.98 /

 

= 097 /
0.96

0.95 /
s L/

/

18 20 22 24 26 28 30 32 34 36
SHIM ROD POSITION (in.)

 

 

 

 

 

 

 

 

 

 

 

 

 

 

093

 

Fig. J.1. Shim Rod Position as a Function of
k at Criticality.

 

e w b

wd

o ald

-

srarer v W AT e Bt

»*W Lm_,p e

oy v

 
-k -

e

O ke Al

g, i S, Rl i, e, |

e R

. Bk

TABLE J.1. CALIBRATION OF SHIM RODS FROM FUEL ADDITION DURING CRITICAL EXPERIMENT

 

Rod Position

 

Average Insertion

Movement of

 

 

Rods ‘
Interval Taken AV;;Zg:iiod of Shi.m Rods Ave]:clge Ak Shim f"?ods, [(Azk/k)/in.] Av{:'Z}jjk;J;rnfiOd
(in.) {in.) Ad (in.)
(in.)
36 to 35 35.5 0.5 1.0 0.001 1 0.001 0.00033
35t 34 34.5 1.5 0.9992 0.0015 1 0.00150 0.00050
34 to 33 33.5 2.5 0.9975 0.0020 1 £.00200 0.00067
32 to 33 32.5 3.5 0.9953 0.0025 1 0.00251 0.00083
31 to 32 31.5 4.5 0.9926 0.0028 1 0.00282 0.00094
30 te 31 30.5 5.5 0.9895 0.0032 1 0.00323 0.00108
29 te 30 29.5 6.5 0.9862 0.0035 1 0.00355 0.00118
28 to 29 28.5 7.5 0.9825 0.0038 1 0.00387 0.00129
27 to 28 27.5 8.5 0.9785 0.0041 1 0.00419 0.00140
26 to 27 26.5 9.5 0.9743 0.0045 1 0.00462 0.00154
25 to 26 25.5 10.5 0.9695 0.0050 1 0.00516 0.00172
24 to 25 24.5 11.5 0.9643 0.0055 1 0.00570 0.00190
23 to 24 23.5 12.5 0.9585 0.0060 1 0.00626 0.00209
22 to 23 22.5 13.5 0.9523 0.0064 1 0.00672 0.00224
21 to 22 21.5 14.5 0.9458  0.0049 1 0.00730 0.00243
20 to 21 20.5 15.5 0.9386 0.0075 1 0.00799 0.00266
TABLE J.2. FORMULA (AND INTEGRAND) FITTED TO SHIM ROD SENSITIVITY CURVE

 

Range of Use

 

Formula No. i) Type Formola*
1 0to 8 Differential = ~0.000642 + 0.01874 + 0.0239
2 8 to 16 Differential o = 0.000542 + 0.0055d + 0.059
3 16 to 18 Di fferential o = —0.004542 + 0.1655d — 1.221
4 0to8 Integral p = ~—0.00024° + 0.009354% + 0.0239d
5 8 to 16 Integral p = 0.000167d° + 0.002754° + 0.0594
6 16 to Integral p = —0.00154> + 0.08754% — 1.221d

18

 

xa = (Ak/E)/in.
p = De/k

was divided and the correspending formulas which
were fitted to the curve are given in Table J.2.

In ‘the case of the integrated curves, the Ak/k
found by integration in the range of the curve was
added to the total Ak/k of the previous curve to
give the total Ak/k to the point of integration.
The resulting worth, (Ak/E), of the shim rods over
the first 18 in. of their movement is shown in
Fig. J.3 and the data are tabulated in Table J.3.

It should be noted that the last 2 in. of the curve

of Fig. J.2 (16 to 18 in.) was extrapolated. The
basis for the extrapolation is the curve shown in
Fig. D.4 of Appendix D, which is a calibration of
an ARE regulating rod made during some pre-
liminary experiments done in the Critical Experi-
ments Facility., The shim rods were assumed to
give the same type of curve,

If it were assumed that the last 18 in. of the
shim rods gave a shape of Ak/k vs rod insertion
which was essentially the image of the first

173

S

e e e g

 

W e g -

 

T

T
 

 

0.32
0.30
0.28
0.26
0.24
0.22

.20

016
044
042

040

REACTIVITY (Ak /&) /in. of SHIM ROD (% /in.)

0.08

0.06

0.04

0.02

0 { 2 3 4 5 6 7 8

 

9

ORNL-LR—DWG 6424

SHIM ROD NO.3
CALIBRATION AGAINST
REGULATING ROD AT
LOW POWER (EXP.L-6)

7/

’

®
CALIBRATION FROM FUEL ADDITION
DURING CRITICAL EXPERIMENT

{AVERAGE OF ALL RODS)

® CRITICAL EXPERIMENT DATA
O LOW-POWER EXPERIMENT DATA

10 i 12 13 14 15 16 i7 18 19 20

INSERTION OF SHIM RODS {in.)

Fig. J.2. Differential Shim Rod Sensitivity.

18 in. (see Appendix D, Fig. D.3), then the total
worth of a shim rod was 5.8% (Ak/k) and the total
worth of all three rods was about 17% (Ak/E).

CALIBRATION AGAINST THE REGULATING
ROD

In Exp. L-6, with the reactor at a power of
1 watt (nominal), shim reds Nos. 1 and 2 were
set at predetermined positions. Then shim rod
No. 3 was moved over various small increments
of its travel and thus calibrated against the requ-
lating rod. With the reactor on servo the regulating
rod then moved automatically a compensating
distance, this distance being roughly 10 in. of
its travel. From the previous measurement of the
worth of the regulating rod of (Ak/k)/in. of

174

0.033%/in. and the ratio of the movement of shim
rod No. 3 and the regulating rod, a calibration of
No. 3 shim rod was obtained over the first 14
inches of its travel (starting from the out positicn).

Figure J.2 shows the results of this method of
calibration (dashed curve), and the data are listed
in Table J.4. The agreement between the two
methods of calibration was surprisingly good.
Since the increments of shim rod movement were
larger in this method thdn in the preceding method
and since there were other sources of error, such
as the error in the movement of the shim rod
position indicators on the reactor console, no
attempt was made to use this curve (dashed curve,
Fig. J.2) to find the integrated value of Ak/k over
the red,

 

 

w

! ST O et b7 T W Fpimsrp RO T

P

-~

Fop

e s e ey e — -

 

T T wgmm—— P

pt' e w R e we

 
“ralf e

QL1

 

 

 

 

 

Formula 3: (Ak/E)/in. = —0.00454% + 0.1655d — 1.221
b|ntegrc[ formulas were of the form Ak/k =A’d° + B4 + C’d

Formula 4: Ak/k =~0.00024° + 0.0093542 + 0.0239d

Formula 5: Ak/k =0.0001674° + 0.0027542 + 0.059d

 

T TS W T P S

Ranges: Formulas 1 and 4, d=0to 8 in.
Formulas 2 and 5, d = 8 to 16 in,
Formulas 3and 6, 4 =16 to 18 in.

4T otal Ak/k, adding area from each section

 

c ) ke k. A ey Sl la G we AR R MR, e+ Kk i W e e o e e s o R MR L - ek e weiok E e & il Ges L b e ok WG AR
3 . + ¥ »
TABLE J.3. CALCULATION OF SHIM ROD REACTIVITY FROM FUEL ADDITION
Integral Calibration
C
Insertion of Differential Calibration O:fer}:frq;;glz Tofclj
Shim Red, d .
(in.) d? 43 Formula Ad? Bd (Ak/k.)/m. Formula? A’d3 B’d? C’d Ak} “ .é.k.;}
(%/in.) — .
b 0
% %)
(%)
1 ] 1 1 —0.0006 0.0187 0.0420 4 —-0.0002 0.00935 0.0239 £.0331 0.0331
2 4 8 1 ~0.0024 0.0374 0.0589 4 -0.0016 0.0374 0.0478 0.0836 0.083%
3 9 27 1 ~0.0054 0.0561 0.0746 4 —-0.0054 0.0842 0.0717 0.151 0.151
4 16 64 1 —-0.0096 0.0748 0.0891 4 —0.0128 0.1496 0.0956 0.232 0.232
5 25 125 1 ~0.0150 0.0935 0.102 4 —0.0250 0.2238 0.1195 0.328 0.328
6 36 216 1 -0.0216 0.112 0.114 4 -0.0432 0.33466 0.1434 0.437 0.437
7 49 343 1 —0.0294 0.131 0.125 4 —-0.0684 0.4581 0.1673 0.557 0.557
8 64 512 1 -0.0384 0.150 0.135 4 —0.1024 0.5984 0.1912 0.687 0.687
2 +0.032 0.0055 0.135 5 +0.0855 0.1760 0.4720 0 0.687
9 81 729 2 0.0405 0.0495 0.149 5 +0.1217 0.2228 0.5310 0.142 0.829
10 100 1000 2 0.0500 0.0550 0.164 5 0.1670 0.2750 0.5900 0.299 0.986
11 121 1331 2 0.0605 0.0605 0.180 5 0.2223 0.3328 0.6490 0.470 1.158
12 144 1728 2 0.0720  0.0660 0.197 5 0.2886 0.3960 0.7080 0.659 1.346
13 169 2197 2 0.0845 0.0715 0.215 5 0.3669 0.4648 0.7670 0.865 1.552
14 196 2744 2 0.0980 0.0720 0.234 5 0.4583 0.539%90 0.8260 1.117 1.804
15 225 3375 2 0.1125 0.0825 0.254 5 0.5636 0.6187 0.8850 1.334 2.021
16 256 4096 2 0.1280 0.0880 0.275 5 0.6840 0.7040 0.9440 1.599 2.286
3 -1.152 2.648 0.275 6 —6.144 21.184 ~19.536 0 2.286
17 289 4913 3 -1.300 2.8135 0.292 6 -7.3695  23.9147 -20.757 0.3112 2.597
18 324 5832 3 ~1.458 2.979 0.300 6 —8.7480  26.811 -21.978 0.612 2.898
Differential formulas were of the form (Ak/k) in. = Ad? + Bd + ¢ Formula 6: Ak/k = ~0.00154% + 0.082754% - 1.221d
Formula 1: (Ak/k)/in. = -0.0006d42 + 0.0187d + 0.0239 cd] = lower limit of formula
Formula 2: (Ak/k)/in. = 0.000542 + 0.00554 + 0.059 d, = position integrated o

T T T O T e
 

e coel Somiimes L b

 

 

 

 

 

 

S
TABLE J.4. CALIBRATION OF SHIM ROD NO. 3 AGAINST REGULATING ROD
Experiment L-4
Shim Rod  Shim Rod Shim Rod No. 3 Regulating Rod Sh::: R;d - 5“:1"; R3°d
Ron Mool o Nee2 s o Step  Avewgs Movemant St o Stop | Mowmert  (Mfiy/in,  Aversse
e iy s Gm Gmd Gy Gny ey OO R/ neerer
(% /in.) (in.)
1A-1B 35.0 36.0 30.0 33.3 3165 3.3 13.0 3.0 10.0 0.100 4.35
1C-1D 35.0 36.0 30.0 18.1 29.05 1.9 6.5 134 6.9 0.120 6.95
3-4 27.0 266 35.5 30.6 33.05 4.9 3.1 13.2 10.1 0.068 2.95
5-6 28.0 274 30.6 28.0 29.3 2.6 3.0 13.1 10.1 0.128 6.7
7-8 29.1 28.9 28.0 26.0 27.0 2.0 3.0 13.2 10.2 0.168 9.0
9-10 30.0 30.0 26.0 24.2 25.1 1.8 2.95 13.0 10.05 0.183 10.9
11-12 32.8 31.9 24.2 22.7 23.45 1.5 2.9 13.1 10.2 0.224 12.55
13~14 35.0 36.0 22,7 214 22.05 1.3 3.2 13.2 10.0 0.254 13.95
wr e e v wmer wemens s e wp v v e rree oyt

BIXC RO PETCONET EEE S v ommeTs e ECREUE T OTUFTC . "o N

 
b

Sl o

vk

il

ORN[-LR-DWG 6422

3.4

3.2

30

SHIM ROD REACTIVITY (A.A’//() OF SHIM ROD (%}
o - - - - - N N N no o
© o ™ S o @ o N s o @

o
o

0.4

0.2

 

0 2 4 6 8 10 12 14 16 18 20
TOTAL INSERTION OF SHIM ROD (in.)

Fig. J.3. Integral Calibration of Shim Rods as

a Function of Position.

CALIBRATION BY USING THE FISSION
CHAMBERS

A third calibration of the shim rods was at-
tempted during the critical experiments in which
the counting rate of the neutron detectors was
taken as a function of shim rod position for two
different uranium concentrations.

The reactivity was then obtained from the
counting rates by using the relationship

1
=1 - —,
M

and a plot of & as a function of rod position then
gave a check on the general shape of the curve.
Figure J.4 shows the £ vs rod position for fission
chambers 1 and 2. Because the uranium concen-
tration in the system was low for both the runs
the fission chambers were showing a subcritical

ORNL~-LR-DWG 6423

 

0.7 ! J- P

} i
U235 N SYSTEM=197 1b
; M>
=
|
I

    
    
  
  

]

I
U23d |y SYSTEM =164 1b

 

     

 

 

 

 

 

 

 

 

e
06 T{/i *
; r
FISSION ; 23‘5 I
CHAMBER #1235 |\ SYSTEM =166 Ib
= i '
2 0.54 :
5 “FISSION CHAMBER NO.2
i |
A ‘ I ‘
s | |
U235 N SYSTEM=19.7 1b
04 | : |
0/+ !
03
0 5 10 15 20 25 30 35 40

SHIM ROD POSITION (in.)

Fig. J.4. Calibration of Shim Rods from Fission
Chamber Data,

ORNL-LR-OWG 6424

 

0.2

Q.15

04

 

3(ak/kY/in(%/in)

 

0.05

    

FISSION CHAMBER NO. 1 DATA
| U235 |y SYSTEM: 197 1b

 

 

 

 

 

 

 

1@t p\'

 

 

o 5 10 15 20 25 30 35 40
SHIM ROD POSITION {in.)

Fig. J.5. Reactivity as o Function of Shim Rod
Position,

multiplication M greater than the actual value, as
discussed in Appendix E. Therefore, the absolute
values of &, as calculated from the counting rates,
are in error. Nevertheless, the general shape of
the curve is experimental verification of the curve
shown in Fig. D.3 of Appendix D, Figure J.5
shows the value of (Ak/k)/in. vs rod position,

177

TR T MR v

 

k.
b

 
 

as obtained from one of the curves of Fig. 4.10.  but its absolute magnitude is not meaningful,
Again, the general shape of this curve is about The data from which Figs. 4.10 and 4.11 were
the same as shown in Fig. D.3 of Appendix D, plotted are given in Tables J.5 and J.6.

 

 

 

 

 

i TABLE J.5. REACTIVITY OF THE SHIM RODS vs ROD POSITION g
;, y23s Rod Fission Chamber No, 1} Fission Chamber No, 2 : BF; Counter E
4 ° i
é in System Position Counting Rate Counting Rate Counting Rate i
o (1b) (in.) N (counts/sec) NO/N k N (countgs/sec) NO/N £ N (counts/sec) NO/N k -
: 16.6 0 17.52 0.5457 0.4543 25.00 0.6512 0.3488 6.00 0.7483 0.2517 g
f 5 17.60 0.5375 0.4625 25.52 0.6379 0.3621 6.05 0742 0,258 ;
] 10 18.85 0.5019 0.4981 26.69 0.6100 0.3900 6.51 0.690 0.310 ‘
: 15 20.93 0.4520 0.5480 31.33 0.5196 0.4804 6.40 0.702  0.298 E |
4 20 23.44 0.4036 0.5964 35.63 0.4569 0.5431 6.67 0.673 0.327
25 25.23 0.3750 0.6250 38.91 0.4184 0.5816 7.01 0.640 0,360 E |
30 26.00 0.3638 0.6362 40.93 0.3978 0.6022 E |
30 25.87 0.3657 0.6343 41.12 0.3952 0.6048 6.72 0.668 0.332
35 26.83 0.3526 0.6474 41.76 0.3898 0.6011 6.80 0.660  0.340 %
‘ 35 26.85 0.3523 0.6477 41.65 0.3909 0.6091 6.77 0.663  0.337 i
19.7 0 18.99 0.4982 0.5018 26.83 0.6068 0.3932 6.27 0.716  0.284
5 19.31 0.4899 0.5101 26.88 0.6057 0.3943 6.43 0.698 0.302 E
10 21.07 0.4490 0.5510 29.28 0.5560 0.444 6.56 0.684 0.316 B
15 23.63 0.4003 0.5997 34.43 0.4728 0.5272 6.91 0.650  0.350 &' '
20 26.21 0.3609 0.63N 38.99 C.4175 0.5825 7.04 0.638 0.362 E
25 28.08 0.3369 0.6631 42.67 0.3815 0.6185 7.28 0.617 0.383 - f
27.5 29.23 0.3236 0.6764 44.59 0.3651 0.6349 7.15 0.628 0,372 i
30 29.55 0.3201 0.6793 45.97 0.3541 0.6459 7.01 0.640  0.360 i
32.5 29.76 0.3179 0.6821 46.19 0.3525 0.6475 7.04 0.638 0.362 .
35 29.76 0.3179 0.6821 46.83 0.3476 0.6524 7.09 ©0.633  0.367 . :
36 29.90 0.3164 0.6836 46.81 0.3478 0.6522 7.16 0.627 0.373 ;

 

 

178

e g

 
G i

= iy

bbbk

ke el

k.

TABLE J.6. REACTIVITY AS A FUNCTION OF SHIM ROD POSITION (FROM FISSION CHAMBER DATA)

 

y235

in System

 

 

(Ib) Limits of d d_, k_, Ak Ad (in.) 3(Ak/R)/in,

19.7 0 to 2 1 0.502 0.04 2 0.0398
2 to 4 3 0.5065 0.05 2 0.0494

4106 5 0.5125 0.08 2 0.0780

61to8 7 0.523 0.12 2 0.1147

8 to 10 9 0.532 0.16 2 0.1503

10 to 12 " 0.555 0.20 2 0.1801

12 to 14 13 0.574 0.21 2 0.1829

14 o 16 15 0.596 0.22 2 0.1853

16 to 18 17 0.6155 0.18 2 0.1462

18 10 20 19 0.632 0.16 2 0.1266

20 to 22 21 0.6465 0.13 2 0.1005

22 to 24 23 0.6575 0.09 2 0.0684

24 to0 26 25 0.6655 0.07 2 0.0526

26 to 28 27 0.6715 0.05 2 0.0372

28 to 30 29 0.676 0.04 2 0.0296

30 1o 32 31 0.6795 0.03 2 0.0221

32tc M4 33 0.6815 0.01 2 0.0073

34 to 36 35 0.6825 0.01 2 0.0073

179

 

 

T

 
b

G

 

 

 

Appendix K
CORRELATION OF REACTOR AND LINE TEMPERATURES

An effect which was noted late in the operation
of the ARE was that fuel and sodium line tempera-
tures indicated in the basement disagreed quite
radically with those recorded in the control room.
It so happened that all the basement indicators
gave line temperatures measured by thermocouples
outside the reactor thermal shield, while the
temperatures recorded in the control room were
measured by thermocouples all located within the
thermal shield. When this was first discovered it
was thought that helium from the rod-cooling system
blowing on the thermocouples inside the thermal
shield was making them read low, Turning the rod
cooling blowers on and off, however, was demon-
strated to have no effect on the thermocouples, and
therefore at the end of the experiment no positive
explanation for these temperature discrepancies
had been found. .

This situation has created serious difficulties in
trying to analyze the data for this report. There
was evidence that the discrepancies were intensi-
fied during operation at high power. Nevertheless,
when nearly isothermal conditions existed, it was
found that the absolute magnitude of the measure-
ments made by the thermocouple within the thermal
shield were incorrect, but the rates of change
obtained from the data were correct, The tempera-
tures read on the basement instruments were
considered to be correct for several reasons, First,
the temperature indications available in the base-
ment were much more numerous than those in the
control room, and the thermocouple sensing elements
for these indicators were all located along lines
outside the thermal shield; the temperatures agreed
with each other to 110 deg from the average. Equi-
fibrium conditions prevailed across the pipes be-
cause of the insulation, Also, there were many
reasons for the thermocouples to indicate higher
than actual temperatures, but none for lower than
actual indications.

In order to use the temperature data from the
experiment it was necessary fo correlate the tem-
perature data obtained in the control room with
those obtained in the basement, The correlations
were then used to correct temperature data ob-
tained in the control room.

The results of the temperature correlations for
the fuel system are shown in Figs. K.1, K.2, and
K.3. The agreement between the low-power line

180

temperatures (Fig. K.1) reflects the attention which
was given to calibrating these thermocouples in
the isothermal condition. The curves in Figs. K.2

ORNL—LR—DWG 6563

 

1350

REACTOR
INLET LINES ———

 

 

1300 é{
A

- — REACTOR
OUTLET LINES

 

 

 

FUEL LINE TEMPERATURES INSIDE THERMAL SHIELD (°F)

 

 

 

 

1250 1
/4
O OUTLET TEMPERATURé
® NLET TEMPERATURE
1200 fl
1200 1250 1300 1350

FUEL LINE TEMPERATURES QUTSIDE TRERMAL SHIELD (°F)

Fig. K.1. Correlation Between Fuel Line Tem-
peratures Measured Inside and Outside the Reactor
Thermal Shield During Subcritical and Low-Power
Experiments,

ORNL—-LR~DWG 6564

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

1500
. /./
|, o 1450 _ L
y L d
5o REACTOR OUTLET LINE —u| ~*
% o 1400 t
5=
| /
w ™ /
b
S 1300} . 1
52 < ;
L2 / REACTOR INLET LINE ?
- {250 7 - - - !
1200

 

1200 {300 1400 1500 1600
FUEL LINE TEMPERATURES OUTSIDE THERMAL SHIELD {°F)

Fig. K.2. Correlation Between Fuel Line Tem-
peratures Measured Inside and Outside the Reactor
Thermal Shield During High-Power Experiments.

e e ger ot e 1

R

" e oo e

e e

e s——

 

wererry 7-m~”r * e

S € - grEA S B R e ey Py err W per T REowerw

Ce oy d

&
k
.
[

 
. Bk

LA,

G B g

e e,

ke SR

e,

ookl Bo ko, o Wk

i, G R e,

B,

ol S, @i

ORNL-LLR-DWG 6565

./

250 &

7

150 /
100 /+/
v

 

300

 

 

 

 

 

50 /’
0 50 10C 150 200 250 300 350 400
FUEL LINE A7 OQUTSIDE THE THERMAL SHIELD {°F)

 

 

 

 

 

 

 

 

 

FUEL LINE A7 INSIDE THE THERMAL SHIELD (°F)

 

Fig. K.3. Correlation Between Fuel Temperature
Differentials Measured Inside and Outside the

Thermal Shield.

ORNL-LR-DWG 6566
1350

 

REACTOR INLET LINE

1300 b

 

 
 

ZREACTOR CUTLET LINE

 

1250 / /

 

 

 

 

SODIUM LINE TEMPERATURES INSIDE THERMAL SHIELD (°F)

1200

 

1250 1300 1350 1400
SODIUM LINE TEMPERATURES QUTSIDE THERMAL SHIELD (°F)

Fig. K.4. Correlation Between Sodium Line
Temperatures Measured Inside and Outside the
Thermal Shield During Low-Power Experiments.

and K.3, both of which were obtained from data
taken during operation at high (>200 kw) power,
show the anomaly in question,

The analogous data for the sodium system are
shown in Figs. K.4, K.5, and K.6, It is evident
from Fig. K.4 that an isothermal condition was
never attained in the sodium system, since both

 

 

 

 

 

 

 

 

 

o
0
o ORNL-LR-DWG 6567
o 1350
: | |
N
2 REACTOR OUTLET LINE
g
9
Léj / /-’
4300
o o /
o
= /
$ *
g /
— O
I 41250 #
w
o..
=
w
= REACTOR INLET LINE
=
- | |
s 1200
g 1200 1250 1300 1350 1400
R SODIUM LINE TEMPERATURES OUTSIDE THERMAL SHIELD (°F)

Fig. K.5. Correlation Between Sodium Line
Temperatures Measured Inside and Ovutside the
Thermal Shield During High-Power Experiments,

ORNL-LR—-DWG 6568
150

 

 

{100

 

50 e ®

SODIUM LINE A7 INSIDE THE THERMAL SHIELD(®F)

 

 

 

 

 

0
0 50 100 {50
SODIUM LINE A7 QUTSIDE THE THERMAL SHIELD (°F)

Fig. X.6. Correlation Between Sodium Temper-
ature Differentials Measured Inside and Outside

the Thermal Shield.

the inlet and the outlet line temperature correla-
tions have slopes of 1 but intercepts of ~25 and
—~40°F, respectively. This was probably the re-
sult of the long sodium inlet and outlet lines being
used as a convenient means of adding heat to the
system to maintain a thermal equilibrium in the
reactor of ~1300°F. Good data for the high-power

181

T P Y

 

TR O T R e T TR e

¢
f
b

 
3

 

T

 

 

correlations, Figs. K.5 and K.6, were rather scarce,
but the curves shown are believed to be reasonable
approximations to the actual correlation, The
data in Fig. K.6 have been corrected for known
error in the temperature differential at zero (or low)
power. The correction, as applied, was to reduce
the value of the temperature differential determined
from the thermocouples outside the thermal shield
by 10°F,

All the correlations were based upon data which
were obtained during runs long enough for the
establishment of equilibrium conditions in the
system or at times when isothermal conditions
prevailed. The most important conclusion that can

‘be made from these data is that the temperature

differentials across the reactor in both the fuel
and sodium systems were about a factor of 2 {ow.
In the fuel system the outlet line temperature
changes read in the control room (inside pressure
shell temperatures) were only one-half as great as
the corresponding temperature changes read in the

basement (outside pressure shell temperatures)., In
the sodium system both outlet and inlet line tem-
perature curves had different intercepts but the
same slopes (as during the low-power runs).

All temperature data used in this report that
pertained to operation at high power were corrected
to agree with line temperatures obtained outside
the thermal shield (basement readings) by using
curves K.l through K.6, Data from runs made
under isothermal or equilibrium conditions where
equal temperature differences were obtained on
instruments either in the control room or in the
basement needed no temperature corrections. The
temperature correlation data from which the curves
were obtained are presented in Tables K.1 and
K.2. Because of the lack of coordination between
the control room and basement operations during
the experiment, much potentially useful data had
to be discarded because the exact times of the
readings were not known or the data were taken
before equilibrium conditions were established,

TABLE K.1. FUEL LINE TEMPERATURES MEASURED BY THERMOCOUPLES INSIDE AND
~OUTSIDE THE REACTOR THERMAL SHIELD

 

Fuel Inlet Line

Experiment Temperatures (°F)

Date Time No. (Line 120)

Fuel Outlet Line
Temperatures (°F)

(Line 111)

Temperafure

D ifferential (o Fy©

 

Basement? Control Room

b

Basement® Control Room? Basement Control Room

 

10/26 2115 Before 1290 1285
operation

10/27 0630 Before 1293 1293
operation

11/4 1100 L-1 1300 1299

11/6 0340 L-4 1305 1310

1111 0300 H-11 1207 1212

1215 H-11 1210 1214

11/12 0001 H-13 1316 1308

0630 H-13 1315 1307

1003 H-14 12464 1243

11/13 0915  After 1264 1260

operation

1290 1287 0 2
1296 1293 3 0
1297 1295 -3 -4
1304 1300 ~1 -10
1522 1418 315 206
1524 1418 314 204
1323 1308 7 0
1320 1306 5 =1
1587 1475 341 232
1260 1250 -4 ~10

 

2Fuel line temperatures outside thermal shield.

bFuei line temperatures inside thermal shield.

€ Thermocouple differences for no-power runs indicate extent of thermocouple errors (highest is £10 deg) and therefore

the accuracy of the readings.

dEsfimcfed.

182

 

L
:

IR F

“ETm e S ey Ny O B TETT 0 b tep T

wt e emerrr gy *ymm'w o W ¥ e TP v ey s’ ¥

 
gl e Ve

e

e i b s MGG R e uma i, Sk ok lelee s il ca e e

S 4

o W ot

Fdtle

TABLE K.2, SODIUM LINE TEMPERATURES MEASURED BY THERMOCOUPLES INSIDE AND
OUTSIDE THE REACTOR THERMAL SHIELD

 

 

 

Sodium Inlet Line Sodium Outlet Line Temperature
Date Time Experiment Temperatures (°F) Temperatures (°F) Differential (°F)
No.
Basement Control Room Basement Control Room Basement Control Room

10/26 2115 Before 1305 1278 1320 1278 15 0

operation
10/27 0630 Before 1312 1292 1325 1290 13 ' 2

operation
11/4 1100 L-1 1320 1295 1330 1288 10 7
11/6 0400 L-4 1325 1300 1335 1295 10 5
11/11 0900 H-11 1223 1224 1332 1271 109 47
17/12 1003 H-14(1) 1246 1258 1369 | 1307 123 49
11/9 1907 H-3(7) 1313 1290 1373 1313 60 23
11713 0915 After 1282 1258 1292 1252 10 -6

operation

 

This points out the need in future operations of the
scale of the ARE for some sort of timing device
that would automatically stamp the date and time
every few minutes on the charts of all recording
instruments, and a more systematic method of

manually recording data for each experiment per-
formed in which conditions are held constant long
enough for all pertinent data to be recorded. These
and other similar recommendations will be found
in Chapter 7 of this report.

183

 

 

 

 

 
 

 

 

 

 

Appendix L
POWER DETERMINATION FROM HEAT EXTRACTION

The power level of the reactor was determined
from the total heat extracted by the fuel and the
sodium. Data for the heat extraction determination
were obtained from continuous records of the re-
actor fuel inlet, outlet, and mean temperatures,
the temperature differential across the reactor,
and the flow rates of the fuel and the sodium.
The flow rates could also be calculated from the
pump speed, and there was independent experi-
mental evidence from which a plot of speed vs
flow rate was obtained, The power level of the
reactor was calculated by use of heat capacities,
flow rates, and temperature differentials for both
sodium and fuel, it was found that the fuel and the
sodium accounted for 99% of the extraction of
power generated in the reactor. The data referred
to above were all obtainable in the control room.

The controls for preheating and maintaining heat
on the system were located in the basement, and
along with these controls were temperature re-
corders and indicators for the whole system, ex-
clusive of the reactor. The temperatures of both
fuel and sodium lines to and from the reactor were
also recorded in the basement. These basement
data were used for a separate power extraction
determination,

An independent source of power level information
was also available, since the fuel and sodium were
cooled by a helium stream flowing over a heat ex-
changer and the helium in turn was cooled by a
helium-to-water heat exchanger., No flow or tem-
perature measurements were made of the helium,
but since the water exchangers were very close to
the fuel or the sodium exchangers, all the heat that
was taken out of the fuel and sodium should have
appeared in the water, The water flow through
these exchangers was metered by orifice-type
flowmeters, and the outlet temperatures were meas-
The fuel loop
exchangers and the two sodium loop exchangers
were metered separately.

Up to the time of the 25-hr xenon run the reactor
was operated at high power for very short periods
of time, and the extracted power was determined
from the control room data. During the 25-hr xenon
run a comparison was made of the extracted power
determined from the control room datq, that de-

ured by thermocouples in wells.

termined from the basement data, and that from
the water data (Table L.1). Large discrepancies

184

were found, and the indications were that the
control room data were low.

The disagreement of the various data led to the
examination of the inlet and outlet line tempera-
tures, as described in Appendix K, During the
xenon run the fuel outlet line temperatures were
about 100°F higher than the fuel outlet manifold
temperatures (control room data), and the fuel
inlet line temperatures (basement data) were
essentially the same as the fuel inlet manifold
temperatures (control room data). The sodium
outlet line temperature {(basement data) was 60°F
higher than the sodium outlet temperature at the
reactor (control room data), (The only check on
power extraction by the rod cooling system, which
was only 1% of total power, was the water dota
for the rod cooling helium-fo-water heat exchanger.
In most discussions this 1% is neglected.)

The power extracted in both the fuel and the
sodium systems has been calculated by using
temperature data from several experiments, and the
results are presented in Table L.2. There were
three separate measurements of the fuel temperature
differential available in the control room which
were usually in fair agreement. The largest dis-
crepancy was, as mentioned, between the control
room data and the basement data. The temperature
differential obtained from the basement data was
an average result from a number of line thermo-
couples, while the control room data came from
thermocouples located on the fuel headers inside
the reactor thermal shield.

The so-called secondary heat balance was ob-
tained by determining the heat dumped by the water
from the various heat exchangers. These data are
tabulated in Table L.3 for the high-power experi-
ments.

The various estimates of reactor power from
both primary and secondary heat extraction are
then listed in Table L.4, together with the power
estimated from the calibration of the nuclear in-
struments, that is, the log N recorder and the micro-
microammeter, These latter instruments were
normalized to agree with the primary extracted
power as determined by the line temperature
(basement data) during the xenon experiment (H-11).
Equilibrium conditions were certainly attained
during this 25-hr experiment and such error as there
may be in using these data as the criteria for

n

 

£
»
&
>
B
.

o

T NPT W TR T pM g Wear s YT P > meEw o oo

gy o

*

 

x TR T DET (R ¢ BT e T P L MY T g S e I T
Wk R Gk m;‘\

TABLE L.1. COMPARISON OF POWER EXTRACTION DETERMINATIONS MADE FROM DATA
OBTAINED DURING 25-hr XENON RUN

 

.

 

 

sl da-

vy,

Ll il w

 

Temperatures (°F) Flow Power Extracted Total Power? Extracted L
Inlet Qutlet AT (gpm) (Mw} (Mw)

Fue! System ‘
Control room data 1212 1418 206 44 1.02 |
Basement data 1209 1522 313 44 1.52 i

Sodium System E
Control room data 1225 1271 46 153 0.244 ‘
Basement data 1226 1335 109 153 0.577

Total Power
Control room data 1.28
Basement data 2,12

Water Data
In fuel loop 61 117 56 205 1.68 :
In sodium loop 61 114° 53 38.3% 0.588° ‘

Total Power from 2.28
Water Data

 

S okl

il nl R AR AL ¢

s

%|ncludes ~ 1% for rod cooling.
bAveroge for both loops.
€Sum of both loops.

TABLE L.2, PRIMARY POWER EXTRACTION

 

 

 

 

 

 

 

B BB e i

Contre] Room Data Basement Data
Average
Experiment Run Fuel System Sodium System TT[:‘::UZ‘E‘WGT Fuel System 2;:;:: T?’:::UZ?:;r k
N o AT,Outlet AT, AT, Tube o b . Rod-Cooling AT b AT P Rdfifximg
o Minsinler Recorder Aveiese (b iy (o o) CF) 0N e o ok em o e ]
H.3 14l 40 33.5 0.196  0.163 15 0 0.199 ¥
2 44l 66 52 0323 0.254 0 0.308 1
3 44, 15 116 0.562  0.567 0 0.584 ;
4 4y 120 14 124 0588 0.560 0.606 153 14 00734 0678 198 0970 45 0.236 123 :
5 4l 2% 231 29 1128 1129 147 153 105 00548  1.21 :
6 4l 2y 234 23 106 1045 119 12 10 00523 124 138 0677 60 0315 1012
7wy 13 19 20 0.0653 0.0946 00975 152 22 0115 0221
H-6 a4l 26 240 245 1158 1075 LIsg 152 3.5 00182 1.2 320 1.572
H8 45 266 225 131 L 153 21 0110 134 293 1452 75 03%4 187 E
K11 44 206 213 21 0984 1017 1018 153 464 0244 127 313 1520 10 0.577 2,12 b
H-13 44 7 0  0.0484 153 1.4 00073 00757 312 1.530 19 0099  1.45 1
H-14 45 182 245 0.900 1.212 153 50 0262 145 355 1760 127 0.667 245 '
L
|
{
185

 
 

 

 

TABLE L.3. SECONDARY POWER EXTRACTION

 

 

 

 

 

 

 

 

E No. 1 Sedium No. 2 Sodium Rod-Ceoling
U?I Heat Exchanger Heat Exchanger Heot Exchanger Heat Exchanger Total
Experiment  Run Extracted Power
No., No,  Water AT Pioerr  Water AT Py Water AT Pyo  Water AT Pr.c. Secondary System
: Flow (°F) ower  Flow (°F) Power  Flow ©F) Power  Flow (°F) Power (Mw)
(gpm) (Mw)  (gpm) (Mw)  (gpm) (Mw)  (gpm) (Mw)
H-3' i 204 13 0.389 38.4 0 0 38.4 0 0 17.3 3 0.0076 0.397
2 204 13 0.389 38.4 0 0 38.4 0 0 17.3 3 0.0076 0.397
3 204 30 0.897 38.4 0 0 38.4 0 0 17.3 3 0,0076 0.905
4 204 29 0.866 38.4 26 0.146 38.4 21 0.1188 173 3 0.0076 1.14
5 204 59 1.765 38.4 26 0.146 384 25 0.141 17.3 3 0.0076 2.06
6 204 59 1.765 38.4 27 0.152 38.4 24 0.135 17.3 3 0.0076 2.06
7 204 3 0.0903  38.4 26 0.146 38.4 25 0.141 17.3 2 0.0051 0.382
H-6 204
H-8 206 63 1.901 38.0 37 0.207 38.4 40 0.226 17.3 3 0.0076 234
H-11 205 56 1.685 38.2 55 0.300 38.4 51 0.228 17.6 2 0.0052 2,28
H-13 205 3 0.0902 38 10 0.0561 38.4 1 0.0056 17.1
H-14 204 63 1.883 38 57 0.318 38.4 56 0.316 16.9 6 0.0150 2.53
TABLE L.4. REACTOR POWER SUMMARY
Primary Power (Mw) Secondary
Nuclear Power Power (Mw);
Experiment  Run (Mw) Control Room Data Basement Water
No. No. Data, Heat Exchanger,
Log N Micromicroammeter PAT Pp Pav AT L Pp Py
H-3 1 0.449 0.425 0.216 0,183 0.199 0.3966
2 0.449 0.463 0.343 0.274 0.308 0.3966
3 0.848 0.910 0.582 0.587 0.584 ¢.9046
4 0.923 0.988 0.681 0,653 0.699 0.678 1.23 1.1384
5 2.47 2.05 1.20 1.20 1.24 1.21 2.0596
6 2.12 2.34 1.23 1.22 1.26 1.24 1.012 2,0596
7 0.281 0.288 0.200 0,230 0.232 0.221 0.3824
H-6 2.21 2,16 1,20 1,21 1.25 1.22 1.93
H-8 2.12 2.38 1.44 1.24 1.34 1.87 2.342
H-11 2.12 2.12 1.25 1.28 1.28 1.27 2.12 2.278
H-13 0.125 0.151 0.0757 0.0757 0.1562
H-14 2.41 2,53 1.42 1.45 1.43 2.45 2.53

 

amount of extracted power is conservative, since
the water heat balance at the same time showed
the power to be 7% higher.

The sources of error in the various measurements
of extracted power were associated with the tem-

186

perature measurements on the fuel and sodium lines
at the reactor, which gave the reactor AT, These
thermocouples were unique in several respects;
that is, they were located within the reactor thermal
shield; they were exposed to the reactor pressure

 

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shell; they were exposed to high nuclear radiation
fluxes, etc. These unique aspects immediately
suggest several possible explanations for the
erroneous temperature measurements. However,
upon further examination, each of these aspects,
with the dubious exception of nuclear radiation,
has been shown to be incapable of producing the
observed anomaly.

The control and shim rods and the fission
chambers were cooled by forced helium circulation
in the rod-cooling system, Part of the helium that
was blown down through the rod tubes was de-
flected back up across the outlet manifold and
between the pressure shell and the thermal shield.
While it was thought that perhaps this was cooling
the fuel outlet manifold thermocouples and making
them read low, this was disproved when changing
the speed of the blower or stopping it had no effect
on the fuel outlet temperature.

It was thought that possibly the heat radiation
from the fuel outlet manifold to the colder bottom
of the pressure shell was great enough to actually
lower the manifold wall temperatures 100°F below
the fluid temperature. This has been disproved by
heat transfer calculations. 1t was also thought that
since the outlet fuel lines passed through the
2-in, plenum chamber that the resulting surface
cooling of the fuel might account for the lower
wall temperatures, although the mixed mean fuel
temperature was considerably higher. This was
also disproved when no increase in wall tempera-
ture was observed for thermocouples within the
thermal shield but located progressively farther
away from the pressure shell bottom.

It is of interest that after the final shutdown of
the reactor, the fuel outlet line and fuel outlet
manifold temperatures agreed; in other words, with
no power generation the basement data and control
room data agreed.

187

 

T —————r

 

 

 
 

 

 

~ Appendix M |
THERMODYNAMIC ANALYSES

Some two years before the Aircraft Reactor Ex-
periment was placed in operation a comprehensive
report was written on the ‘‘Thermodynamic and
Heat Transfer Analysis of the Aircraft Reactor
Experiment.”'!  Not only was the design of the
reactor system based, in part, on the studies culmi-
nating in that report, but the material therein
served as a guide during the acceptance fest of
the experiment, Therefore a comparison of the
calculated and the actual thermodynamic per-
formance of the reactor system was attempted, [t
soon became apparent that the experimental data
of the type needed for thermodynamic analyses
were inadequate to permit a meaningful comparison
with any of the calculated situations. Perhaps the
most valid comparison may be made between the
calculated insulation losses, the heater power
input at equilibrium, and the heat removed by the
space coolers. Even here the agreement is less
than 50%. An illustration of the inefficacy of
attempting to calculate such thermodynamic con-
stants as the heat transfer coefficients of the fuel
and the sodium is presented below.

INSULATION LOSSES, HEATER POWER INPUT,
AND SPACE COOLER PERFORMANCE

At equilibrium the electrical heat required to
maintain the system at o mean temperature of
1325°F should have been equivalent to the heat
removed in the eight pit space coolers. The maxi-
mum amount of heat ever removed by the space
coolers was the 250 kw attained with 66.5-gpm-total
water flow with a temperature rise of 30°F. Al-
though this value agrees very well with calculated!
heat loss of 220 kw for the entire system (the
reported value of 240 kw was calculated for an
earlier design system and was reduced by about
20 kw for the actual system), the actual electrical
power input to the heaters averaged around 375 kw
after the system reached equilibrium.

There are several possible explanations for the
discrepancies betwaen the calculated heat loss,
the electrical power input, and the heat removed
by the space coolers. First, the calculated loss
was low because it assumed idealized conditions

 

1B, Lubarsky and B. L. Greenstreet, Thermodynamic
and Heat Transfer Analysis of the Aircraft Reactor Ex-
periment, ORNL-1535 (Aug. 10, 1953),

188

and not the octual insulation which had cracks,
clips, etc. Second, the electrical heat load was
high, since it did not allow for transformer, variac,
and line losses. Also, the space cooler heat load
may not have represented an equilibrium condition
if the pit walls were still heating up and/or radi-
ating heat,

EXPERIMENTAL VALUES OF HEAT
TRANSFER COEFFICIENTS?

 An attempt was made tocalculate both the sodium
and the fuel heat transfer coefficients from the
measured values of temperatures and flows in the
various heat transfer media, that is, fuel, sodium,
helium, water, These data are given in Table M.}
for a typical operating condition.

The desired heat transfer coefficient should be
determinable from the calculated values of the
various thermal resistances in the heat exchangers.
To obtain the thermal resistance on the fuel side
or the sodium side of the respective heat ex-
changers, the helium side and metal resistances
had to be subtracted from the over-all resistance.
The expression for the over-all resistance is

1 q

. — =

! (UgA) A2 AT

’

where

U = over-all heat transfer coefficient,

A area across which heat is transferred,
q rate of heat transfer,

AT = over-all temperature difference,

H

I

A knowledge of the inlet and outlet temperatures
on the helium side of the heat exchanger is re-
quired by the above expression, and lack of this
information immediately introduces a large un-
certainty in the results, Of greater importance,
however, is the relative magnitude of the resist-
ances involved, The calculated ratio for the fuel-
to-helium exchanger is

RHe

 

— =103,
/

where RHe is the resistance on the helium side

 

2H. H. Hoffman, Physical Properties Group, Reactor
Experimental Engineering Division.

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TABLE M.1. EXPERIMENTAL HEAT TRANSFER DATA IN FUEL AND SODIUM LOOPS

 

 

 

Primary Coolant Helium Water
Temperature Temperature Temperature
*

Flow (°F) Flow (°F) Flow (°F)
(gpm) (cfm) —_— {gpm} —/——
Inlet Qutlet Inlet  Outlet Inlet Qutlet
Fuel loop 44 (¥5%) 1209210 1522 £10 8000 (£30%) No data 194 (+5%) 61+t2 11712
Sodium loop 152 (£1%) 1225 %10 1335110 1700 (£20%) No data 766 (£1%) 6112 11412

 

*Estimated.

and R, is the resistance on the fue! side, and for
the sodium-to-helium exchanger is

 

where R_ is the wall resistance, Hence, the
liquid-side thermal resistance is the small dif-
ference between two much larger numbers, A small
error in R.. or R, is greatly magnified in R,
Unfortunately, the error in R, is not small, being
perhaps as great as 60%., Thus, except for over-all
heat balances, little useful heat transfer informa-
tion can be extracted from the available data on
the ARE heat exchangers.

189

TR e

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PR WL TR T W

 

 

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Appendix N
. COMPARISON OF REACTOR POWER DETERMINATIONS
W, K. Ergen

In order to get an estimate of the reactor power
and to be able to calibrate the instruments, the
ARE was run for 1 hr ot a power which was esti-
mated, at the time, to be 10 watts (exp. L-4). A
sample of the fuel was then withdrawn and the
gamma activity was compared with that of a sample
which had been irradiated at a known power level
in the Bulk Shielding Reactor (BSR). This method
of power determination is discussed in Appendix H.

A curve showing the gamma counting rate of a
BSR-irradiated sample as a function of time after
shutdown had been obtained. The irradiation time
was | hr at a constant power of 1 w, From this
curve a decay curve corresponding to the actual
power history of the ARE was synthesized by
taking into account not only the *‘10-watt'’ run,
but also the previous lower power operation.

When this synthetic curve was compared with the
one obtained from measurements on the ARE sam-
ple, the shapes of the curves did not agree, the
ARE curve having a smaller slope than the syn-
thetic curve, Also, the power determined by this
method was lower than the power ultimately de-
termined by-the heat balance. Both these effects
can be explained in a qualitative manner by the
loss of some of the radioactive fission fragments
from the ARE sample. This would reduce the total
radioactivity of the sample and hence the apparent
power level. Furthermore, the loss of radioactive
fission fragments would reduce the counting rate
at short times after irradiation more than at long
times, because among the most volatile fission
fragments are the strong gamma emitters that have
relatively short half lives (notably, 2.77-hr Kr38),
This would flatten the slope of the decay curve
of the ARE sample. |

it has not been possible so far to treat this
matter in a quantitative way, but in order to elimi-
nate some computational complications a sample
was irradiated in the BSR under conditions exactly
duplicating the power history of the ARE, except
for a proportionality factor. A comparison of the
decay curve of this sample and that of the ARE
sample is shown in Fig. N.1. The BSR sample
contained 0.1166 g of U235: the fission cross
section at the temperature of the BSR is 509 barns, !
The BSR power at the final T-hr irradiation was

190

10 w, corresponding to a flux? of 1.7 x 108 n/cm?:sec,
and a self-shielding factor, estimated to be 0.8,
had to be applied to this sample. Hence, during
the 1-hr irradiation there were

0.1166 g x 0.6 x 1024 atoms/g-atom x 509
x 10~ 24 fissions/(atoms:n/cm?) x 1.7

x 108 n/cm?-sec x 0.8/235 g/g-atom

n

20.6 x 10° fissions/sec

The ARE sample contained 0.1177 g of U233 and
the total U?33 in the ARE at the time of the run
was 59.1 kg. The reactor power, as determined
fater by the heat balance, was actually 27 w,
instead of the expected 10 w, Hence there were

27 w x 3.1 x 100 fissions/wesec x 0.1177 g
59,100 g

 

= 1.66 x 10¢ fissions/sec

in the ARE sample.

The ratio of the radioactivities of the BSR
sample and the ARE sample should thus have been
20.6/1.66 = 12.4. As may be seen from Fig. N.1,
the measured ratio is 27 at 1 hr 40 min, and 17.5
at 38]/2 hr. As pointed out above, the discrepancy
can be explained by the loss of radioactive fission
fragments from the ARE sample.

A small amount of the radioactivity of either
sample was contributed by the capsule and the fuel
carrier, especially the sodium content of the
carrier. This was measured by irradiating a non-
uranium-bearing capsule with carrier, and counting
its radioactivity. However, since this correction
proved to be small and since both samples con-
tained about the same amount of uranium in the
sagme amount of carrier, the results would not be
appreciably affected.

 

13, L. Meem, L. B. Holland, and G. M, McCommon,
Determination of the Power of the Bulk Shielding Re-
actor, Part Ill. Measurement of the Energy Released
per Fission, ORNL-1537 (Feb. 15, 1954).

2E, B, Johnson, private communication.

 

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161

ACTIVITY OF SAMPLE (arbitrary units )

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SAMPLE

SAMPLE

 

10 15 20 25 30 35 40 45
TIME AFTER SHUTDOWN (hr)

Fig. N.1. Decay of Irradiated Fuel Samples.

 

 

 

T WIS e T My S ST e g e e e - e R Cw T o : T Ot
 

 

i Kb i s i o b ase e Aale i3

 

Appendix O
ANALYSIS OF TEMPERATURE COEFFICIENT MEASUREMENTS

'IMPORT‘ANCE OF THE FUEL TEMPERATURE
- COEFFICIENT

One of the most desirable features of a circu-
lating-fuel reactor is its inherent stability because
of the strong negative temperature coefficient of
reactivity. This was conclusively demonstrated
by the ARE. It had been predicted, on the basis
of the temperature dependence of the density,
which was

p (g/cm®) = 3.98 — 0.00093T (°F)

for the final fuel concentration in the reactor, that
the ARE fuel would have a negative temperature
coefficient of sizable magnitude. At the operating
temperature of 1300°F the fuel density was 3.33
g/em®. The resulting mass reactivity coefficient
was '

o ~0.00052

(AM/M)/°F = — ———= ~1.56 x 1074/°F ,

3.33

and, for Ak/k = 0.236 AM/M, the predicted temper-
ature coefficient that would result from the changing
density of the fuel alone was

(Ak/R)/°F = ~3.68 x 10~5/°F .

Actually, as was stated in the body of this report

(ef., Fig. 6.4), the fuel temperature coefficient was
~9.8 x 10-%/°F,

As long as the over-all temperature coefficient
of a reactor is negative, the reactor will be a slave
to the load demand. However, the fuel temperature
coefficient was the important factor for reactor
control in the ARE because it took a significant
time for the bulk of the material of the reactor to
change temperature and, therefore, for the over-all
coefficient to be felt. In experiment H-5 (cf., Fig.
6.4), it was found that the fuel temperature coef-
ficient predominated for 6 min.

EFFECT OF GEOMETRY IN THE ARE

When the reactor was allowed to heat up by
nuclear power, as in experiment H-5, the measure-
ments of the temperature coefficient were quite
definitive. The experiments in which the heat
extraction by the helium blower was suddenly
increased, i.e., experiments E-2, L-8, and H-4, did
not give clear-cut results. In fact, it was noticed
throughout the operation that whenever the fuel

192

loop helium blower speed was increased there was
always a sudden marked increase in reactivity. In
attempting to determine this instantaneous temper-
ature coefficient by observing the rod movement
by the servo as a function of the fuel temperature,
values of the fuel temperature coefficient of a
magnitude much larger than ~9.8 x 10™% could be
obtained.

The change in the apparent temperature coef-
ficient as a function of time during experiment H-4
is shown in Fig. O.1. The time intervals chosen
were of 15-sec duration. The apparent peak tem-
perature coefficient of reactivity was in excess of
~3.5 x 1074 (Ak/k)/°F, and after 5 min the coef-
ficient leveled out to a value of about ~6 x 1073,
A more complete discussion of this experiment is
given below in the section on ‘‘High-Power Meas-
urements of Temperature Coefficients of Reac-
tivity,"" in which the time lag of the thermocouples
is taken into account. Two coefficients were ob-
tained: an initial fuel temperature coefficient and
an over-all coefficient, which was the asymptotic
value (=6 x 10~3) given by Fig. 0.1,

The effect shown in Fig. O.1 was due partly to
geometrical considerations. By reference to Figs.
2.2 and 2.3, it can be seen that as the fuel entered
the reactor it passed through the tubes closest to

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“TIME OF DAY, NOV. 9, 1954

Fig. O0.1. Variation in Apparent Temperature
Coefficient of Reactivity with Time, Experiment

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the center where the reactivity effect was greatest.
With o sudden increase in heat extraction, a slug
of cooled fuel initially entered the center of the
core and caused a large reactivity change. Since
the mean reactor temperature was the average
temperature of all the fuel tubes, a comparison of
the initial rate of rod insertion by the servo with
the initial rate of change of the mean reactor
temperature could give an apparent temperature
coefficient much larger than its actual value. For
example, in experiment L.-8 {cf., Figs. 5.13 and
5.14) a comparison of the slopes of the red position
curve and the mean temperature curve near the
beginning of the run was made. If it is assumed
that the temperature response of the thermocouples
lagged behind the response of the regulating rod
movement, it is interesting to compare the points
of steepest slope for the two curves. By using
the slope from Fig. 5.13 at time 0218:30, and the
slope from the mean temperature curve (Fig. 5.14)
at 0220, a temperature coefficient of —1.6 x 10~ *
(Ak/k)/°F is obtained, as opposed to the actudl
fuel temperature coefficient of ~9.8 x 1073, A
detailed discussion of this experiment is given
in the section on ‘‘Low-Power Measurements of
Temperature Coefficients of Reactivity.” This
effect was characteristic of the geometry of the
ARE but not necessarily of circulating-fuel re-
actors in general.

TIME LAG CONSIDERATIONS

As mentioned above and in the “*Reactor Kinetics”’
section of Chapter 6, one of the most consistently
noted phenomena of the ARE operation was the
time lag in reactor témperature response during
every phase of the experiment. These time lag
effects can be partly explained by the geometry
effect just described; and, in this sense, the time
fag is not a true time lag but only an apparent lag
due to the differences in location and response of
the thermocouples and the neutron detectors (and,
hence, regulating rod movements ). Other possible
causes for the time lags were the fuel transit time
around the system, the heat transfer phenomenon
within the reactor, the design and location of
thermocouples, and a mass-temperature inertia
effect. These effects were all discussed briefly
in the section on ‘‘Reactor Kinetics,”” Chapter 6.

 

MThe regulating rod was controlled by o flux servo
mechanism which received its error signal from o
neutron detector (cf., App. C).

Whether or not the time lag was all or partly real
is academic. The fact that it did give an observed
effect during many phases of the ARE operation
made it mandatory to take the lag into account in
interpreting much of the data. The manner in which

this was actually done was to assume that the.

response of the thermocouples lagged behind the
nuclear response of the reactor (by as much as
2% min for low-power operation and by about 1 min
during the high-power regime) in those experiments
in which the equilibrium between the reactor and
its load was upset (i.e., rapid fuel cooling rates).
The thermocouple readings were then ‘“moved up"’
by that amount, and the new readings were com-
pared with the appropriate nuclear instrument
observation. For those experiments in which equi-
{ibrium prevailed, but in which cooling was taking
place, it was only necessary to correct for temper-
ature readings (cf., app. K).

The temperature coefficient measurements were
probably more affected by the temperature-time
lags than any other single type of measurement
made on the ARE, mainly because of the short
duration of the experiments and their great de-
pendence on time correlations (for example, corre-
lations between regulating rod motion and mean
temperature changes). The results of temperature
coefficient measurements which contained time lag
corrections were not included in the main body of
the report because such corrections needed to be
discussed in detail inappropriate to the context of
the report. These experiments are described in
the following sections of this appendix.

SUBCRITICAL MEASUREMENT OF
TEMPERATURE COEFFICIENT

The suberitical measurement of the temperature
coefficient {exp. E-2) was described in Chapter 4.
Briefly, the procedure followed was to cool the
fuel by raising the heat barriers on the fuel heat
exchangers, turning on the fuel helium blowers,
and then observing the increase in multiplication
with the two fission chambers and the BF; counter.
The BF, counting rate was so low that the sta-
tistics were poor; therefore, the fission chamber
data had to be used. The subcritical multiplication
of the fission chambers was then subject to the
phenomenon discussed in Appendix E. In this
experiment the cooling was rapid and equilibrium
conditions v_vé;e not attained. | Consequently, in
order to find a value of the temperature coefficient
from these measurements, it was necessary to

193

TR TRETTTSE T T W

¥ o

 

 
(170 T

 

 

apply three corrections: a correction for the fission
chamber multiplication error, a time lag correction,
and a temperature correction of control room ob-
served temperatures {app. K).

Throughout the subcritical experiments ample
data were taken simultaneously on the BF counter
and the two fission chambers for a correlation plot
between the counting rates of the BF, counter and
the fission chamber to be easily obtained, as
shown in Fig. 0.2. A plot of the raw data obtained
from the fission chambers before corrections were
applied is shown in Fig. 0.3,2 which shows the
counting rates of the chambers plotted as a function
of the reactor mean temperature. The hysteresis
effect is the result of the time lag. The progress
of the experiment can be read from the curves by
starting on the right side at 1004 and proceeding
counterclockwise around the loops. The fuel
blower was turned on at 1004 and allowed to cool
the fuel for 5 min, after which the blower was turned
off and the system then slowly returned to its initial
condition. The fission chamber counting rates
increased while the blower was on and decreased
after the blower was turned off again in immediate
response to the cooling. The reactor mean temper-
ature change, on the other hand, iagged behind
both when the blower was turned on and when it
was turned off. The blower was turned off shortly
after 1009, but even though the counting rate
started to decrease immediately, the reactor mean
temperature continued to fall for approximately
2% min before it began to show a warming trend.
Undoubtedly the fuel temperature did actually
follow closely the changes introduced by the
blower (otherwise the counting rate changes would
not have been observed as promptly as they were),
but because of the various effects noted in this
appendix and in the “Reactor Kinetics'' section
of Chapter 4 the thermocouples were slow in
responding. |f the thermocouples had shown instant
response there would have been no hysteresis
effect observed, and the plot of £ vs mean temper-
ature would have been a straight line with a nega-
tive slope proportional to the temperature coef-
ficient.

The plot of & vs mean temperature was obtained
by applying the three corrections noted above in

 

2This plot is actually a cross plot of the curves of
Fig. 4.5, which show both the counting rate and reactor
mean temperature plotted as a function of time.

194

the following way. The first correction was applied
by changing the fission chamber data to BF,
counter data by using the curves of Fig. 0.2. The
resulting points obtained from each fission chamber
were averaged and then plotted on a time scale
along with the reactor mean temperature to produce
a plot similar to the curves of Fig. 4.5. The maxi-
mum of the counting rate curve and the minimum
of the temperature curve were then matched up (the
temperature curve was effectively moved up 2]/2
min in time), and new temperatures were read from
the temperature curve corresponding to the time
that the counts were taken. From the counting
rates the multiplication factor £ = 1 — (1/M) was
determined for each point, and then a plot of & as
a function of mean temperature was drawn up, as
shown in Fig. 0.4. A straight line could reasonably
be drawn through the points.

The third and final correction to be applied to
the mean temperature was obtained from Appendix
K. Since this experiment was one in which the
fuel was cooled rapidly and equilibrium conditions
were not met, the curves of Fig. K.2 are applicable.
From Fig. K.2 it can be shown that a change of
1°F in the true mean temperature corresponds to
a change of 1.40°F in the mean temperature read
from the control room instruments. Thus the mean
temperature change observed had to be increased
by a factor of 1.4. A measurement of the slope of
the curve of Fig. 0.4 gives a (Ak/k)/AT of 1.65 x
10~4. The average & over the plot is 0.922. By
applying the factor 1.4 to the observed mean tem-
perature change, the fuel temperature coefficient
was calculated to be

~1.65 x 104 4
=~1.28 x 10~

— (ARR/AT = 7
o= (Ak/k) 0.922 x 1.4

This value is about 30% higher than that given
by the results of experiment H-5, but it is in fair
agreement in consideration of all the necessary
corrections. A consideration of the errors involved
showed that the maximum error was of the order of
magnitude of 2.4 x 10~3. Therefore,

a = ~(1.28 + 0.24) x 10~4

[f the lower limit of this value is taken, the
agreement between this value and the accepted
value is fairly good. This experiment did not yield
an over-all temperature coefficient.

 

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FISSION CHAMBER COUNTING RATE (counts/sec)

 

4 2 5 10 2 5 10° 2 5 10
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Fig. 0.2. Correlation Between the Counting Rates of the Bl'-'3 Counter and the Two Fission Chambers.

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REACTOR MEAN TEMPERATURE (°F)

Fig. 0.3. Subcritical Measurement of Temperature Coefficient of Reactivity (Uncorrected Fission

Chamber Data).

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Fig. 0.4. Reactivity as a Function of Reactor
Mean Temperature as Determined from the Sub-
critical Temperature Coefficient Measurement. Ex-
periment £-2.

LOW-POWER MEASUREMENTS OF
TEMPERATURE COEFFICIENTS
OF REACTIVITY

The low-power measurements of the temperature
coefficients of reactivity were similar to the sub-
critical measurements, except that with the reactor
critical and on servo at 1-w power, reactivity
introduced by cooling the fuel was observed by a
change in the regulating rod position. The experi-
ment is described in Chapter 5.

As shown in Fig. 0.5, which is a plot of the
regulating rod position vs the observed mean
temperature during the experiment, a hysteresis
phenomenon was obtained. 1t is significant that
no time lags needed to be taken into account in
the interpretation of this data, because thie experi-
ment proceeded slowly enough for equilibrium
conditions to prevail. However, since this was a
cooling experiment, a temperature correction had
to be applied. Cooling took place along the {ower
half of the figure and heating occurred along the
upper portion. Each of the curves had an initial
steep slope corresponding to an initial fuel temper-
ature coefficient and a less steep slope from which
an over-all reactivity coefficient was found.

After correction for the mean temperature read-
ings, the average of the two initial slopes gave
a fuel temperature coefficient of reactivity of
~9.9 x 10~2, and the other slopes gave an average
over-all reactivity coefficient of about —5.8 x 102,
These values agree well with the accepted values
of ~9.8 x 107> and —6.1 x 10> for the fuel and
over-all temperature coefficients of reactivity,
respectively.

197

 

 

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REACTOR MEAN FUEL TEMPERATURE (°F)

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Fig. 0.5. Regulating Rod Position as a’ Function of Observed Reactor Mean Temperature During

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HIGH-POWER MEASUREMENTS OF
TEMPERATURE COEFFICIENTS
OF REACTIVITY

During the high-power operations an experiment
was conducted at a power of 100 kw which was
similar to the low-power experiment just described.
The fuel helium blower was turned on with the
reactor on servo, and the fuel was cooled. How-
ever, the action took place so rapidly that a 1-min
time lag® of the fuel mean temperature had to be
accounted for in plotting the data in addition to
the temperature correction. Figure 0.6, in which
the regulating rod movement is plotted as a function
of the mean fuel temperature, shows the experi-
mental measurements., Two distinct slopes were
observed that corresponded to an initial fuel
temperature coefficient and to an over-all temper-
ature coefficient of reactivity.

From the steep slope (curve No. 1) a fuel temper-
ature coefficient of =1.17 x 10~% (Ak/k)/°F was
obtained, and from the other slope an over-all coef-
ficient of ~5.9 x 103 (Ak/k)/°F was found. These
two values are in fair agreement with the accepted
values of =9.8 x 107> and =6.1 x 10~ for these

coefficients.

 

3Time lags at high power operation were observed to
be shorter than those at low or no power. For a dis-
cussion, see Chapter 4,

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1240 1250 1260 1270 128C 1290 1300 1310 1320
REACTOR MEAN FUEL TEMPERATURE (°F)

Fig. 0.6. Regulating Rod Position as a Fl.'mc-
tion of Reactor Mean Fuel Temperatyre for Experi-
ment H-4,

199

 

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Appendix P
THEORETICAL XENON POISONING

The xenon poisoning which existed in the ARE
was determined experimentally to be an almost
negligible amount. It was therefore of interest to
compute the amount of such poisoning that would
have been present if no xenon had been lost due
to off-gassing of the fuel so that a measure of the
effectiveness of the off-gassing process could be
obtained.

The poisoning by Xe
has reached equilibrium, is given by!

135 when the xenon content

o lyy + ¥9) B¢ 2/
- (Ay + 0py) Ez:

where
P, = the ratio of the number of thermal
neutrons adsorbed in the xenon to
those adsorbed in the fuel,
o, = microscopic xenon cross section for

thermal-neutron absorption,

(v, +y,) = the total fractional yield of Xel35

from fission, both from iodine decay
and direct Xe formation = 0,059,

1\2 = the decay constant for Xel3% -
2.1 x ]0-5/sec,

E/Eu = the ratio of the macroscopic thermal-
neutron cross section fo;sfission to
that for absorption, for U 5. /5 =

f Tu
0.84,

$o = the average thermal-neutron flux in
the fuel (since the entire fuel vol-
ume, 5.33 ft3, is equally exposed to
this flux, although there is only
1.37 #1% of fuel in the core at one
time, the average flux in the fuel is
1.37/5.33 or 0.26 times the average
flux in the reactor, which is 0.7 x

1013),

The Xel35 absorption cross section, Oy is
smaller in the ARE than at room temperature, be-
cause the average neutron energy in the ARE
exceeds the average neutron energy corresponding
to room temperature and because the Xe!33 ab-
sorption cross section drops off rapidly with

 

 

]S. Glasstone and M. C. Edlund, The Elements of
Nuclear Reactor Theory, D. Van Nostrand Co., Inc.,
New York (1952), p 333, 11.57.2.

200

increasing neutron energy. R. R. Bate, R. R.
Coveyou, and R. W, Osborn investigated the neutron
energy distribution in an absorbing infinite moder-
ator by using the Monte Carlo method and the
Oracle. By assuming a constant scattering cross
section and a constant 1/v cbsorpfion cross
section, they found that the neutron energy distri-
bution is well represented by a Maxwellian distri-
bution corresponding to an effective temperature,

T, provided
v 2 < 0,06
K = Y- — 06 ,
3 =,

where the macroscopic absorption cross section
2 is measured at the moderator temperature and
ES is the macroscopic scattering cross section.
The effective temperature

T, = T (1 + aAn) ,

where T  is the moderator temperature, a is a
constant approximately equal to 0.9, and A is the
atomic weight of the moderator.

The present state of the theory does not permit
consideration of the inhomogeneous distribution of
the various constituents of the ARE core, and
therefore the main constituents were considered to
be evenly distributed over the core. The nuclei
per cubic centimeter were thus

Oxygen 5.5 x 1022
Beryllium 5.5 x 1022
u23s 7.8 x 1017

Other elements made only a negligible contribution
to the cross section of the core, The following
cross sections were used;

Oxygen, scattering 4 barns
Beryllium, scattering 7 barns
Uranium, absorption 360 barns?

 

20btained by converting the room temperature value
of the cross section to the value at the reactor operating
temperature by multiplying by the square root of the
ratio of the temperatures:

\ / 293 (°K)
630 (bams) X —— = 360 (barns) .
1033 (°K)

 

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K = - = 0.038 .
3 (4+7) x 555 x 1022
For the atomic weight, A, the average of the

values for beryllium and oxygen, 12.5, was as-
sumed. Thus

T = 10331 + (0.9 x 12.5 x 0.038)]

e

= 1474°K
= 2200°F .

The Xe!33 absorption cross section in the reactor
was then determined, as shown in Table P.1,
which gives the energy intervals of the neutrons,
E., the fraction of neutrons in these energy inter-
vals according to the Maxwell-Boltzman distri-
bution, n(E.)/n, and the total Xel33 cross section,
o, for each energy interval,® from which the
average Xe'3% cross section is obtained. In the
energy range in question the xenon adsorption
cross section is approximately equal to the total
xenon cross section, o,. As shown in Table P.1,

t
the value of this cross section is 1.335 x 10°

barns or 1,335 x 10~ 18 cm?2,

The anticipated xenon poisoning during the 25-hr
xenon run (exp. H-11) may then be computed from
Eq. 1 by using the value determined above for the
Xe'35 absorption cross section:

 

TABLE P.1. Xe!3% ABSORPTION CROSS SECTION
IN THE REACTOR

 

 

 

E (o) /= 2o Ap 0 E) fx 0 (E)
: " (barns) (barns)
0.02 0.060 2.50 x 10%  0.150 x 108
0.04 0.073 2,75 x 10% 0,200 x 108
0.06 0.076 3.25 x 10%  0.248 x 108
0.08 0.075 3,30 x 106 0,248 x 10°
0.10 0.072 2.82 x 105 0,201 x 10°
0.12 0.067 1.92 x 10% 0,129 x 10°
0.14 0,062 1.27 x 10% 0,079 x 108
0.16 0.057 0.75 x 10%  0.042 x 10°
0.18 0.051 0.45 x 10  0.023 x 108
0.20 0.046 0.32 x 10 0.015 x 10°

Average Utxe 1.335 X 108

 

(such as Inconel, etc.), and during the 25 hr of
operation, if the xenon had all stayed in the fuel
it would have reached 69% of its equilibrium
concentration, ? By applying these various cor-
rections, it is found that the xenon poisoning in

the ARE at the end of the 25-hr run should have

 

1.3 x 10=18 x 0.059 x 0.7 x 10'3 x 0.26

 

 

PO == X 0.84
(2.1 x 1075) + (1.3 x 1018 x 0.7 x 10'3 x 0.26)
0.0117 x 10=3 0.0117
4 = = 0.005 .
(2.1 + 0.24) x 10~5  2.34

This value has to be corrected because about
one-~third of the fissions occur at energies above
thermal and therefore have only little competition
from xenon absorption. Furthermore, the reactivity
loss due to poison was about 89% of that computed
above because of the absorption in other poisons

 

been 0.2% in Ak/k if no xenon had been off-

gassed.

 

3BNL-170.
4Glasstone and Edlund, op. cit., p 333,

201

O R all R

 

 

 
 

 

Appendix Q
OPERATIONAL DIFFICULTIES

The operational difficulties described here are
only those which occurred during the nuclear phase
of the operation, that is, from October 30 to No-
vember 12, the period of time covered by this
report. With a system as large, as complex, and
as unique as that which constituted the ARE, it is
amazing that so few difficulties developed during
the crucial stages of the operation. Furthermore,
such troubles were, without exception, not of a
serious nature, This is in large measure attribut-
able to the long period of installation and testing'
which preceded the nuclear operation, the safety
features inherent in the system, and the quality
of workmanship which went into its construction.
All major difficulties and impediments which arose
are discussed below and are grouped by systems,
with special regard for chronology of occurrence.

ENRICHMENT SYSTEM

As mentioned previously, the fuel enrichment
system was changed (shortly before the critical
experiment was to begin) from a remotely operated
two-stage system, in which the transfer was to
start with all the fuel concentrate in a single con-
tainer, to a manually operated two-stage system,
in which small batches of the available concentrate
were transferred, one at a time. A portion of the
equipment used is shown in Fig. Q.1. Although
this change resuited in an improvement both in
safety and control, the temperature control of the
manually operated system was persistently diffi-
cult. Furthermore, in order to avoid plugged lines
because of the concentrate freezing at cold spots,
the lines had to be continuously purged with gas
and the exit gas lines then plugged as a result of
concentrate-vapor condensation. Both these diffi-
culties could have been avoided with proper design.

The temperature control was a greater problem
here than anywhere else in the system because of
the small (5/16 and 3/8 in.) tubing used in the trans-
fer lines and an inferior technique of heater instal-
lation. This problem was aggravated by the virtual
inaccessibility of the connection between the
transfer line and the pump at the time the system
was revised. The final heater arrangement used,
which proved to be satisfactory, consisted of

 

IDesz’gn and Installation of the Aircraft Reactor Ex-
periment, ORNL-1844 (to be issued).

202

double tracing of the line with calrod heaters
staggered so that successive pairs of calrods did
not meet at the same point. Even with this arrange-
ment, thermocouples were necessary at every
heater junction and each calrod or each pair of
calrods should have had a separate control.

After it became an established part of the enrich-
ment procedure to continuously bleed gas through
the transfer line into the pump (in order to keep
the line clear), the gas was vented through an
extra line at the pump. The extra line was not
properly heated and soon plugged with vapor
condensate. |t then became necessary to use the

‘primary pump vent system which had a vapor trap.

By reducing the bleed gas flow to a minimum, this
vent system could be used without becoming
plugged with vapor condensate.

The transfer line to the pump served adequately
throughout the critical experiment, although it had
to be reworked four times either because of the
formation of plugs or the development of leaks (in
Swagelok connections where the line was cut and
replaced). During the last injection for rod cali-
bration the transfer line again became plugged at
the fitting through the pump flange. This fitting
was a resistance-heated concentric-tubing arrange-
ment which provided an entrance for the 1300°F
transfer line through the 700°F pump flange. The
oxidized fuel from previous leaks had shorted the
heating circuit, and the fitting had therefore cooled
and plugged. Attempts to clear the fitting caused
it to leak; the leak was sealed but the fitting was
then inoperable.

The final injection of concentrate, which was
required for burnup at power, xencn poison, etc.,

was therefore made through the fuel sampling line. -

A special batch of concentrate was prepared and
pressurized into the pump through the sample line.
This technique worked satisfactorily but had previ-
ously been avoided because the line had not been
designed to attain the high temperatures > 1200°F
required by the melting point of the concentrate.
Furthermore, the sample line was attached to the
pump below the liquid level in the pump, and if
a leak had developed in the line a sizeable spill
would have resulted.

 

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Fig. Q.1. Enrichment System Transfer Pot and Transfer Lines,

€0t

 

 

- R T T " 2 T NI Y T T WA Ty T T e ST T [T T TR e W T e T b Aot e AN
 

 

- PROCESS INSTRUMENTATION

For an appreciation of the generally excellent
performance of the instrumentation, knowledge of
the number and complexity of the instruments is
required. There were at least 27 strip recorders
(mostly multipoint), 5 circular recorders, 7 indi-
cating controllers, 9 temperature indicators (with
from 48 to 96 points apiece), about 50 spark plugs,
20 ammeters, 40 pressure gages, 16 pressure regu-
fators, 20 pressure transmitters, and numerous flow
recorder indicators and alarms, voltmeters, tachome-
ters, and assorted miscellaneous instruments,
Furthermore, many of these were employed in sys-
tems circulating fuel and sodium at temperatures
up to 1600°F. Since all instruments were subject
to routine inspection and service, petty difficulties
were kept to a minimum.

In the course of nuclear operation, oniy the
following instruments gave cause for particular
concern: fuel flowmeter, main fuel pump level
indicator, several sodium system spark plugs, fuel
pressure transmitter, and several pump tachometer
generators. Of these only the tachometer generator
““failures’’ were not caused by the matericls and
temperatures being instrumented. The tachometers,
as installed, were belt driven rather than direct-
coupled and were not designed to withstand the
side bearing loads to which they were subjected.
The tachometers were replaced, however, before
the pits were sealed, and they performed satis-
factorily during the high-power operation.

The fuel flowmeter and the fuel pump level indi-
cator were similar instruments in which a float or
bob was attached to a long tapered iron core sus-
pended in a *‘dead leg.”” Coils were mounted out-
side the dead leg which located the position of the
core. The position of the core could be interpreted
as a measure either of fuel level or of fuel flow up
past the bob. The coil current, however, was very
sensitive to the fuel temperature, which had to be
maintained above the fuel melting point. In ad-
dition to the temperature sensitivity, several coils
(spare coils were provided on each instrument)
opened up during the experiment, presumably due
to oxidation of the coil-to-lead wire connection.
The fuel flowmeter oscillated rapidly over a 10-
gpm range throughout the later stages of the experi-
ment, although the electronics of the instrument
appeared to be in order. On the other hand, oper-
ation of the main fuel pump level indicator was
satisfactory up to the last day of operation, at
which time the spare coil opened up (the main coil

204

had previously opened). It is felt that these instru-
ments would have performed satisfactorily if the
iron core were designed to move in a trapped gas
leg above the float rather than in a dead leg below
the float so that the operating temperature could
be reduced. Furthermore, the coil reading should
be ““balanced out'’ to eliminate the temperature
sensitivity.

Of the numerous spark plugs which were employed
in the various fuel and sodium tanks as the measure
or check on the liquid level, only three of those in
the sodium system showed a persistent tendency
to short. These shorts could not always be cleared
by *‘short-burning,’”’ but they frequently cleared
themselves as the liquid level dropped. Although
the probes were located in a riser above the tank
top to minimize shorts, the shorts could probably
have been eliminated by using larger clearances
than were afforded by the use of ]/8—in.—OD probes in
a ]/2-in.-IPS pipe riser.

The high-temperature fuel system pressure trans-
mitters suffered a zero shift during the course of
the experiment.  These transmitters employed
bellows through which the liquid pressure was
transmitted to gas. 1t is probable that the bellows
were distorted at times when the gas and liquid
pressure were not balanced as a result of oper-
ational errors or plugged gas supply lines. In any
case, the gas ports in the transmitter occasionally
plugged, and a zero shift in the instrument was
observed.

In view of the large number of thermocouples in
use throughout the experiment (in the neighborhood
of 1000), it is not surprising that a small number
were in error. However, those which gave incorrect
indications included some of the most important
ones associated with the entire experiment. As
discussed in Appendix K there was serious dis-
agreement between line thermocouples inside and
outside the reactor thermal shield, although it
would appear that they both should have given the
same indication. Other misleading temperature
readings were obtained from the thermocouples on
the tubes in the fuel-to-helium heat exchangers.
These thermocouples were not properly shielded
from the helium flow and read low.

Although most of the thermocouple installations
were designed to measure equilibrium temperatures
and did so satisfactorily for a number of experi-
ments involving fast transients, it was important
that the response of the thermocouples be > 10°F
per sec in order to correlate the changes in fluid

 

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temperatures with nuclear changes. Unfortunately
it is not certain that this was the case, and, in
addition, in certain instances, as with one of the
thermocouples on each of the reactor AT and re-
actor mean temperature instruments, the thermo-
couples were mounted on electrical insulators
which increased the thermal lags.

The above discussion covers most of the instru-
mentation difficulties that arose. This is not
meant to imply, for example, that all the 800-odd
thermocouples lasted throughout the experiment;
there were open thermocouples scattered through-
out the system. Furthermore, the instrument me-
chanics were kept busy; when not doing installation
work, they were usually involved in routine service
and maintenance work.

NUCLEAR INSTRUMENTATION AND CONTROLS

It is difficult in the case of the nuclear instru-
ments to separate operation problems from those
inherent in routine installation and debugging,
since these latter operations were continued right
up to the time the instruments were needed. How-
ever, during actual operation, all nuclear instru-
mentation performed satisfactorily. The control
mechanism operated as designed, except that one
shim rod had a higher hold current (it was sup-
ported by an electromagnet) than the other two.
This situation was improved by cleaning the magnet
face and filtering the gas surrounding the magnet.

ANNUNCIATORS

The control system included an annunciator
panel which anticipated potential troubles and
indicated off-design conditions by a fight and an
alarm. During the course of the experiment, cer-
tain annunciators consistently gave false indi-
cations, and therefore the bells (but not the lights)
of these annunciators were finally disconnected.
Included were the standby sodium pump lubrication
system flow, the standby fuel pump cooling water
flow, the fuel heat exchanger water flow, and the
fuel heat exchanger low-temperature alarm. The
first three annunciators had mercury switches
which were either improperly mounted or set too
close to the design condition, but the fuel heat
exchanger low temperature alarm error was due to
faulty thermocouple indication; that is, instead of
indicating fuel temperature, the thermocouple,
which was located in the cooling gas stream, read
low. '

HEATERS AND HEATER CONTROLS

During the time the system was being heated, the
heater system power was over 500 kw. This heat
was transferred to the piping and other components
by the assorted ceramic heaters, calrods, and strip
heaters that covered every square inch of fuel and
sodium piping, as well as all system components
which contained these liquids. Although there
were numerous heater failures resulting from
mechanical abuse up until the time the pits were
sealed, all known failures were repaired before
the pit was sealed. Only four heater circuits were
known to be inoperable at the time the reactor was
scrammed ~ two heaters showed open, two shorted.

Except for the fuel enrichment system in which
the heating situation was aggravated by the higher
temperatures as well as the small lines, the
available heat was adequate everywhere. The
control of the various heater circuits was, however,
initially very poor; it consisted of four voltage
buses to which the various loads could be con-
nected plus variacs for valve and instrument
heaters. However, in order to obtain satisfactory
heater control for all elements of the system, 13
additional regulators were installed in addition to
numerous additional variacs. With the additional
regulators it was possible to split up the heater
load to get the proper temperatures throughout the
system without overloading any distribution panel.
Even with the helium annulus, one function of
which was to distribute the external heat uniformly,
it must be concluded as extremely desirable, if
not an absolute necessity, that all components of
any such complex high-temperature system be
provided with independently controlled heater units
in order to achieve the desired system eguilibrium
temperature.

SYSTEM COMPONENTS

The only major components of either the sodium
or fuel system which caused any concern during
the nuclear operation were the sodium valves
(several of which leaked) and the fuel pump (from
which emanated a noise originally believed to
originate in the pump bearings). In addition, there
were problems associated with plugged gas valves
and overloaded motor relays.

The sodium valves that leaked were the two
pairs that isolated the main and standby pumps
and at least one of the fill valves in the lines to
the three sodium fill tanks. The leakage across

205

T S TR TR T

 

 
 

 

 

- the pump isolation valves was eliminated from
concern by maintaining the pressure in the inoper-
ative pump at the value required to balance with
the system pressures. The leak (or leaks) in the
fill valves was of the order of % to 1 1 of
sodium per day, and was periodically made up by
refilling the system from one of the fill tanks. In
the course of these operations, two tanks eventually
became empty, and it was then apparent that most
of the leakage had been through the valve to the
third tank. This leakage was initially abetted by
the high (™~ 50 psi) pressure drop across the valve,
but even though the pressure difference was re-
duced to 5 to 10 psi the leckage rate appeared to
increase during the run.

It was of interest, as well as fortunate for the
ultimate success of the project, that the fuel carrier
fill valves were tight. However, in the fuel system,
only two valves were opened during the filling
~operation and only one of these had to seal in
order to prevent leakage. This valve did seal, and
it was opened only one other time, i.e., when the
system was dumped.

Each sodium pump {main and standby) and each
fuel pump (main and standby) was provided with
four microphone pickups to detect bearing noises.
Shortly after the fuel system had been filled with
the fuel carrier, the noise level detected on one
of the main fuel pump pickups jumped an order of
magnitude, while that on the other increased
substantially. At the time the noise was believed
to be due to a flat spot on a bearing, and operation
was therefore watched very closely. When the
noise level did not increase further (in fact, it
tended to decrease), it was decided to continue
the experiment without replacing the pump (a very
difficult job which could conceivably have resulted
in contamination of the fuel system). The pump
operated satisfactorily throughout the experiment.
Subsequent review of the pump design and behavior
of the noise level indicated that the noise probably
originated at the pump discharge where a sleeve
was welded inside the system to effect a slip
connection between the pump discharge duct from
the impeller housing and the exit pipe, which was
welded to the pump casing. Vibration of the sleeve
in the slip joint could account for all the noise.

Although the vent header was heated from the
fuel and sodium systems to the vapor trap filled
with NaK (which was provided to remove certain
fission products but also removed sodium vapor),

206

the temperature control of the header and the
individual vent lines connected to it was not
adequate to maintain the line above the sodium
melting point and yet not exceed the maximum
temperature limit (400°F) of the solenoid and
diaphragm gas valves. Consequently, these vent
lines became restricted by the condensation of
sodium vapor. |t was apparent that a higher temper-
ature valve would have been desirable, that the
gas line connecting the tank to the header should
have been instailed so that it could drain back into
the tank, and, also, that good temperature control
of the line and valve should have been provided.
As it was, exceptionally long times were required
to vent the sodium tanks through the normal vent
valves. |t would have been necessary to use the
emergency vent system, which was still operable
at the end of the experiment, if a fast dump had
been required.

In addition to the above failures or shortcomings,
there were numerous probiems of less serious
nature in connection with auxiliary equipment,
motor overloads, water pipes which froze and split,
and air dryers which burned out. However, nothing
occurred in such a manner or at such a time as to
have any significant bearing on the conduct of the
experiment.

LEAKS

The only sodium or fluoride leaks that occurred
have already been discussed. These included one
minor sodium leak in the sodium purification sys-
tem (discussed in chap. 3, *‘Prenuclear Operation’’)
and two fuel concentrate lecks from the enrichment
system. That neither the reactor fuel system nor
sodium system leaked is a tribute to the quality
of the workmanship in both welding and inspection
that went into the fabrication of these systems. In
all, there were over 266 welded joints exposed to
the fivoride mixture, and over 225 welded joints
were exposed to the sodium.

In contrast to the liguid systems, there were
several leaks in gas systems. 1t is felt that these
feaks would not have occurred if the gas systems
had been fabricated according to the standards
used for the liquid systems. The notable gas leaks
were from the fuel pumps into the pits, from the
helium ducts in the heat exchangers into the pits,
and from the pits into the building through the
various pit bulkheads, as well as the chamber
which housed the reactor controls.

  

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The combination of the leak out of the fuel pump
and that out of the pits required that the pits be
maintained at subatmospheric pressures in order
to prevent gaseous activity from contaminating the
building. Accordingly, the pit pressure was lowered
by about 6 in. H20 by using portable compressors
which discharged the gaseous activity some 1000 ft
south of the ARE building. The activity was of
such a low magnitude that, coupled with favorable
meteorological conditions, it was possible to

operate in this manner for the last four days of the
experiment.

The ledks out of the helium ducts in the fuel and
sodium heat exchangers resulted in a maximum
helium concentration in the ducts of the order of
50%, and to maintain even this low concentration,
it was necessary to use excessive helium supply
rates, i.e., 15 cfm to the ducts alone and another
10 cfm to the instruments.

207

R

 
 

 

 

Appendix R
INTEGRATED POWER

The total integrated power obtained from the
operation of the Aircraft Reactor Experiment does
not have any particular significance in terms of
the operational life of the system. However, a
total integrated power of 100 Mw was more or less
arbitrarily specified as one of the nominal ob-
jectives of the experiment. While at the time the
experiment was concluded it was estimated that
this, as well as all other objectives of the experi-
ment, had been met, the estimate was based on a
crude evaluation of reactor power. Since subsequent
analyses of the data have permitted a reasonably
accurate determination of the reactor power, it is
of interest to reappraise the estimated value of the
total integrated power,

The total integrated power could be determined
from either the nuclear power or the extracted
power (cf., section on ‘‘Reactor Kinetics'' in
chap. 6). The power curves, which should have
equivalent integrals, could be obtained from any
of a number of continuously recording instruments,
i.e., the nuclear power from either the micromicro-
ammeter or log N meter, and the extracted power
from any of the several temperature differential
recorders in the fuel and sodium circuits., The
total integrated power has been determined both
from integration of a nuclear power curve (log N)
and from the sum of the extracted power in both
the sodium and fuel circuits, as determined by the
temperature differentials in each system together
with their respective flows (which were held
constant). For both power determinations a calcu-
lation of the associated error was made.

EXTRACTED POWER

The integrated extracted power was determined
from the charts which continuously recorded the
temperature differential (AT) in each system. As
a matter of convenience the sodium system AT
was taken from a 24-hr circular chart, while the
fuel AT was taken from one of the six Brown strip
charts which recorded, in the control room, the
AT across each of the six parallel fuel circuits
through the reactor.  The individual circuit
recorders had a much slower chart speed than
that of the over-all AT recorder and were therefore
much easier to read, To keep the results well
within the accuracy of the whole experiment, tube
No. 4 was selected for the determination because

208

the chart trace very closely corresponded to that
of the over-all fuel AT across the reactor. From
Fig. K.3, Appendix K, the control-room-recorded
AT was converted to what was accepted as the
correct AT, The power extraction was then calcu-
lated and plotted against time in Fig. R.1.

The sodium AT across the reactor was obtained
from the circular charts of the recorders located
in the control room. These AT's were also
corrected by using Fig. K.6, Appendix K, and
the plot of the power extracted by the sodium as a
function of time is also shown in Fig. R.1 on the
same abscissa as that of the fuel plot. The total
integrated (extracted) power, i.e., the sum of the
area under both the fuel and sodium system power
curves, was then determined by using a planimeter;
it was found to be 97 Mw-hr,

A calculation was also made of the magnitude
of the ““maximum’’ possible error in the determi-
nation of the reactor power. The power equation,
which was calculated for both the fuel and sodium
systems, was

P = kf AT ,
where
P = reactor power,
k = a constant containing heat capacity and
conversion units,
f = flow rate of fuel or sodium,
AT = temperature difference across reactor,

The consequent error equation is

AP = Ef A(AT) + EAT Af + AT Ak ,

where
Ak = maximum error in the heat capacity,
Af = maximum error in the flow rate,
A(AT) = maximum error in the temperature

difference.
Nominal average values of these factors for the
fuel system were

f = 46 gpm
AT = 350°F
k= 0.11 kw/day gpm
Af = 2 gpm = 5%
AMAT) = 10°F = 3%
Ak = 0.011 £ 10%

Therefore

AP (for fuel system) = 20%

   

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EXTRACTED POWER {Mw) EXTRACTED POWER (Mw)

EXTRACTED POWER (Mw)

o
1300

0
G100

2.8

2.0

0.8

0.4

1300

1500

{700 4900
NOVEMBER 8

PRACTICE
OPERATION

0300 0800 Q700

1500

{700 {1900
NOVEMBER i

2100

OPERATION

PRACTICE
EXP H-7

0900

IEMONSTRATION

D
Iro AlR FORCE

2100

FUEL SYSTEM POWER

——=—— SODIUM SYSTEM

2300 0100 0300 050C O70C 0900 400 1300
NOVEMBER 9

OPERATION
TO AIR FORCE

PRACTICE
EXP. H-10

RUN {

1100 1300 1800 {700 1900 2100 2300 0100
NOVEMBER 40

DEMONSTRATION
QOF HIGH-POWER
QOPERATION
EXP, H-i4

2300 0100 0300 0500 O70C 0900 00 1300
‘ NOVEMBER 12

Fig. R.1. Power-Time Curve.

ORNL-LR-DWG 63555

 

1500

0300

1500

1700

S

1800

0500 0700
NOVEMBER 1

{700

1200

2100

0900

2100

 

2300 0100

1400

 

2300

 

1300

209

 

 

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The values for the sodium system were

f = 152 gpm
AT = 50°F
k = 0.0343 kw/day.gpm
Af = 4 gpm = 3%
A(AT) = 10°F = 20%
AR =0

Therefore
AP (for sodium system) = 11%

The weighted, over-all percentage error was
therefore

AP P,»XOQO PSXOO]]

—_— = + =17,
p P pP

 

where P, and P_ are the power extracted in the
fuel and sodium systems, The errors made in the
power integration by using the AT chart are only
those associated with the determination of power
extraction. Since Fig. R.1 is a reproduction of the
AT trace and it was integrated by using the
planimeter, any integrating errors were assumed
to have been averaged out. Therefore, the error
that applies to the integrated exfracted power is
the 17% that is applicable to the power level
determination; therefore the integrated extracted
power was

97 £ 16.5 Mwshr .,

NUCLEAR POWER

During most of the time the reactor was operating
at any appreciable power level, the nuclear power
level was kept fairly constant and was recorded in
the log book. During the times the power level
was either not constant or not recorded in the log
book, the nuclear power was determined from the
log N recorder chart by integrating the area under
the power trace, (The log N trace was used
rather than the micromicroammeter because no
record was kept of the micromicroammeter shunt
value as a multiplying factor.) Since the log N
chart gave a log of power vs time plot, exact
integration under the curve would have been an
extremely difficult task; accordingly, the area
under the curve was integrated graphically. The
curved portions of the trace were approximated by
straight lines, and an average log N valuve was
determined for each line segment. |t was assumed
that there were as many positive as negative
errors in this method and that the errors cancelled
out, The total integrated power was then obtained

210

by adding up all the incremental areas which were
in terms of average log N units times time,

From the 25-hr constant-power xenon run a
relation between the log N reading and the
extracted power was obtained, |t was found that
22.5 log N units = 2.12 Mw, or 10.6 log N units = ]
Mw; therefore

log N
10.6

for any segment under the curve. By using this
correlation between actual power and log N
reading, the total integrated nuclear power was
calculated to be 96.6 Mw-hr. This value of the
integrated nuclear power, O, as determined from
the log N chart, was, in effect, found from the
following relation

x time (hr) = Mw-hr

 

0 = MCA: ,

where
M = average log N recorded value over an
increment of time At,

At = increment of time in hours,
C = a constant which converts the log N value
to Mw-hr,

The constant, C, in the expression is equal to the
percentage error in the extracted power level
determination, which was calculated in the
preceding section to be 17%. Therefore the error
in C is (0.17) (1/10.6) = 0.016. The 74-hr period
of high-power operation was divided into two
parts, one part being the 25-hr period of constant
power during the xenon run and the remainder
being the 49-hr period of variable power. The
errors were of different magnitude for each part,
since during the xenon run there was no error in
time or in the determination of the average value
of M,

For the 25-hr xenon run the integrated power was

53 Mw-hr. Thus

M = 22.5 log N units

At = 25 hr
C = 0.094 Mw/log N
AM = 0
A(A?) = 0
AC = 0.016 = 17%

Therefore
AQ] = MC AAt) + M AC At + AM C At = 9 Mw-hr .

For the balance of the operating time (49 hr) and
integrated power (43.6 Mw-hr), the average M must

 

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be determined as

H=-Q—= 43.6

— . = 9.4
Ct  (0.094) (49) ¢

and for AQ,

M = 9.46 log N units
49 hr

= 0.094 Mw/log N
AM =1

a2
I I

AAL) = 1 hr
AC = 0.016 = 17%

AQ,=MC A(At) + M AC At + AM C At = 12.8 Mw-hr,

The total error was therefore

AQ = 12.8 + 9.0 = 21.8 Mw-hr or 22.5% .

The total extracted nuclear power then was

96.6 + 21.8 Mw-<hr .

211

 

:
g

e o - T

 
When excess reactivity is being introduced into
a reactor at a given rate, the period meter will
show @ pertod which is not a true period but a
combination of the true period and its time rate of
change. If the time rate of change of the period is
known, then the true period may be found by the
following means.

First, it is assumed that whenever any excess
reactivity is introduced into the reactor the power
will rise according to the following equation:

P o= PO(.’AI ,
Where
PO — initial power,
A = ]/./""I

T = reactor period,

 

t = time power is increasing.

Then, to introduce a time rate of change of the
period, the power is expressed in the form of an
infinite series

) A2f2 ;
1=I01+)U+2! T

If the higher order terms are neglecfed, the rate

 

of change of power is

dP A
= Py A+t ,
dt dt

and therefore

 

 

212

 

Appendix $
INTERPRETATION OF OBSERVED REACTOR PERIODS DURING TRANSIENTS

and

-«

dp dr
—_—
dt dt

In experiment H-8 {cf., chap. 6) it was noted

that when the regulating rod was withdrawn the :

period observed was not a true period and that it
kept changing over the 35 sec it took the regulating
rod to travel its full distance. For a calculation
of the true period, the following values were

taken from Fig. 6.10 and Table 6.4:

Py = 2.34 Mw,
P = 3.94 Mw,
7, = 25 sec (initial period observed),
7, = 50 sec (final period observed),
t = 35 sec.
From these values, it was found that
dP
— = 0.0457 Mw/sec
dt
and :
1 1
A 50 25
t —— = 35 = -0.02 .
t 35 ~
Therefore
~ 2.34
T = = 36 sec .

0.0457 + 0.02

This calculated valve for the true period corre-
sponds fairly closely to the period of 42 sec
measured from the slope of the log N recorder
trace,

 

 

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Appendix T
NUCLEAR LOG

The following information was copied verbatim from the nuclear log book, The only material omitted
was the running uranium inventory, which was kept during the critical experiment, and several calcu-
lations and graphs, which were inserted in the log book in an attempt to interpret the data, All the
omitted information is presented in much better form elsewhere in this report.

The reactor power data referred to in the log are not consistent, because the power was estimated
first from a calculation of flux at the chambers, subsequently from the activation of a fuel semple, and
finally fromthe extracted power, If it is assumed that the extracted power value is correct, the original
estimate was a factor of 2,7 low and that calculated from the fuel activation was a factor of 2 low. In
the log book, all power levels mentioned through experiment L -8 were based on the original estimate;
from experiment L-9 to H-14, all power levels were based on the fuel activation; and it was not until

after high-power runs 1 and 2 of experiment H-14 that correct power levels were listed,

213

 

 

< v

TmT

CEETEE T
 

 

 

Source in core fission chamber 2 42.0 counts/sec :

BF, 6.97 counts/sec E.

1845 fission chamber 1 1.49 counts/sec
Source out of core fission chamber 2 2.24 counts/sec 1

BF3 13.44 counts/sec :‘:..

¥

1855 fission chamber 1 27.01 counts/sec f
Source in core fission chamber 2 42.59 counts/sec i

BF, 6.82 counts/sec “

3 1912 Started adding lot from can 7. E
1920 Started transferring first 5-1b, E
1930 Plugged line from transfer tank to pump. E
2015 Chemists report that uranium concentration of sample 6 taken at 1720 is *’
2.47 + 0.01 wt %, E

214

{ fission chamber 1

27.1 counts/sec

| EXPERIMENT E-1 '
: October 30, 1954 .
Objective: Bring the Reactor Criticol b
1 Run Time %
1 1425  Zero fuel flow. ¥
1 1500 Main fuel pump started to bleed down to minimum prime level. , ;’
1 1507 Main fuel pump at minimum prime level. System volume is now Cd
| 4.64 + 0.18 = 4.82 ft* (calculated). b
3 2 Started adding lot from can 6. ::
i 1539 First 51b lot going into pump. . ;
: 1545 It was noticed that pips on the fission chamber occurred in about 2-min 3
: i‘n'rervcls (pump rpm, 520). E

1554 Second 5-lb batch going in. F

1558 Third 5-1b batch going in. Interval is exactly 2 min. This gives a flow of ;

18.8 gpm for a pump speed of 520 rpm, b

1603 Fourth 5-1b batch going in. %

1610 Fifth 5-1b batch going in. .

1614 Last of first 30-Ib lot going in. ;

1623 Speeded up pump. Roughly 26 counts/sec on fission chamber 1 and 40 :

counts/sec on fission chamber 2. 500 rpm on pump. E

1625 Pump speed 1100 rpm, :

1720 Sample 6 taken from pump for analysis. k

~ 1730 Shim rods up, fission chamber 1 26.74 counts/sec f‘y

control rod down, fission chamber 2 41.86 counts/sec ;

fission chambers down. BF, 6.93 counts/sec ,,— E

These count rates checked very well with count rate meters. §

~ 1750 Shim rods down, fission chamber 1 18.74 counts /sec :*

control rod down, fission chamber 2 24.18 counts/sec . "

fission chambers down. BF, 6.03 counts/sec

1812 Rods started up. Control rod back to 8-in. out. ?

~ 1827 Rods out and count started prior to taking source out of core. §

 
Farliidr i ik,

it Bl

e

R

o ey .

e

Y-V TN

Run
)

Time

0035
0230
2000
2006

2230

2300
2307
2315

0008
1315

1320

1335

1349
1403
1413
1415
1421
1505
1615
1617

1707
1720
1731
1742

1753

1804
1838

1900
1940

2028

EXPER‘MENT E-] (continued) Octob 31 1954
croper ’

Attempt to clear line from transfer tank to pump.
Still unsuccessful.

New transfer line into main fuel pump installed.
Sample 7 drawn off for analysis,

Analysis of sample 7 taken ot 2006 shows 1.84 *+ 0.04 wt %
total uranium,
Grimes estimates that no appreciable concentrate was transferred
last night (i.e., Saturday night) at the time the line plugged, Therefore,
this sample should be representative of run 2 and supersedes the previous
sample which showed 2.47% (see above).

Starting transfer of first 51b of 25 Ib in can 7.
0.78 min between pips = 43 gpm, indicator on rotameter shows 46 gpm,
Transfer line plugged again.

November 1, 1954

Rods inserted.

From data on rod position vs count rate, rods appear to be ineffective
above 30 in.

Plug removed from transfer line., Can 7 is still attached to the system.
It is estimated that 5 of the 25 |b in can 7 were injected last night. The

~ other 20 Ib are now to be injected in 5-1b batches.

Started transfer of second 5 Ib of 25 |b in can 7,
Interval between pips on fission chamber 2 was 30 sec. Pips are
becoming hard to see.
Started transfer of third 5 Ib. Interval between pips ~0.86 min.
Started transfer of fourth 5 Ib,
Started transfer of last batch from can 7.
Transfer of 25 |b from can 7 completed. Interval between pips ~0.79 min.
Withdrew rods from 30 to 35 in. Three 10-min counts taken.
Rods inserted.
Set log N to come on scale at an estimated 2 x 103 w,
Sample 8 taken for analysis.

First 5 1b of 30 Ib from can 8 transferred.
Second 5 |b transferred, '

Third 5 Ib transferred..

Fourth 5 Ib transferred.

Fifth 5 Ib transferred.

~Sixth (final) 5 Ib from can 8 transferred.

Analysis of sample 8 taken at 1617 shows 3.45 + 0.01 wt %
total uranium, _ | ‘

Rods inserted, Check on noise on pile period recorder started,

Pile period meter shows microphonics, Epler put log N back on normal
setting. Log N should come on scale ~1/2 w.

Taking sample of fuel system, sample 9,

215

e

A T T s

 

 
 

EXPERIMENT E-1 (continued) November 1. 1954
V! ’

Run Time

5 2112 First 5 Ib from can 9 transferred.
2123 Second 5 Ib transferred.
2133 Third 5 Ib transferred,
2142 Fourth 5 Ib transferred.
2153 Fifth 5 Ib transferred.
2203 Sixth (final) 5 |b transferred.
2255 Chemists report 5.43 wt % total uranium for sample 9 taken at 2028. .

November 2, 1954

s e b o i e

6 0104 First 5 Ib from can 10 transferred.
0117 Second 5 Ib transferred,
0128 Third 5 Ib transferred. Two 5-min counts taken,
0153 Fourth 5 b transferred,

0202 Fifth 5 Ib transferred.

4 0213 Sixth (final) 5 Ib transferred.

7 0506 First 5 Ib from can 12 transferred,
0521 Second 5 Ib transferred.
0533 Third 5 [b transferred. Two 5-min counts taken.
0551 Fourth 5 Ib transferred.
0600 Fifth 5 |b transferred,
0610 Sixth (final) 5 1b transferred.

8 0831 First 5 Ib from can 5 injected. Leak in line occurred.

 

EXPERIMENT E-2
Objective: Preliminary Measurement of Temperature Coefficient

1003 Heat barriers started up, >
1005 Heat barriers up.
1005 Blower started.
1005:10 Blower up to 275 rpm.
1010 Blower off.

Water heat exchanger rose 25°F,

Manometer reading 5.45 in.; corresponds to ~ 170 gpm,

YR TRt M M TR o e » PR WY amp WA Y MO omey e vy MR R gt AR VR wewr rSeesr oy TSI o Wi

1615 Sample from fuel system taken, This is sample 10,
1820 Results from sample 10 show 9.58 + 0.08% uranium,
2015 Sample 11 taken for analysis.

2020 Referring back to run 8, experiment E-1:
it is estimated that 5.5 Ib from can 5 went into transfer tank. Approxi-
“mately 0.2 Ib was lost in the leak and 5.3 |b went into the system,
Counts vs shim rod position were taken at 5-in, intervals on shim rods,

, Time of day was 1442 to 1512,
2215 Sample 11 showed 9.54 + 0.08% total uranium, =

November 3, 1954
0136 Second 5 b batch from can 5 injected. Rods at 20 in, Counts taken at

intervals for each 5 in. of rod withdrawal,
ol 0152 Rods inserted to 20 in.

! 0205  Injection nozzle shorted.

T R MR TR A A ER e

 

“*_.
; .
“
&
i.
i
g
£
fr

216

 
BXReE .

W

Run Time

(8) 0226

0256

9 0459

0523

10 0836

0911
0941

1018
11 1110

1147

1230
1245

12 1453

1536

1545
1547
1603

1604
1626

0830
0840

EXPERIMENT E-2 (continued)
November 3, 1954

Third 5 [b batch from can 5 injected, Rods at 20 in. Counts taken as
rods withdrawn,

Final batch from can 5 injected, Chemists estimate about 21/2 Ib, Counts
taken as rods withdrawn.

First batch consisting of 5.5 b from can 11 injected. Rods at 20 in.
Counts taken as rods withdrawn,

Remainder of can 11 injected. Rods at 20 in. Counts taken as rods
withdrawn,

5.5 Ib from can 22 injected. Rods at 20 in. Counts taken as rods
withdrawn,

Second 5.5 Ib from can 22 injected. Counts taken as rods withdrawn,

Completing injection from can 22, Can 22 empty. Counts taken as rods
withdrawn,

Reactor monitored. Reads ~§ mr/hr.

5.5 Ib from can 31 injected. Rods at 20 in. Counts taken as rods
withdrawn.

Second 5.5 Ib from can 31 injected. Rods at 20 in.

Fission chamber 2 withdrawn 2 orders of magnitude.

Balance of can 31 injected. Rods at 20 in, Counts taken as rods
withdrawn,

5.5 Ib from can 20 injected, Rods at 20 in. Counts taken as rods
withdrawn,

Source withdrawn — reactor subcritical,

Second 5.5 Ib from can 20 injected. Rods at 20 in. Counts taken as rods
withdrawn,

Reactor critical.

Control given to servo.

Radiation survey made, Reactor reads 750 mr/hr at side; ~ 10 mr/hr on
grill above main fuel pump.

Reactor shut down,

Shim rods brought out to about 18 in, with source in core,

November 4, 1954

Sample 12 taken.
Sample 13 taken.

EXPERIMENT L-1

Objective: One-Hour Run at 1 w! (Estimated) to Determine Power Level;

1107
1118:40

1218:40
1250
1300
1400

.Radiatien Level Check

Shim rods coming out,

Reactor up and leveled out. Estimated 1 watt.! On servo,

Micromicroammeter 1 x 10=9; 48.6 on Brown recorder,

Reactor scrammed.

Sample 14 taken._'

Sample 15 taken,

Estimated power from sample 15, 1.6 w.
repeated at 10 w.

2 Count was low. Must be

 

YActual power subsequently determined to be 2.7 w,

250 Appendix H, **Power Determination from Fuel Activation."’

217

T T TR

 

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Run

218

Time

1525

1651
1657
1707

1710
1746

1758
1804
1814

2009
2018
2024
2030

2230

0142
0158
0210
0235

0240
0245
0250
0252

0346
0350
0354
0357
0401

- i d
EXPERIMENT L-1 (continued) November 4, 1954

Four samples taken from the fuel system since going critical.
Analyzed as follows:

Time Sample No. Uranium (total) (wt %)
0830 12 12.11 £ 0.10
0840 13 1221 £ 0.12
1250 14 12,27 + 0.08
1300 15 12.24 + 0.12

EXPERIMENT L.2
Objective: Rod Calibration vs Fuel Addition

Reactor brought critical before injecting first penguin,
Rods coming out.

Reactor up to ~1 w. Regulating rod at 13.1 in,
Reactor subcritical,

Penguin 11 injected,

Reactor up to ~1 w. Regulating rod at 10.6 in. Rod moved 2.5 in,,
from 13.1 to 10.6 in.

Readjusted rods. Regulating rod at 13.0 in.

Reactor subcritical.

Penguin 19 put in furnace,

Penguin 19 injected. Time between pips, 0.75 min,

Reactor up to ~1 w. Red at 5.7 in.

Reactor scrammed. '

Started changing from rod 4 (19.2 g/cm) to rod 5 (36 g/cm). Counts

on fission chambers and BF , taken before changing rod.

EXPERIMENT L-3

Fuel system characteristics were obtained,

EXPERIMENT L.2 (continued)
November 5, 1954

Started withdrawing rods,

Upto~1lw,

Reactor subcritical.

Up to ~1 w again. Temperature had drifted. Regulating rod
position, 12,1 in,

Reactor subcritical,

Penguin 4 injected.

Reactor at ~1 w again. Regulating rod position, 12.1 in,

Reactor subcritical, Apparently penguin didn't come over
(no dip line).

Reactor at ~1 w, Regulating rod position, 11.5 in.

Reactor subcritical,

Penguin 14 injected,

Reactor at ~1 w, Regulating rod position, 9.9 in.

Reactor subcritical. Regulating rod movement, 1.6 in.

~j

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Run Time

4 0429
0431
0439
0442
0445

5 0519
0524
0529
0536
0542

6 0607
0612
0619
0626
0631

0715

1015
1040
1105

1350
1403
1413
1414

1416

1417

1418

EXPERIMENT L-2 (continued) N ber 5. 1954
ovember 5,

Reactor at ~1 w. Regulating rod position, 9.7 in.

Reactor subcritical,

Penguin 16 injected.

Reactor at ~1 w. Regulating rod position, 7.8 in. Rod moved 1.9 in.
Reactor subcritical,

Reactor at ~1 w. Regulating rod position, 7.65 in.
Reactor subcritical.

Penguin 13 injected.

Reactor critical at ~1 w. Regulating rod position, 6.80 in.
Reactor suberitical.

Reactor critical at ~1 w. Regulating rod position, 6.5 in.
Reactor subcritical,

Penguin 5 injected.

Reactor critical. Regulating rod position, 5.50 in.
Reactor subcritical,

Shim rods 1 and 2 inserted. Reactor shut down due to increase in
radiation level above fuel pump tank upon adding concentrate to pump.
During fuel addition for run 9 the background picked up to ~ 50 mr/hr
and, on run 5 addition, increased to 55 mr/hr. The normal background at
the point of measurement being ~1 mr/hr.

Trimming pump level to normal operating probe level.
Trimming pump level to estimated ‘4 in. below normal operating probe,
Took fuel sample 16.

EXPERIMENT L-2-A

Objective: Test on Activity of Yent Lines by Operating Reactor at

~1 w for ™~ 10 min and then Venting as Though Adding Fuel

Preliminary data recorded.

Rods 1 and 2 being withdrawn.

Rods at 30 in., rod 2 being withdrawn., Fission chamber 1 at full scale,
Period meter shows slight period, ~ 400 sec.

Rod 2 at upper limit,
Rod 1 being withdrawn,

Period meter reaches 100 to 50 sec.
L.og N reading, 2 x 10~4,

Fission chamber 2 at 500 counts/sec,
Rod 1 at upper limit

Regulating rod being withdrawn.

Micromicroammeter reading, 10.
Fission chamber 2 at 1000 counts/sec.
Log N, 3 x 10-4.

Period, ~400 sec,

219

Term

 

 

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EXPERIMENT L<2<A (continued)
November 5, 1954

Run Time

] (6) 1420 Log N, 5 x 10-4
: Fission chamber 2, 2 x 103,
Micromicroammeter, 20.

Period, ~400 to 100 sec.

1421 On servo, Micromicroammeter, 38.
Log N, 103. Fission chamber 2, 3.5 x 103 counts/sec.

1432 Regulating rod inserted. Shim rods 2 and 1 inserted,

1515 Log N and period channel normal and operating correctly
except 5-sec period reverse is disconnected for duration of

 

calibration run.

1940 The activity observed at 0715 was explained during the run 1420

to 1432 as activity in the vent line from the pump.

»
Tt e e g e THY T pR W

1945 Analysis of sample 16 taken at 1105 after the 70-1b removal gave: %
Total Uranium 12.54 wt % uvranium or 11,7% U235 é
Chromium 372 ppm ' F
fron 5 ppm %
Nickel <5 ppm Y
EXPERIMENT L-2 (continued) E
7 1947  Started pulling shims. g
2002 Upto~1w. ok
2009 Reactor subcritical, ;j
2014 Move shim rods — must reset, Started pulling shims.
2016 Up to ~1 w. Regulating rod position, 6.9 in. ¢
2020 Shim rods going in. -
2025 Penguin 15 injected. :
2032 Up to ~1 w, Regulating rod position, 6.2 in.
2050 Shim rods going in. Regulating rod movement, 0.7 in.
8 2148 Up to power of ~1 w, Regulating rod position, 6.8 in.
2156 Reactor subcritical. Shim rods going in.
2204 Penguin 18 injected,
2211 Up to ~1 w, Regulating rod position, 5.7 in.
2220 Shim rods going in. Reactor suberitical, Regulating rod
movement, 0.6 in.
9 2238 Started up to ~1 w.
2242 Reactor at ~1 w, Regulating rod position, 6.9 in.

2250 Shim rods going in. Reactor subcritical.

2258 Penguin 17 injected.

2305 Reactor at ~1 w, Regulating rod position, 3.3 in,
2312 Shim rods going in. Reactor subcritical,

10 2332 Reactor at power of ~1 w. Regulating rod position, 13.5 in.
2340 Shim rods going in.
2345 Penguin 12 injected,
2352 Reactor critical at ~1 w. Regulating rod position, 12,95 to 13.0.
2400 Shim rods going in, Reactor subcritical.

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11 0020
0026
0100
0212

0223
0228

0305
0315
0318
0320:11

0420:11

0505
0516
0524

EXPERIMENT L-2 (continued) N ber 6, 1954
ovember o,

Reactor critical, Adjusted regulating rod to ~3.5 in. with shims 1 and 2,

Shim rods inserted, Reactor subcritical.

Developed gas leak in injection system during injection of penguin 10,

Leak repaired; apparently penguin 10 was injected at 0100. Rods coming
out,

Reactor critical. Regulating rod position, ~2.44 in.

Reactor scrammed.

EXPERIMENT L-4

Objective: One<Hour Run at 10 w to Make Radiation Survey

and Observe Pile Period

Shim rods coming out.

Leveled off manually at ~0.1 w. Regulating rod position, 7 in.

Pulled regulating rod to 8 in.

10 w estimated power.3 Rod moved from 6.8 to 7.9 on chart paper.
40-sec period according to the chart, Flux servo demand, 500.
Micromicroammeter 51 on range 1 x 10-8,

Reactor scrammed after ~1 hr at 10 w estimated power. Pile period
from slope of log N, 51 sec.

Sample 16 taken.
Sample 17 taken,
Sample 18 taken,

EXPERIMENT L-5

Objective: Calibration of Regulating Rod from Reactor Periods

1 0651

2 0705

0925
0934
0939
0940
0942
0945
0947
0948

Leveled out at ~1 w,
Shim 3 at 30 in. Regulating rod at 13 in.
Shim 3 ot 33.3 in. Regulating rod at 3 in,
Shim 3 at 30 in. Regulating rod at 6.5 in.
Shim 3 at 28.1 in. Regulating rod at 13.4 in.

Pulled regulating rod from 6.53 to 8.61 in. Period, 22.1 sec,

Shim rods coming out.

Reactor at 1 w,

Reactor suberitical.

Pump stopped.

Reactor at 1 w.

Pulled regulating red,

Reactor scrammed at ~50 w.

Pump started, From log N, 21.3-sec period. From recorder:
" regulating rod moved from 6.61 to 12.66.

 

3 Actual power subsequently determined to be 27 w.

221

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1 12 0136
% 0138
4 0231
; 0237
| 0245
Q 0249
0423
1

 

4 1605
1611
1613

5 1804
1808
1810

6 2020
2022
2024
2026
2034
2036
2037

7 2201
2208
2210

2 2307
2353

EXPERIMENT L«2 (continued)
November 7, 1954

Rod Position (in.)

 

 

Shim Rods
Regulating Rod ] 2 3
Reactor up to ~1 w, 8 32.8 324 321

Reactor suberitical,
Start injection from cans 120 and 125.
Injection completed,
Reactor up to ~1 w, 8 28.2 28.2 27.8
Reactor shutdown. :
Sample 19 taken for chemical analysis. All fuel injection and sampling
lines removed from pump. It is estimated that the liquid level is about
0.1 ft3 below the normal operating level probe; therefore final systemvolume
is 5.35 ft3,

Results on sample 19:

Uranium (total) 13.59 = 0.08 wt %
Chromium 445 ppm

4

EXPERIMENT L+5 (continued)

Up to 1 w.

Pulled regulating rod.

Scrammed at 100 w. Wrong paper on log N. By super-imposing proper
paper got 27.2 sec.

Regulating rod withdrawal! from 2.80 to 4.80 in., from Brown,

Up to 1 w,
Pulled regulating rod.
Scrammed at 100 w. From log N, period is 22.4 sec; Ak/k, 0.06%.

Up to 1 w. Pump at 48 gpm,

Reactor subcritical,

Pump stopped, zero flow.

Up to T w.

Pulled regulating rod.

Reactor scrammed at 60 w.

Started pump. Period, 23.0 sec from slope on log N. Rod moved from
2.60 to 8.70 in., or 6.1 in,

Reactor up to | w,
Pulled regulating rod.
Scrammed at 100 w. Period, 20.6 sec from slope on log N. Rod moved from

4,19 to 6.27 in., or 2.08 in, (Brown)

EXPERIMENT L-6

Objective: Colibration of Shim Rods vs Regulating Rod

One set of data taken before final fuel addition,

Up to 1 w. Data recorded.
End of run.

 

222

 

4lrwemory gave 5.33 #t3 for the volume of the fluoride in the system.

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11

12
13
14

15

Time

2353
2358
2400

0147
0050
0057
0107
0110
0113
0115
o1z
0119

0122
0125
0127

0131
0136
0139

0141
0142

0142
0157

0214
0215
0218
0221
0222

0230
0231

0233

EXPERIMENT L-5 (continued) November 7, 1954
ov ’

Start of run, reactor already at 1 w,

Pulled regulating rod.

Scrammed at 200 w. Period, 23.0 sec from slope on log N..
Regulating rod moved from 8,32 to 10.38 in., or 2.06 in.

EXPERIMENT L-7
November 8, 1954

Objective: Effect of Fuel Flow Rate on Delayed Neutrons

Up to 1-w power,

Fuel Flow rate, 48 gpm. Regulating rod position, 12.2 in.
Fuel Flow rate, 41 gpm. Regulating rod position, 11.5 in.
Fuel Flow rate, 34 gpm. Regulating rod position, 11.0 in,
Fuel Flow rate, 30.5 gpm. Regulating rod position, 10.5 in.
Fuel Flow rate, 25 gpm. Regulating rod position, 10.0 in,
Fuel Flow rate, 20 gpm. Regulating rod position, 9.35 in.
Fuel Flow rate, 16 gpm. Regulating rod position, 9.0 in.

Fuel Flow rate, 0 gpm. Regulating rod position, <2 (off-scale).
Reset shims.

Fuel Flow rate, 11.7 gpm. Regulating rod position, 12.3 in.
Fuel Flow rate, 0 gpm. Regulating rod position, 3.6 in.

Fuel Flow rate, 12 gpm. Regulating rod position, 12.2 in.
Reset shims.,

Fuel Flow rate, 12 gpm. Regulating rod position, 3.7 in.
Fuel Flow rate, 43.5 gpm. Regulating rod position, 6.45 in.

Fuel Flow rate, 37.5 gpm. Regulating rod position, 8.0 in.
- Stopped sodium flow.

Fuel Flow rate, 37.5 gpm. Regulating rod position, 7.9 in.
End of experiment,

EXPERIMENT L-8

Objective: Measure Reactor Temperature Coefficient

Reactor already up to 1 w,
Took data,

Start prime mover.

Started raising barrier doors,

Helium blower, 255 rpm.

Readjust shims to put regulating red to top of scale.

Heat exchanger water flow, 195 gpm. Temperature rise, 62 to 76°F.
Steady state, :

Low heat exchanger reverse. Prime mover off,

Barrier doors down.,

Took data,

223

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Run
(3)

224

Time

0242

0700

0950

0950

. 0954

1004
1010
1018

1028
1033
1101

1106
1113

1116
1128

1132
1150

1155
1159

1445
1448

1452
1505

1525

1611
1619

1125
1127

1138

1208
1224

EXPERIMENT L-8 (continued) November 8, 1954
ovember o,

Readjusted shims. Regulating rod position, 3.15 in. Reactor mean,
1257°F,

Regulating rod position now 9.55 in,, moved 6.40 in, since 0242
when final rod adjustment was made. Reactor mean temperature,
1286°F; 29°F rise since 0242.

End of experiment,

EXPERIMENT L-9
Objectivet Set Reactor at 1 kw and Adjust Chambers

Reactor already at 1 w,
Took data.

Pulled safety chamber 1.
Raised power to ~20 w,
Raised power to ~40 w.

Started pulling log N chamber,
Reactor scrammed.
40-w nominal power,

Pulled log N chamber all the way within shield.
Pulled log N out of shield about 20 in.

Data recorded.
Raised power to ~400 w.

On manual. Pulling micromicroammeter chamber,
Set micromicroammeter to exactly 500 w.
Micromicroammeter 50 on 1 x 10~8; corresponds to 500 w.
Put reactor on 30-sec period and let safety chambers scram reactor,
EXPERIMENT H-.1
Objective: Approach to Power (10<kw run)
Instruments on scale
Reactor up to 10 kw.

Took data.
Pulled safety chambers,

Took data.
Took data,
Scrammed, Fission gases in basement,
EXPERIMENT H-2
Objective: 10-kw Run to Monitor Gas Activity

Novémber 9, 1954

Instruments on scale,

Up to 10 kw.
Data taken.

Data taken.
Decreased power to 5 kw,

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4

Time
1227

1245
1246

1520
1523
1526

1529
1530
1551
1600
1621
1625
1626
1643

1656

1704

1715
1727

1740
1749
1801

1810
1820
1825

1845

1854
1900
1907
1915
1919
1938
1945
2027

EXPERIMENT H-2 (continued) N ber 9, 1954
ovember 7,

Data taken.
Data taken,

Reactor scrammed.
During run, the estimated pit activity from RIA-4 was 0,032 pc/em3,

EXPERIMENT H-3
Objective: 100-kw Run

Instruments on scale. Period, ~17 sec,

Reactor at ~ 100 kw.

Fuel and sodium system barrier doors raised, Fuel system prime mover
started,

Blower at ~300 rpm.

Safety chambers scrammed reactor at ~ 130 kw.

Instruments on scale, Period, ~30 sec.

Reactor scrammed by safety level (set too low).

Instruments on scale. Period, ~27 sec,

Reactor on servo at 50 kw, Safety chambers reset,

Reactor off servo, Power raised.

Power leveled out at ~250 kw. Data taken,

Elevated mean temperature (withdrew shim rods). Changed scales on
reactor AT,
Data taken,

Power raised to ~ 500 kw.
Data taken,

Front sodium system blower brought to ~ 500 rpm.
Back sodium system blower brought to ~ 500 rpm.
Data taken, |

Increased power,
1-Mw nominal power,
Data taken.

Data taken.

Reduced power.

Fuel system blower off,
Data taken, ~ 115 kw.
Rods going in.

Sodium system blowers off,
Started increasing power,
Reactor power, 50 kw.
Reactor power, ~ 100 kw,

EXPERIMENT H-4

Objective: Fuel and Reactor Temperature Coefficient Determinations

2045
2100
2102
2103

Reactor power level at about 100 kw.

Turned on fuel system helium blower, 350 rpm.
Readjusted shims,

Blower off,

225

 

 
 

 

 

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Run.

226

EXPERIMENT H-4
November 9, 1954

Time

2121 Fuel system blower on, 340 rpm.,
2124 Readjusted shim.

2126:30 Prime mover off. Servo off,
2129 End of run.

2156 Reactor at 10 kw,

EXPERIMENT H-5

Objectives Over-all Reactor Temperature Coefficient Determination

Reactor at 100 kw on flux servo. Allowed to heat up from 1260 to 1313°F,

2225 Data taken,
2243 Data taken.
2300 Data taken,

EXPERIMENT H-6

Objective: Startup on Temperature Coefficient

2307 Increased power,

2315 Power, ~ 1.3 Mw {fuel system blower speed at maximum).

2327 Took data,

2339 Reduced power to ~200 kw, Pressure shell temperature getting high.
2348 Inserted shims to drop reactor mean temperature,

2353 Ran blower up from 0 to 1700 rpm (maximum speed).

2400 Mark charts,

2405 Ran regulating rod in from 7 in,; then pulled out all the way.

November 10, 1954

0016 Reduced fuel system helium blower speed to zero,

0019 Inserted regulating rod from 14 to 2 in. Reactor subcritical.

0026 Ran fuel system helium blower speed from 0 to 1700 rpm.

0028 Low heat exchanger temperature reverse,

0033 Stopped fuel system helium blower,

0037 inserted shim rods and reduced power, Stopped sodium system
helium blowers.

0106 Reactor left at 10 to 100 kw to maintain temperature on fuel lines.

0115 Reactor scrammed due to calibration of fuel flowmeter,

0137 Reactor power at ~ 100 kw,

0204 Reactor mean temperature at ~ 1300°F; power at ~75 kw and falling
slowly due to heating of fuel, 5.25 in, of regulating rod movement = ~30°F
temperature rise = 5.5 x 10=3 (Ak/k)/°F,

PRACTICE OPERATION

Objective: To Familiarize Crews with Reactor Operation

0503 Reactor power, ~ 250 kw.

0538 Reactor power, ~ 500 kw,

0545 Stopped fuel and sodium system blowers. Power leveled at ™ 50 kw,
0548 All blowers back on at about 500 rpm,

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(2)

Time

0552
0554
0600
0602
0615
0625
0630
0844
0852
0855
0858
0905
0909

0914
0917
0924

1040
1045
1057

1110
1114
1128
1150
1200
1206
1229
1241

1300
1305
1310
1313
1315

PRACTICE OPERATION (continued)
November 10, 1954

Reactor power, ~ 750 kw.

All blowers off.

Blowers on. Power, ~750 kw.

Blowers off.

Reactor power, ~ 500 w,

Reactor power, ~ 100 kw,

Reactor power, ~ 50 kw,

Reactor power, ~50 kw, Took data.

Started fuel system blower prime mover,

Started fuel system blower.

Leveled off power at ~200 kw,

K. Z. Morgan reports short-lived air activity in crane bay.
Blower off, Prime mover off. End of experiment. Reactor leveled out

at ~ 50 kw.

EXPERIMENT H-7

Objective: Sodium Temperature Coefficient Determination

Ran sodium system helium blowers from 0 to ~ 2000 rpm,

Stopped sodium system blowers. Peak power, ~ 150 kw.

Leveled out. Charts pulled. Sodium temperature coefficient is
negative,

EXPERIMENT H-8

Oblective: Measure Effect of a Dollor of Reactivity

Reactor power, ~ 50 kw, Started sodium system blowers,
Reactor power, ~ 1 Mw,

Took data,

Pulled regulating rod one dollar in 0.61 min.
Inserted regulating rod one doltar, Charts pulled,
Scrammed while pulling charts,

Reactor power, ~ 150 kw.

 Reactor power, ~1 Mw,

Sodium system blowers up to 1000 rpm each,
Reduced blower speed to zero,
Reactor power, ~ 100 kw,

PRACTICE OPERATION (continued)

Blowers started, Blowers at 500 rpm.

Reactor power, ~500 kw, Extracted power, ~ 570 kw.

Blowers to 1000 rpm.

Reactor power, ~ 900 kw,

Regulating rod withdrawn 2 in. Extracted power, 920 kw (fuel)
+ 96 kw (sodium) = 1.016 Mw.

227

 

 

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228

EXPERIMENT H-9
November 10, 1954

Objective: Over-all Reactor Temperature Coeffizient Determination

Time

1400
1422

1452
1458
1505
1515

1523

Reactor power, 300 kw. Fuel mean temperature raised to 1350°F,

Reactor on servo at 36 kw. Front sodium system blower on, 600 rpm,
Reactor cooling.

Regulating rod reached lower limit,

Reactor at 100 kw on servo,

Going up to 1 Mw.

Mean temperature was raised to 1350°F; now lowered to 1325°F to avoid
hot places on lines.

Mean temperature lowered to 1315°F at which point low heat exchanger
temperature interlock reversed helium blowers.

EXPERIMENT H-10

Objective: Moderator Temperature Coefficient Determination

1550
1600

1605
1606

1611
1613
1625

1630
1735

1737:30
1745

1747
1752

1755
1802

1805
1808

Lowered reactor to ~ 400 kw,

Fuel temperature, 1317°F; regulating rod position, 7.2 in.; reactor outlet
sodium temperature, 1280°F; reactor power, 500 kw.

Front sodium system blower, 870 rpm. Back sodium system blower, 1260 rpm,

Reduced back sodium system blower speed to 870 rpm. Regulating rod
position, 7.5 in,; fuel temperature, 1317°F; sodium temperature, 1280°F.

Temperatures same as at 1606. Regulating rod position, 8.3 in.

Front sodium system blower speed reduced to 670 rpm; back sodium system
blower speed to 850 rpm. Regulating rod position, 8.7 in,; fuel temper-
ature, 1319°F; sodium temperature, 1278°F.

Front blower at 650'rpm; back blower at 640 rpm. Regulating rod position,
7.6 in,; sodium temperature, 1278°F,

Reactor power at 1 Mw for demonstration to Air Force personnel.

Regulating red position, 8.0 in.; reactor mean temperature 1313°F; front
blower at 960 rpm; back blower at 1050 rpm; sodium inlet temperature,
1265°F (thermocouple 3-CR-1) sodium outlet temperature, 1282°F
(thermocouple 3-CR-5).

Front blower at 1180 rpm.

Reactor mean temperature, 1314°F; regulating rod position, 8.0 in. Front
blower at 1160 rpm; back blower at 1050 rpm. Sodium inlet temperature,
1255°F; sodium outlet temperature, 1280°F.

Back biower at 1280 rpm.

Reactor mean temperature, 1313°F. Regulating rod position, 7.6 in.
Front blower at 1170 rpm; back blower at 1260 rpm. Sodium inlet
temperature, 1248°F; sodium outlet temperature, 1278°F.

Front blower raised to 1350 rpm.

Regulating rod position, 6.6 in. Reactor mean temperature, 1308°F.
Front blower at 1340 rpm; back blower at 1250 rpm. Sodium inlet
temperature, 1238°F; sodium outlet temperature, 1272°F,

Front blower raised to 1490 rpm, '

Regulating rod withdrawn 1.3 in. to regain 5°F lost in reactor mean
temperature, Rod position = 7.3,

)

 

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Run

Time

1812

1816
1825

EXPERIMENT H-10 {(continued) N ber 10, 1954
ovember ’

Regulating rod position, 8.4 in. Reactor mean temperature, 1311°F.
Front blower at 1480 rpm; back blower at 1240 rpm. Sodium inlet
temperature, 1230°F; sodium outlet temperature, 1268°F.

Back blower speed changed to 1500 rpm.

Regulating rod position, 8.9 in. Reactor mean temperature, 1313°F,
Front blower at 1470 rpm; back sodium blower at 1480 rpm.

Sodium inlet temperature, 1225°F; sodium outlet temperature, 1268°F.

EXPERIMENT H-11

Objectives 25-Hour Run at Full Power to Observe Xenon Poisoning

1935

0304

0400
0635

0750
0825
0915
0925

1020
1100
1130

1145
1206
1435

1445

Since 1825, the reactor has been steady at 90% full power on servo; it
is now off servo, The 25-hr xenon run is considered to have started

at 1825.
November 11, 1954

Cooling water flow to back sodium heat exchanger decreased. Sodium
AT fell ~3°F. No explanation as yet.

Drop in sodium AT due to heaters for annulus helium being shut off.

Withdrew regulating rod from 10.3 to 10.4 in. Reactor mean temperature
dropped to 1309°F from an initial temperature of 1311°F.

Regulating rod withdrawn from 10.4 to 10.5 in.

Micromicroammeter took a small dip (53 to 52) for no apparent reason.

Adding helium to sodium and fuel system ducts.

Fuel flowmeter coils calibrate OK. There is a possibility of something
binding in the Rotameter to give the erratic readings, Switched to
operation on No, 3 coil.

Regulating rod withdrawn ~0.05 in. Indicator now reading 9.25 in.

Pile period shows unexplained excursion to 400 sec.

Noticed fluctuations in the fuel pump ammeter recorder corresponding
to fluctuations in the Rotameter chart, These fluctuations appear to
be between 9.0 and 93/4 amp.

Noticed same pump current and flow level fluctuations as at 1130.

Switched back to No. 5 coil on fuel flow Rotameter.

Regulating rod withdrawn 0.1 in.

For the past 3/4 hr have noticed corresponding fluctuations in the
following instruments: fuel flow Rotameter, fuel motor current
recorder, period meter, log N recorder, safety levels 1 and 2, reactor
AT, the six individual tube AT’s, and the micromicroammeter. All
these fluctuations at present are small, especially on the micro-
microammeter. The largest fluctuations appear to be on the pump
motor current, between 9 and 10 amp; the pile period meter, ~ to + 400
sec; and the six AT recorders, + 2 to 3°F, No extraneous audio noise
has been detected. Noticed increase in fuel pump pressure since (0800
of 0.2 psi. At 0800 pressure was 0.2 psi, at present it is 0.4 psi.

229

 

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E

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230

 

Run

()

Time

1935

2032

EXPERIMENT H-11 (continued) N ber 11, 1954
ovember '

End of xenon run. Regulating rod has been withdrawn ~ 0.3 in., which
corresponds to 0.01% Ak/k as compared to ~0.30% Ak/k if all xenon
had stayed in system,

Had demonstration for Air Force personnel, mostly 1-Mw nominal power,

Set reactor back to original condition left at 1935.

EXPERIMENT H-12

Objective: Stop Sodium System Blower and Observe Effect on Power

2105

2113
2119

2150
2155
2158
2200
2201
2204

2210
2230

2237

2237

0605
0735

0835

Objective:

0837
0855

0907

0910
0917

Took data.

Sodium system blower motors off.

Took data, Blowers back on, Maximum deviation of micromicroammeter was
from 53 to 45, 15%.

Reduced power by cutting off fuel system blower,

Reduced speed on sodium system blower,

Regulating red inserted; reactor power, ~ 10 kw,

Started fuel system blower motor,

Started fuel system helium blower,

Approached low heat exchanger temperature reverse (1150°F) which shut off
fuel system blower motor, Withdrew regulating rod.

Reactor power back up to 1 Mw,

Cut off fuel system biower, Cooling reactor with sodium system blowers
only.

Reactor power, ~ 100 kw.

EXPERIMENT H-13

Objective: Measure Xenon Buildup at 1/]0 Full Power
Consider experiment to have started at 2237 at ]/10 full power, i.e., ~100 kw.

November 12, 1954

Regulating rod withdrawn from 5.05 to 5.15 in. Reactor mean temperature
dropped from 1304 to ~1303°F.

Sodium AT up due to addition of helium in rod cooling system. Reactor
mean temperature down to 1302°F,

No appreciable xenon buildup. End of experiment.

EXPERIMENT H-14

Determine Maximum Reactor Power and Characteristics of Power Operation

Demonstration of reactor operation.

Reactor power up to 1 Mw., Maximum fuel system helium flow. Reactor
mean temperature, ~1340°F,

No. 2 rod cooling helium blower on slow speed. Helium pressure,
25 in. H, 0.,

No. 2 rod cocling blower off. Helium pressure, 9 in. H,0.

No. 2 rod cooling blower on,

,.
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gy I o g N TN T W T

 
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e

Run
(2)

Time

0919

0922
0935

1015
1055
1102
1109
1110
1111:30
1112
1113

1115
117

1125

1140
1150
1230
1244
1306
1306
1308
1325
1326
1330
1331

1334

1336
1350
1400
1615

1930
2004

EXPERIMENT H-14 (continued) N ber 12, 1954
ovemoer ’

No. 2 rod cooling blower off. Small wobble in AT probably due to
helium blowing on thermocouple. No systematic effect on micro-
microammeter can be observed.

Started demonstration. Average reactor power, ~ 1 Mw,

Demonstration ended. Reactor power back to T Mw,

Reactor mean temperature raised to 1350°F.

Fuel system blower off.

Power leveled off at ~350 kw. Sodium system blowers off.

Power leveled off at ~210 kw, Rod cooling blowers off,

Power leveled off at 200 kw.

Started bringing reactor up to full power,

14-sec period observed,

Heat exchanger outlet temperature reduced from 1380 to 1175°F.
Log N from 4 to 30.

Back to power of >1 Mw.

Moved shim and regulating rods to obtain new heat balance.

Sodium system blower speed up to 2000 rpm, Extracted power: from
fuel, 1920 kw; from rod cooling, 20 kw; from sodium, 653 kw; total,
2.6 Mw,

No. 2 rod cooling blower started.

Front sodium system blower off; reactor power reduced to ~ 100 kw.

Reactor mean temperature and power leveled off; power, ~ 1.5 Mw,

Reduced fuel system helium blower speed to minimum,

Fuel AT nearly constant.

Raised fuel system helium blower speed to maximum,

Fuel AT essentially restored.

Fuel system blower turned on; speed reached 1500 rpm,

Blower speed reduced to 1000 rpm.

Reactor AT about 200°F,

Regulating rod withdrawn to raise mean temperature from 1320 to
1325°F.

Regulating rod inserted to lower mean temperature from 1325 to
1320°F.

Regulating rod withdrawn to raise mean temperature,

After demonstration, power leveled off to 1.5 Mw.

Demonstration of reactor operation.

End of demonstration of reactor operation, Reactor run more or less
steadily at 2 Mw until time for scram,

Reactor power up to ~2.5 Mw,
Reactor scrammed.

231

R e p——— e

 
i

 

 

Appendix U
THE ARE BUILDING

The ARE building, shown in Fig. U.1, is a mill
type of structure that was designed to house the
ARE and the necessary facilities for its operation.
The building has a full basement 80 by 105 ft,
a crane bay 42 by 105 ft, and a one-story service
wing 38 by 105 ft. The reactor and the necessary
heat disposal systems were located in shielded
pits in the part of the basement serviced by the
crane. One half the main floor area was open to
the reactor and heat exchanger pits in the basement
below; the other one half housed the control room,
office space, shops, and change rooms.

In one half the basement were the shielded
reactor and heat exchanger pits; the other one
half of the basement was service area and miscel-
laneous heater and control panels. The control
room, office space, and some shops were located

on the first floor over the service area. The first
floor does not extend over the one half of the
basement that contains the pits. The crane is
a floor-operated, 10-ton, bridge crane having a
maximum lift of 25 ft above the main floor level.
Plan and elevation drawings of the building are
shown in Figs. U.2 and U.3.

The entire reactor system was contained in three
interconnected pits: one for the reactor, another
for the heat exchangers and pumps, and a third
for the fuel dump tanks. These pits, which were
sealed at the top by shielding blocks, were located
in the large crane bay of the building. The crane
bay was separated from the control room and
offices, and the heating and air conditioning
systems maintained the control room at a slightly
higher pressure than that of the crane bay.

 

Fig. U.1. The ARE Building.

232

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4 4! A . 3§
LEGEND
ORNL-LR~-DWG 6553
STANDARD MIX CONCRETE
|—" c V277777 SOLID CONCRETE MASONRY-STANDARD MIX
81 ft : SOLID CONCRETE BLOCKS LAID DRY - STANDARD MIX
i ; A REMOVABLE SPECIAL MIX BLOCKS
; 4zt i : C
;
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i = o pemeeseecs o 1 ey — ,
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CRAFT FOREMAN INSTRUMENT ROOM
CRANE BAY AND SHOP
:, FIELD ENGINEERS — :
DOWN E ‘ CONTROL PIT
: A :
T 1 + §
! l
|
__ AMPLIFIER /<
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d OFFICE BASEMENT INSTRUMENT
PANEL
TEST |[i{PITS § D TEST PIT
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b S 8 K
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Z 2 /<
= o =
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. & L H HEAT EXCHANGER PIT
B8 o N g 8 B 8
. L OFFICE :
t 1+t A
| ] Y ‘
i RELAY CABINET
I
i POWER EGQUIPMENT ROOM
| ;
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FIRST FLOGOR PLAN BASEMENT PLAN

Fig. U.2. Plan of the ARE Building.

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LEGEND

STANDARD MiX CONCRETE

SOLID CONCRETE MASONRY-STANDARD MIX

SOLID CONCRETE BLOCKS LAID DRY - STANDARD MIX

- REMOVABLE SPECIAL MIX BLOCKS

 

 

 

 

 

 

 

 

 

 

 

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. I i

BIBLIOGRAPHY

Aircraft Nuclear Propulsion Project Quarterly Progress Reports for the Periods Ending:

March 10, 1951
“June 10, 1951
September 10, 1951
December 10, 1951
March 10, 1952
June 10, 1952
September 10, 1952
December 10, 1952
March 10, 1953
June 10, 1953
September 10, 1953
December 10, 1953
March 10, 1954
June 10, 1954
September 10, 1954
December 10, 1954

ANP.60

ANP-65

ORNL-1154
ORNL-1170
ORNL-1227
ORNL-1294
ORNL-1375
ORNL-1439
ORNL-1515
ORNL-1556
ORNL-1609
ORNL-1649
ORNL.-1692
ORNL-1729
ORNL-1771
ORNL-1816

235

 

T T

 
 

 

 

 

DATE

1948

12-15-49
5-21-51
4-17-51

6-5-51

7-25-51
8-13-51

8-29-51

10-15-51

1-5-52
1-8-52

1-11-52
1-14-52

1-22-52
1-29-52

1-29-52

3-10-52

3-20-52

3-24-52
3-28-52

4-16-52

4-22-52

5-8-52
6-2-52

8-8-52

236

TITLE

Metals Handbook (Nicke! and Nickel Alloys,
p 1025-1062)

The Properties of Beryllium Oxide
Activation of Impurities in BeQ

Perturbation Equation for the Kinetic Re-

sponse of a Liquid-Fuel Reactor

The Contribution of the (n,2rn) Reaction to
the Beryllium Moderated Reactor

Radiation Damage and the ANP Reactor
Alkali Metals Area Safety Guide

Physics Calculations on the ARE Control
Reods

Some Results of Criticality Calculations on
BeO and Be Moderated Reactors

Physics of the Aircraft Reactor Experiment

Statics of the ANP Reactor — A Preliminary
Reporf

The ARE with Circulating Fuel-Coolant

A Flux Transient Due to a Positive Reac-
tivity Coefficient

Heat Transfer in Nuclear Reactors

Safety Rods for the ARE

Effect of Structure on Criticality of the ARE
of January 22, 1952

A Simple Criticality Relation for Be Moder-

ated Intermediate Reactors

Health Physics Instruments Recommended
for ARE Building

Induced Activity in Cooling Water - ARE

Optimization of Core Size for the Circu-
lating-Fue! ARE Reactor

Physics Considerations of Circulating Fuel

Reactors

Note on the Linear Kinetics of the ANP

Circulating-Fuel Reactor
Statics of the ARE Reactor

Reactor Program of the Aircraft Nuclear

Propulsion Project

The ARE Critical Experiment

AUTHOR(s)

American Society for Metals

M. C. Udy, F. W, Boulger
W. K. Ergen
N. Smith et al,

C. B. Mills
N. M. Smith, Jr.

L. P, Smith

Y-12 Alkali and Liquid Metals
Safety Committee

J. W. Webster
R. J. Beeley

J. W. Webster
0. A, Schulze

C. B. Mills
C. B. Mills
C. B. Mills
C. B. Mjlls
R. N. Lyon
W. B. Manly
E. S. Bomar
C. B. Mills
C. B. Mills

T. H, 1. Burnett

T. H. J. Burnett
C. B. Mills

W. K. Ergen

F. G. Prohammer

C. B. Mills
Wm. B. Cottrell (ed.)

C. B. Mills
D. Scott

REPORT NO.

BMI-T-18
Y-F20-14
ANP-62

Y-F10-55

ANP-67
Y-8

Y-F10-71
ANP.66

Y-F10-77
Y-F10-81

Y-F10-82
Y-F10-63

CF-52-1-76
CF-52-1-192

Y-F10-89
Y-F10-93
CF-52-3-147

CF-52-3-172
Y -F 10-96

Y-F10-98
Y-F10-99

Y-E10-103
ORNL-1234

Y-F10-108

-

o}

T TR TR I R T W -

e o vy

e b vy EEE T

-

¥ vemes W R ROTBP

o R R R T T

 

i
¥
S
gr-*.
¥

Y LT

e t

reeee

 
"

DATE

7-30-52

9-52
9-52
9-25-52

10-22-52
11-1-52

11-24-52

11-13-52
11-17-52

11-25-52

1-7-53
1-9-53
1-12-53
1-27-53

1-27-53
1-27-53

2-45
2-11-53

2-26-53

3-17-53

3-30-53

4-7-53

TITLE

Loading of ARE Critical Experiment Fuel
Tubes

Welding Procedure Specifications

Welders Qualification Test Specifications

Radiation Through the Contrel Rod Pene-
trations of the ARE Shield

Xenon Problem — ANP

Structure of Norton’s Hot-Pressed Beryllium
Oxide Blocks

Aircraft  Reactor  Experiment Hazards

Summary Report
Effects of lrradiation on BeO

Structure of Be0 Block with the 11/8

in,

Central Hole

An On-0Off Serveo for the ARE

Delayed Neutron Damping of Non-Linear

Reactor Oscillations

ARE Regulating Rod

A Guide for the Safe Handling of Molten
Fluorides and Hydroxides

Delayed Neutron Activity in a Circulating
Fuel Reactor

Components of Fluoride Systems

Delayed Neutron Activity in the ARE Fuel
Cireuit

Some Engineering Properties of Nickel and

High-Nickel Alloys

Heating by Fast Neutrons in a Barytes Con-
crete Shield

Methods of Fabrication of Control and
Safety Element Components for the Air-
craft and Homogeneous Reactor Experi-

ments

Corrosion by Molten Fluorides

The Kinetics of the Circulating-Fuel

Nuclear Reactor

Minutes of the Final Mee*ing of the ARE

Design Review Committee

AUTHOR(s)

D. Scott

P. Patriarca
P. Patriarca

H, L, F, Enlund

W. A, Brooksbank
L. M, Doney

J.H. Buck
W. B, Cottrell

G. W, Keilholtz
L. M. Doney

S. H. Hanauer
E. R. Mann
J. J. Stone

W. K, Ergen

E. R. Mann
S. H, Hanauer

Reactor Components Safety
Committee

H. L. F. Enlund

Wm, B, Cottrell
H. L. F. Enlund

B. B, Betty
W. A. Mudge

F.H. Abernathy
H. L. F. Enlund

J. H. Coobs
E. S. Bomar

L.S. Richardson
D. C. Vreeland
W. D. Manly

W. K. Ergen .

W. R. Gall

REPORT NO.

Y-B23-9

PS-1
QTS-1
Y-F30-8

CF-52-10-187
CF+52-11-12

ORNL-1407

CF-52-11-85
CF.52-11-146

CF.52-11-228

CF-53-1-64
CF-53-1-84
Y-B31-403
CF-53-1-267

CF.53-1-276
CF-53-1-317

Mech. Eng. 73,
123 (1945)

CF-53-2-99

ORNL-1463

ORNL-1491

CF-53-3-231

CF-53-4-43

237

 

T T T T T e T R B Ty

- e rrrm——

 
 

 

 

 

 

 DATE.

5-18-53
6-19-53
7-20-53
7-20-53
7-20-53

8-3-53
8-10-53

9-1-53
9-1-53
9-3-53
9-22-53

© 9-25-53
10-18-53

12-1-53
12-22-53

12-18-53

3-2-54
4-7-54
5-7-54
7-20-54

7-27-54
.7-31-54

-

8-28-54

To\"lv_ae issued

TITLE

ARE Control System Design Criteria

The Stability of Several |ncone!-UF4 Fused
Salt Fuel Systems Under Proton Bombard-

ment

. Current Status of the Théory of Reactor

Dynamics
Interpretation of Fission Distribution in

ARE Critical Experiments

Composition of ARE for Ciéticality Calcu-

lations
ARE Fuel Requirements

Thermodynamic and Heat Transfer Analysis
of the Aircraft Reactor Experiment

ARE Fuel System

Static Analysis of the ARE Criticality Ex-
periment (A Preliminary Report Pending
Reception of the ARE Criticality Report)

Experimental Procedures on the ARE (Pre-
limindry)

The General Methods of Reactor Analysis
Used by the ANP Physics Group

Supplement to Aircraft Reactpr Experiment
Hazards. Summoary Report (ORNL-1407)

Preliminary Critical Assembly for the Air-

craft Reactor Experiment
ARE Design Data

The Inhour Formula for a Circulating-Fuel
Nuclear Reactor with Slug Flow

Analytical and Accountability Report on
ARE Concentrate

ARE Design Data Supplement
ARE Instrumentation List
Analysis of Critical Experiments

ARE Operating Procedures, Part |, Pre-

Nuclear Operation

ARE Operating Procedures, Part 1l, Nuclear

Operations

Fuel Activation Method for Power Determi-
nation of the ARE

Critical Mass of the ARE Reactor

Stress Analysis of the ARE

AUTHOR(s)

F. P, Green

W. J, Sturm
R. J. Jones
M. J. Feldman

W. K Ergen
Joel Bengston
J. L, Meem
J. L. Meem

B. Lubarsky
B. L. Greenstreet

 G. A. Cristy

C. B, Milis

J. L. Meem
C. B. Mills

E. S. Bettis
W. B. Cottrell

D. Callihan
D. Scott

W. B. Cottrell
¥W. K, Ergen

G. J. Nessle

W. B. Cottrell
R. G. Affel
C. B. Mills
W. B, Cottrell

J. L. Meem
E. B. Johnson
C. B, Mills

R. L. Maxwell
J. W. Walker

REPORT NC

CF-53-5-238
ORNL-1530

CF-53-7-137

CF-53-7-190

Secret rough dr

CF-53-8-2
ORNL -1535

CF-53-9-2

CF-53-9-19

CF-53-9-15

ORNL -1493

- CF-53-9-53

ORNL-1634

CF.53-12-9
CF-53-12-108

CF-53-12-112

CF-54-3-65
CF-54-4-218
CF-54-5-51
CF-54-7-143

CF-54-7-144
CF-54-7-11

CF-54-8-171
ORNL -1650

£

¥

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