ORNL-TM-0911.txt

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OAK RIDGE NATIONAL LABORATORY l
aperated by
UNION CARBIDE CORPORATION
NUCLEAR DIVISION
for the
U.S. ATOMIC ENERGY COMMISSION

ORNL- TM- 911

 

 

RELEASED FOR ANNOUNCEMENT

1IN NUCLEAR SCTENCE ABSTRACTS

 

MSRE DESIGN AND OPERATIONS REPORT
Part XI
TEST PROGRAM

R.H. Guymon
P.N. Haubenreich
J.R. Engel

THIS 07 C-.‘ "T'T H”‘ pEEY wn,rtv-wED

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NOTICE This document contains information of a preliminary nature /’J_]/

and was prepared primarily for internal use ot the Oak Ridge National
Loboratory. It is subject to revision or correction and therefore does
not represent a final report.

  
 
 
  
  
 

 

 

 

LEGAL NOTICE

This report was prepared as an account of Government sponsored work. Neither the United States,

nor the Commission, nor any person acting on behalf of the Commission:

A. Makes any warranty or representation, exprassed or implied, with respect to the accuracy,
completeness, or usefuiness of the information contained in this report, or that the uss of
any information, apparatus, methed, or process disclosed in this report may not infringe
privately owned rights; or

8. Assumes any liabilities with respect to the use of, or for damages resulting from the use of
any information, apparatus, methaod, or process disclosed in this report.

As used in the above, “‘person acting on behalf of the Commission® includes any smployes o

contractor of the Commission, or employee of such contractor, to the extent that such employee

or contractor of the Commission, or employee of such contractor prepares, disseminates, or
provides access to, any infermation pursuant te his empleyment or contract with the Commission,

or his employment with such contractor.

 

 

 

 

 
ORNL-TM-911

RELEASED FOR ANNOUNCEMENT

 

‘k

TENCE ABSTRACTS

 

MSRE DESTGN AND OPERATIONS REPORT

Part XTI

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¥
PREFACE

This report is one of a series that describes the design and opera-
tion of the Molten=-Salt Reactor Experimeant. All the reports are listed

below.

*
ORNIL-TM-T728 MSRE Design and Operations Report, Part I,
Description of Reactor Design by
R. C. Robertson

ORNL-TM-T729 MSRE Design and Operations Report, Part IT,
Nuclear and Process Instrumentation, by
J. R. Tallackson

*
ORNL-TM-T730 MSRE Design and Operations Report, Part III,
Nuclear Analysis, by P. N. Haubenreich and

J. R. Engel, B. E. Prince, and H. C. Claiborne

ORNL-TM-T731 MSRE Design and Operations Report, Part IV,
Chemistry and Materials, by F. F. Blankenship
and A, Taboada

ORNIFTM—TBB* MSRE Design and Operations Report, Part V,
Reactor Safety Analysis Report, by S. E. Beall,
P. N. Haubenreich, R. B. Lindauer, and
J. R. Tallackson

*
ORNL-TM-7373 MSRE Design and Operations Report, Part VI,
Operating Limits, by S. E. Beall and
R. H. Guymon

*
ORNL-TM-907 MSRE Design and Operations Report, Part VII,
Fuel Handling and Processing Plant, by
R. B. Lindauer
*
CRNT~-TM-908 MSRE Design and Operations Report, Part VIII,
Operating Procedures, by R. E. Guymon

*

ORNI~TM-909 MSRE Design and Operations Report, Part IX,
Safety Procedures and Emergency Plans, by
A, N. Smith

¥
ORNL-TM-910 MSRE Design and Operations Report, Part X,
Maintenance Equipment and Procedures, by
E. C, Hise ard R. Blumberg

 

Tagued.

*¥%
These reports will be the last in the series to be published.
iv

*
ORNL-TM-911 MSRE Design and Operations Report, Part XI,
Test Program, by R. H. Guymon,
P. N. Haubenreich, and J. R. Engel

MSRE Design and Operations Report, Part XTI,
Lists: Drawings, Specifications, ILine
Schedules, Instrument Tabulations
(Vol.1l and 2)
FOREWORD

The reader of this report should be aware that the date of issue is
somewhat misleading — this description of the MSRE test program was
written before the fact and has not been updated. Preparation of this
document began in 1964 and continued through 1965, each section being
issued in draft form as it was completed and before operations entered
that phase of the program. Inevitably some situations arose (the offgas-
system troubles, for example) that added some tests and some time to the
planned program. No major deviation from the broad outline has occurred,
however, and we believe it worthwhile to issue this report as a convenient
record of what was planned for the MSRE. For an account of what actually
transpired, the reader should turn to the Molten Salt Reactor Project

semiannual progress reports.
3

vii

 

- CONTENTS

Page

INTRODUCTION e oo vennoeeounnnseseconsssnssssssssonenosncsaonosnaes i- 1
1.1 OBIECTIVES +vvevveensoasscnsannsssss hesesevecnsa Cerseen et 1- 1
1.2 AREAS TO BE INVESTIGATED...esvevevnnssrnnnesvas cesseuncaran 1- 1
1.2.1 Chemistry and MaterialS.eeeeeveevaneas et eeees 1- 1
1.2.2 Engineering...... Gttt s ettt ety 1- 2
L1.2.3 Reacltor PhySiCSieerivienrerocitennoenennesvonsanns 1- 2

1.3 DPHASES OF PROGRAM +vvreovoessons ce e fevrvserties e ranuas 1- 3
1.4 DOCUMENTATION v v eevsoessvessussessstnatnnannensnenseassoss 1- 4k
NONNUCIEAR TESTING seversveversenens Ce et s ettt sty e- 1
5.1 ORTECTIVES secevevrssnsns ceeees cecaeeaas Cieseasenesrnasenaane 2= 1
2.2 EXPIANATION OF PROCEDURES ¢evsvoecscsvsetsons te et asar s vsn s 2- 1
2.3 PROCEDURES v ecsesss e ves e iennssesr et u et et e 0 e T
2.3.1 Fuel System «seeesessesnvausss Ceesasssererrrasresos 2- 2
2.3.2 Fuel-Drain-Tank System eecevesensnve “revesssrarsnanne 2- 5
2.3.3  (C001ant SYSTEm eeervoerecnonenirarsassonsnsransases o~ 7
2.3.4  Coolant Drain-Tank SyStem +teeeceecaeavereeans e 2-10
2.3.5 Cover-Gas and Offgas Systems .«... trereseiesearanas 2=-11
2.3.6 01l SyStemSescrenreerrosreorstssnaacsasosnanccssonns e-12
2.3.7 Chemical Processing System seeeessceveiirinrinanean. 2-13
2.3.8 Ileak-Detector System..ceecesoevsncrinnansusss ceenen 2-1h
2.3.9 Cooling-Water SysStems -cevierrerorrrrenvenrensnons 2-15
2.3.10 Component-Cooling Systems veeeveeesee. Chreeear e, 2-16
2.3.11 Instrument-Air and Auxiliary-Air Systems........... 2-17
2.3.12 Instrumentation............ srs et tesvecsanervasres e 2-18
2.3.13 Electrical Systemetcveecerorevecrnannans besensenan 2-20
2.3.1L Shield and Containment +vevvveceererrvsnsvansans ee. 221
2.3.15 Ventilation System esescveverecrieirrerscrvsoveronons 2-23
2.3.16 Liquid-Waste SysteMmecsveerocoess Ce e e 2-2k
2.3.17 Samplers sscscsassccooans tetsecesrert ettt un 2~25
2.3.18 Control ROAS ccsesesvecsssessetcorssossocescanansssns 2-25
2,3.19 Heaters reeeersavecencanan. Ceeeresieraaean Cer e 2-26
3.

L.

5.

 

2.3.20 TFreeze ValveS ceevoses tesaceesscescssencscssssectras
2.3.2]1 MlscellanNeOusS siescescscrccsosscecosas teevcerraancas
2,3.22 Entire Plant ..ceececoececess ceescscrrareesarnansnanne
ZERO POWER EXPERIMENTS +veeeceeovooosoaassoosaneas teovsesscennns
3.1 OBJIECTIVES ..svveevucesees teeseeseneresens sevesassssreans cees
3.2 PROCEDURES .eeesvecsnssscans tectecearenssscaesrrosaene ceeneee
3.2.1 Initial Critical Experiment....ceeves.. traseneraen .
3.2.2 Calibration of Control ROGS.e.eeess ceveoces erescncs
3.2.3 EBvaluation of Nuclear Parameters ..eeeeeveoss cesens

3.2.4  Preliminary Studies of DynamicsS ..eveeveceocennnans

3.2.5 Evaluation of Neutron Sources and Future

 

 

Requirements of the External Source.......... ceue

3.2.6 Chemical ANBLYSES .vuusveronverrocanrnaroranssns coe

I0W POWER MEASUREMENTS .oeveee ceresssessssaas Peetscnvavarnsncnse
Lol OBJECTIVES sieeeeerncncusocncosnoasncronnsaseans ceeseeraais
4.2 PROCEDURES ...... Gescsanssases creecvane ceennee ceesccessa .o
4.2.1 Shielding and Containment SUTVEYS «eeeveseeraoennn.
L.2.2 Calibration of Nuclear-Power Instruments ..........
k.2.3 Power Coefficient of Reactivity +....... e eenrnenn
L.2.4  Xenon Poisoning ..... Ceveceneaes Cereceerceecnna cees
L,2.5 On-Line Analysis of Operation .eveseeecesevoneesns .
L.2.6 Establishment of Baseline for Chemical Analyses....
4.2.7 Intermediate Dynamics Studies ..... teesetrencecnans
REACTOR CAPABILITY INVESTIGATIONS — APPROACH TO FULL POWER .....
5.1 OBTECTIVES iievcececcencanens Cevessesrtessesansesancsncscons
5.2 PROCEDURES sveeeeeoreencnacsens cecessaceersanns tevarecnanne
5.2.1  Performance of Control Systems ..eeeececveersconses
5.2.2 ohielding and Containment Adequacy +veeeeesececeesn
5.2.3 Calibration of Power Instruments ...ceeceveocs ceees
5.2.4  Xenon PoiSoning .cee.eevesececececceens Ceeerecreeeas
5.2.5 On-Line Analysis of Operation ...evececcoccasvasces
5.2.6 Thermal Effects of Power Operation .......eeeee... .

5.2.7 Capability and Performance of Heat Transfer
SYSTEMS teeveeereesconssorossonvsnncans ceceescononce

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ix

 

| 5.2.8 Chemical Effects of Power Operationecesvececsscesss 5-
) 5.0.9 Dynamics Studies «.... e, e 5~
6. SYSTEM CAPABILITY TINVESTIGATIONS — EXTENDED OPERATION e+ -sessesoe O-
6.1 OBJECTIVES .....vveeoevonsavaconnonconns e tseeeieeaan 6~
6.2 PROCEDURES «eevevoscsseancsscasssasonns feseersieratieanenss 6~
£.2.1 Fuel Chemistryseseesecorecenaesas cereeienaestansans D=
6.2.2 Materials Compatability......... ciecenens crecnaenes 6-
6.2.3 Changes in DynamicS....ceeeeesnoes Ceeeos teeescsenes 6~
6.2.4 Performance of Components and Equipment...... cevees O-
-

Vi RN v O N0
SECTION 1
INTRODUCTION

1.1 OBJECTIVES

The purpose of the Molten Salt Reactor Experiment, stated broadly,
is to demonstrate that many of the desirable features of molten salt re-
actors can presently be embodied in a practical reactor which can be
operated safely and reliably, and can be serviced without undue difficulty.
The program which has been lald out for the MSRE is intended to provide
that demonstration in a safe, efficient and conclusive manner.

Although the complete success of the MSRE depends in part on the
reactor being able to operate for long periods at full power, the test
program recognizes that the success of a reactor experiment is not
measured solely in megawatt-days. The tests and the experiments are
designed to be penetrating and thorough, so that when the experiment is
concluded not only will reliable operation and reasonable maintenance
have been demonstrated, but. there will be as many conclusive answers as
possible to the important Questions pertaining to the practicability of

molten =zalt reactors of this general type.

1.2 AREAS TO BE INVESTIGATED

1.2.1 Chemistry and Materials

 

Some of the most important questions have to do with the behavior
and interactions of the fuel salt, the graphite, and the container
material in the reactor enviromment. The major points to be investigated
in this area are:

1. fuel stability,

2. penetration of the graphite by the fuel salt,

3. graphite damage,

4.  xenon retention and removal,

5. corrosion,

6. behavior of corrosion products and non-volatile fission products.

The principal means of investigation used during operation will be

regular sampling and chemical analysis of the fuel salt, analysis of the
1-2

long-term regctivity behavior, and determination of the isotoplc compo-
sition of the xenon in the offgas. Periodically, during shutdowns, speci-
mens of graphite and of INOR will be removed from the core for examination.
l.2.2 Engineering

The MSRE incorporates some novel features and components which have
been developed and designed specifically for molten salt reactors. The
test program will obtain performance data on these items, permitting
evaluation of ideas and principles which could be employed in future
reactors.

The broad heading of Engineering also covers the extensive start=up
program which must be devoted to the checkout, calibration and preliminary
testing of the many more or less counventional devices and systems in the
MSRE.

1.2.3 Reactor Physics

 

From the standpcint of reactor physics, the MSRE core 1s unique.
But the nuclear design posed no serious problems. One reason for this
apparent paradox is the simplicity of the core, which makes simple spatial -
approximations valid. Another is that the demands for accuracy in the .
predictions are not severe. This follows because the fuel is fluid, -
permitting easy adjustment of the uranium loading and eliminating hot
spot problems assocliated with heat transfer from fuel to core coolant.
For these reasons the design did not employ extremely sophisticated
calculational procedures and there were no preliminary critical experi-
ments. Instead, the physics part of the reactor test program is relied
on to provide such accurate information on nuclear characteristics as
may be required.

The program of reactor physics measurements begins with the experi-
mental loading of uranium to attain criticality. Following this will be
experiments to verify that the system is stable and safe. Accurate
measurements of rod worth and reactivity coefficients will be made to
facilitate later analysis of the reagctivity behavior during powver
operation. This analysis will be concerned, among other things, with the

ransient behavior of *35Xe. The reactivity behavior will also be
scrutinized for possible anomalies, which might indicate changes in

conditions within the core. -
1-3

1.3 PHASES OF PROGRAM

The testing and experimental operation of the MSRE fall naturally

into different phases which must follow in sequence. They are:

1. precritical testing,

Do initial critical measurements,

3. low-power measurements,

L, reactor capability investigations.

The precritical testing phase begins with the new operators, as
part of their training, checking the location of equipment and comparing
the installed piping against the flowsheets. As systems are completed,
leak testing, purging, filling, calibrating and test operation are
started. The precritical testing culminates in shakedown operation of the
entire reactor system, with flush salt in the fuel gystem and coolant
salt in the coolant system.

In the initial critical experiments, fuel salt will be loaded and
enriched uranium will be added in increments to bring the concentration
up to that required for operation. During this phase the control rods
will be calibrated and fuel concentration, temperature and pressure
coefficients of reactivity will be measured. Baseline data on the
fuel chemistry and corrosion will also be obtained during this period.

Following the critical experiments, which will be conducted at a
few watts of nuclear power, the power will be raised to permit cexrtain
tests. These will include tests of the nuclear power servo control
gystem, the automatic load control system, the calibration of power
indicators and surveys of the bioclogical shielding. The nuclear power
will be less than 2 Mw during this period.

Capability investigations consist of two parts: The first, a step-
wise approsch to full power; the second, extended operation. During the
approach to full power, temperatures, radiation levels and the nuclear
power noise will be observed to determine if any unforeseen condition
existe which would restrict the attainable power level. Extended opera-
tion will test the reliability of equipment and long-term corrosion am
fission-product behavior. Maintenance will be carried out as required

and the reactor will be shut down periodically for removal of samples
1-4

from the core. Long-range-plans include chemical processing of the fuel

salt and operation with different fuel sait compositions.

1.4 DOCUMENTATION

This report describes, in rather general terms, the experiments to

be performed with some discussion of the methods to be used and the type

of results expected. It provides the basic plan for the day-to-day
experimental program.

Bach experiment or test, prior to its performance, is the subject
of a Test Memo which describes that particular experiment in complete
detail. A stepwilise procedure with references to applicable established
operating procedures is included. If they are required, supplementary
check lists and sampie data sheets are made a part of the Test Memo.
The Test Memos must be internally reviewed and approved before the
experiment is performed. Since the procedural details are of limited
interes®:, these documents are distributed only to personnel and super-
vision directly connected with the experiment.

As soon as possible after the completion of an experiment, a Test
Report is written to describe the results. This report summarizes the

test experience and data obtained and presents any conclusions that can

be drawn. The scope and importance of the individual experiments deter-

mine the nature and distribution of the Test Reports.
SECTION 2
NONNUCLEAR TESTING

 

2.1 OBJECTIVES

Prior to full-scale operation the MSRE will undergo a number of
shakedown runs and tests. The purposes are:

1. to assure that the design is adequate and that the equipment and
instrumentation function as designed;

2. to cbtain information which may be needed for future operation or
analysis of the reactor (calibration of instruments and equipment,
dimensional changes, etc.);

3. to discover and correct weak points of the system to assure that
it is safe and operable (This includes long term integrated runs
to allow for early failure of defective equipment.);

L, to develop sampling techniques and determine the adequacy of analysis
procedures;

5. to train operating personnel and check out the operating procedures;

6. to determine the effects of equipment or instrument failures or

maloperation.
2.2 EXPLANATION OF PROCEDURES

Most of the nonnuclear tests will be performed before the reactor
is made critical for the first time. However, the testing of some
equipment will be completed at a later date. For example, the vapor-
condensing system and the final closure of the contaimment vessel wilil
not be completed before the critical experiment sc the testing of these
items will be performed after the zero-power nuclear tests. Thus, the
order in which the tests are listed does not indicate the chronological
order of testing.

Because of the large number of nomwuclear tests, a numbering system
wags adopted to facilitate the maintenance of records. The various test
memos and operating procedures that are applicable are referred to by

number in the descriptions whica follow.
a-2

2.3 PROCEDURES

z.3.1 Fuel System
The fuel system consists of the reactor, the fuel pump, the overflow

tank, the heat exchanger, and assoclated piping.

Prior to and during early heatups, the fuel system and graphite will
be purged of moisture. The necessary heater settings for various average
system temperatures will be determined along with temperature gradients,
cool down rates, and adequacy of spring piping supports.

Purge — Oxygen and moisture must be removed from the system prior
to heatup or addition of salt to the system. This will be done by evacu-
ating and refilling the system with helium. The system will be evacuated
through a temporary connection to line 110 in the drain-tank cell. Since
there will be no salt in the freeze valves at this time, the entire drain-
tank system is purged also. Since a large quantity of moisture is
expected to be released from the graphite during heatup, the moisture
content of the helium will be monitored and further evacuations performed
during heatup 1f necessary. Although evacuation will be from line 110
at the drain-tank cell, the venting of purge helium will be through the
offgas system charcoal beds. Details are given in Test Memo XI 2.3.1.1-A.

Heater Settings — During early heatups, all the reactor-cell piping

 

will be heated concurrently. At least that portion of the coolant-salt
system in the reactor cell must be heated concurrently due to thermal
expansion of piping. Thermocouples will be monitored and excessive
thermal gradients, such as might occur at the cell penetrations, will
be minimized by proper heater adjustment. Heater-control settings will
be determined for holding the system at various temperatures. The rate
of cooldown and temperature gradients during a simulated power outage
will be checked. Details are given in Test Memo XI 2.3.1.1-B.

Thermal Stress in Piping and Equipment —— The thermal growth of the

 

piping system and the operation of the fuel-pump and piping supports
will be noted during hesgtup. The piping heaters will be observed while

at operating temperature to detect any apparent difficulties due to the
c-3

expansion of the piping system. Temperature gradients at points of
stress will be analyzed and strain gages will be used if necessary.
Details are given in Test Memo XI 2.3.1.1-C.

Z2.3.1.2 Initial Fill and Operation

During the initial fill, the normal fill procedure, Section 5I,
Part VIII, Operating Proceduresf will be used and no calibration data,
as such, will be taken. Salt will be circulated and sampled for a
period to gain both operating experience and continuous-operation sample
data. The system will then be drained and refilled during which the
system will be calibrated vs the amount of salt added, rate of fill
determined, overflow tank calibrated, cooling tests performed and drain
time established.

Calibration — During the calibration fill, the fuel-system level
and volume will be calibrated vs the amount of salt added as the drain-
tank pressure is increased by increments. After each partial addition
the approximate level in the system and weight of salt in the drain
tank will be determined. From previously obtained drain-tank calibrations
the weight of salt in the fuel system vs elevation will be plotted. The
fuel pump will be overfilled to determine the location of the overilow
inlet. Some salt will be transferred to the overflow tank to test the
level indicators. Detalls are given in Test Memo XI 2.3.1.2-A.

Fuel-Pump Tests — The tests to be performed on the fuel pump in-

 

clude checking operation of the bubble-tube level indicators and deter-
mination of load and no-load power requirements of the fuel-pump motor.
Details are also given in Test Memo XI Z2.3.1.2-A.

Cooling Rates -— Cooling rates for both the coolant and fuel systems
willl be determined from a starting condition of lZOOOF to a minimum of
lOOOOF. Recorder charts and photographs of scanner traces will be used
to determine cold spots and cooling rates. Details are given in Test

Memo XTI 2.3.1.2-C.

 

*R. H. Guymon, MSRE Design and Operations Report, Part VIII, Operating
Procedures, USAEC Report ORNL-TM-908, Oak Ridge National Laboratory,
November, 1965.
2-4

Drain Times — Drgin times will be determined at both minimum and
maximum circumstances — the slowest drain time being that with only the

drain-tank vent open. Details are given in Test Memo XI 2.3.1.2-A.

Initial Operation — The preliminary fill will be done according to

 

a rormal fill procedure which includes the freezing of a salt plug in
the reactor access nozzle. The system will be filled, the freeze valve
frozen, and salt circulation begun. Circulation and normal operating
conditions will be established for a period of days during which salt
samples for chemical. analysis will be withdrawn from the fuel pump
through a temporary sampler. Details of the preliminary fill are given
in Section 5%, Part VIII, Operating Procedures.

2.3.1.3 Krypton Injection

Calculation of a reactivity balance at freguent intervals provides
a valuable indication of conditions in the core during nuclear operation.
Whenever the power is significant (sbove a few kilowatts), the poisoning
of *35Xe is an important term in the reactivity balance. Constants which
are used in the computation of the *>°Xe poisoning must therefore be
available at the outset of nuclear operation.

The 13°Xe poisoning depends strongly on the amount of gas bubbles
circulating with the salt, the stripping in the pump bowl and the mass
transfer of xenon between the salt and the graphite in the core. The
effect of these mechanisms can be predicted from theory, basic data
and pump-loop experiments, but the uncertainties are undesirably large.
Therefore, during the precritical testing an experiment will be done
with radioactive krypton to provide further information on the behavior
of noble gases in the MSRE.

Radioactive B%Kr (10.4 y half-life) will be injected with the
helium flow into the fuel pump while flush salt is being circulated.
The offgas is diverted, just after leaving the pump bowl, past a radi-
ation measuring device. Sampling connections for trapping krypton on
charcoal are also provided at this point. Normally the flow bypasses
the sampling bombs, back into the offgas line, through the charcoal
beds and up the ventilasion stack.

Basically the experiment consists of saturating the sall and graphite

with krypton, then stopping the inflow and obgerving the rate at which
a-5

the krypton is eliminated from the system.

The operation will be done in three phases. The first will be a
short run (about 10 hr) to calibrate the equipment. The second run, of
about 2 days duration, will yield approximate values for the constants
which are to be measured. The results of these two runs will be used
to decide on the rate of krypton injection and the duration of a third,
long run. The anticipated rate of injection is 6 curies/day or less
and the duration, chosen to allow 35Kr concentrations to reach steady
state, 1is expected to be about 20 days.

At the end of the prescribed period, the input of 85Kr will be
stopped and the decreasing ®5Kr concentration in the offgas will be
followed closely until the level becomes insignificant. The concentration
will decrease rapidly at first as most of the krypton in the salt is
stripped, then more slowly as krypton diffuses out of the graphite into
the salt and thence into the offgas stream.

The transients in the ®%Kr concentration in the offgas stream will
be analyzed to yield values for fuel salt-gas holdup, stripping rate
and the rate of transfer from the graphite to the salt. These quanti-
ties can then be used rather directly to predict values for xenon.

These values will be incorporated in the 12 Xe reactivity calculation
for the nuclear startup.
2.3.2 Fuel-Drain-Tank System

The fuel-drain-tank system consists of fuel drain tank No. 1 (¥FD-1),
fuel drain tank No. 2 (FD-2), fuel flush tank (FFT), and associated
piping. The two drain-tank afterheat-removal systems are also included.
Precritical testing will include the following items.
2.3.2.1 Calibration of Steam Drums

The steam drums will be calibrated by adding known increments of
water from the previously calibrated feed-water tanks and by comparing
the amount added with the indicated level. The calibration curves thus
obtained will be used to set the operating parameters. Details are given
in Test Memo X1 2.3.2.2.
2.3.2.2 Initial Heatup

During the initial heatup the system will be purged and stress
relieved, the thermal growth of the piping will be checked, the heatup
2-6

and cooldown rates wili be determined, and the necessary heater settings
for various operating conditions will be obtained.

Purge — The system will be purged at the same time as the fuel
system. Details are given in Test Memo XI 2.3.1.1-A.

Heater Setftings — The heater-control settings will be determined

 

for holding the system at various temperatures. During cooldown, the
effect of loss of electrical power will be checked. From this information,
operating conditions can be established. Mechanical limits will be put
on heater controllers to prevent overheating during future operation.
Possible temperature effects on the weigh-cell readings will be noted.
Details are given in Test Memo XI 2.3.2.2-B.

Thermal -Expansion Effects — Prior to heating the system, key

 

measurements will be taken on the piping and equipment. Stress relieving
will be accomplished by holding the temperatures at 13OOOF for a minimum
of 100 kours. While hot and after cooling down, the key measurements
will be rechecked to determine movement which could cause trouble in the
future. Details are given in Test Memo XI 2.3.2.2-C.
2.3.2.3 Initial Salt Fill

A small amount of flush salt will be added to the system and will
be used to fill the freeze valves. The remainder of the flush salt will
then be added to fuel drain tank No. 1. Weligh cells will be calibrated
when the tanks are cold by adding known increments of weight. They will
be recalibrated as the salt is added. The weigh-cell readings when the
level probes are actuated will be noted. Using the probe locations as
baselines, the elevations of salt in the tank will be plotted ve weight
of salt added and weigh-cell readings. The fuel drain tank No. 2 and
the fuel flush tank will be calibrated by transferring the salt in
increments and by observing the tank weights and level probes. From
these curves, the weight and elevation of the salt in the fuel system
can be determined during fill operations. Details are given in Test
Memo XI Z2.3.2.3.

In ordar to determine how soon salt will freeze in a drain tank
if electric power is lost, the power will be turned off all heaters and

the salt will be allowed to cool approximately ZOOOF. The curve will be
2-7

extrapolated to determine the freezing time. Detalls are given in Test
Memo XI 2.3.2.k4.

Sufficient cooling capaclty is needed to remove fission-product
afterheat from the fuel when it is drained to the drain tanks. With
flush salt in a drain tank, the steam drums will be put into service
and the heat removal rate determined by the cocldown rate of the salt.
The test will be terminated before the salt freezes. Details are given
in Test Memo X1 Z2.3.2.5.

2.3.3 Coolant System

The coolant system consists of the radiator, coolant pump, heat
exchanger, and associated piping. Precritical testing will consist of
the following items.
2.3.3.1L Initial Heatup

During the initial heatup the system will be purged and stress
relieved. The thermal growth of the piping will be checked, the heatup
and cooldown rates, as well as temperature gradients, will be determined,
and the necessary heater settings for various operating conditions will
be obtained.

Purge — Before any salt-containing equipment is heated or salt
added, the entire system will be purged to remove oxygen and moisture.
This will be done by first evacuating and filling the system with helium
followed by an extended purge which will continue through the heafup.
Purging will be conducted in a sequence to insure purging of all gas and
salt piping. The coolant pump will be operated to circulate helium
through the main loop. 8Since there will be no salt in the freeze valves,
the coolant drain tanks will be purged along with the coolant system.
Purging of the coolant oil system is also included at this time. IF
possible, the fuel system, fuel-drain-tank system, cover-gas system,
and offgas system will be purged at the same time. Details are given
in Test Memo XI 2.3.3.1-A.

Heater Sebtings -— During the initial heatup all thermocouples will

 

be monitored to assure that no cold spots or excessive temperature
gradients exist. Necessary adjustments of the heater controllers will

be made using, as a gulde, previously prepared graphs of the indicated
2-8

heater current vs the estimated power per square foot of surface. Heater-
control settings will be determined for holding the system at various
temperavures. The rate of cooldown and temperature gradients during a
power outage will also be checked. From this information mechanical
1imite can be put on the controllers, curves can be made to show the
interrelationship between heater current and equipment temperature, and
the normal control settings can be established. Details are given in
Test Memo XI 2.3.3.1-B.

Thermal Growth — Stress relief of individuval components and piping
welds will be performed during assembly. However, as part of the initial
heatup the entire system will be held at l3OOOF for 100 hours for addi-
tional stress relief. The thermal growth of the piping system and the
operation of the constant-load pipe supports will be noted before,
during and after the initial heatup. The piping heaters inside the
reactor cell will be observed while at operating temperature to detect
any apparent difficulties due to the shifting of the piping system.
Details are given in Test Memo XI 2.3.3.1-C.

During the initial fill the level in the system will be calibrated
vs the amount of salli added. The rate of fill will be determined,
various coolant-pump tests will be made, and the rate of cooling will be
checked under various conditions. The effect of temperature on the
coolant-salt filow meter will be investigated.

Level Calibration — The initial fill of the coolant system will

 

be made by increasing the drain-tank pressure in increments. After

each partial addition, the level of the salt as indicated by the AP,

the loop pipe temperatures, and the weight of the salt in the coolant
drain tank will be determined. From this information and the calibration
of the coolant drain tank vs the elevation made previously, the weight

in the coolant system vs the elevation will be plotted. In order to
establish future operating parameters, the rate of fill will be deter-
mined at various settings of the drain-tank helium-supply valves and

at various salt elevations. The salt level in the pump bowl will be
calibrated using the float indicator and both bubblers. Details are

given in West Memo XI Z2.3.3.2-A.
2-9

Salt Flowmeter — The effects of loop temperatures and temperatures
of the NaK-filled differential-pressure cells on the flow indicated by
the coolant-salt flowmeter will be determined for baseline information.
Details are given in Test Memo XI 2.3.3.2-B.

Cooling Rates — The cooldown rate upon loss of electric power will
be determined with salt circulating in the system and without salt circu-
lation. The response time needed to stop the system cooldown and start
heating will be checked. Tests will also be made to determine the
temperature response with and without circulation using only the emer-
gency electric-power supply. From this information, the operating
policies during power outages can be finalized. Details are given in .
Test Memo XI 2.3.3.2-C.

Freeze-Valve Thaw Rate — The rate at which the drain valves

 

(FV 204 and 206) will thaw with and without electric power and the rate
of drain of the salt from the coolant system will be determined under
various operating conditions. From this information operating procedures
can be established which will prevent freezing of the salt in the radi-
ator. Inventory checks will be made after each drain to determine the
heel left in the coolant system. Details are given in Test Memo
XI 2.3.3.2-D.

The radiator doors, blowers, and dampers will be checked to assure
that they operate as designed and that the conitrol circuits function
properly. The stack flowmeter will be calibrated.

Radiator Door Tests — The operation of the radiator doors will be

 

tested. Most of these tests will be made with the coolant system at
ambient temperature. They include: rate of raise, rate of lowering
under power, and during a load scram, and position change with blowers
on. The operation of the doors will also be checked while at 1200°F.
Warpage will be determined after the heating and cooling cycles.
Details are given in Test Memo XI 2.3.3.3-A.

Radiator Cooling — Tests will be made to determine the cooling

 

rate which would occur if both radiator doors were dropped with salt

in the system. Details of this are given in Test Memo XI 2.3.3.3-B.
2-10

Damper Positior — The measured damper position ve indicated

 

position will be measured and the reproducibiliity checked before heatup.
Several points will be rechecked after heatup and cooldown. The rate
of movement of the dampers will be determined. The operation of the
dampers will be checked whiie the system is at lZOOOF. Details are
given in Test Memo XI 2.3.3.3-C.

tack Flow Rates — The stack air-flow instrument will be calibrated

 

over the range of 20,000 to 200,000 scfm by measuring veiocity profiles.
Flow rates at various combinations of blowers, door positions, and
damper positions will be determined for future reference. Details are
given in Test Memo XI 2.3.3.3-D.

Radiator Air Leakage — Ventilation is provided to maintain the

 

coolant cell at a lower pressure than the high-bay area. Tests will be
made at various door and damper positions with both radiator blowers

in operation to assure that leakage does not pressurize the cell. Details
are given in Test Memo XI 2.3.3.3-E.

2.3.4 Coolant Drain-Tank System

 

The coolant drain-tank system consists of the coolant drain tank
and associated piping. Precritical testing will consist of the following
items.
2.3.4.1 Initial Heatup

During the initial heatup the system will be purged and stress
relieved, the thermal growth of the piping will be checked, the heatup
and cooldown rates will be determined, and the necessary heater settings
for various operating conditions will be obtained.

Purge — The system will be purged at the same time as the coolant
system. Details are given in Test Memo XL Z.3.3.1-A.

Heater Settings — The heater control settings will be determined

 

for holding the system at various temperatures. Attempts will be made
to hold all temperaturss within % 100°F of each other. During cooldown,
the effect of loss of electric power will be checked. From this
information coperating conditions can be established. Mechanical limits
will be put on heater controliers to prevent overheating during future
operation. Possible tempersture effects on the weigh-cell readings

will be noted. Details are given in Test Memo XI 2.3.4.2-B.
Z=-11

Thermal Growth -— Prior to heating the system, key measurements
will be taken on the piping and equipment. Stress relieving will be
accomplished by holding the temperatures at l300oF for a minimum of
100 hours. While hot and after cooling down, the key measurements will
be rechecked to determine movement which could cause trouble in the
future. Details are given in Test Memo XI 2.3.4.2-C.
2.3.4.2 Initial Salt Fill and Calibration

A small amount of coolant salt will be added to the system and will
be used to fill the freeze valves. The remainder of the salt will then
be added to coolant drain tank. The weigh cells will be calibrated when
the tank is cold by adding known increments of weight. They will be
recalibrated vs the welght of salt added. The weigh cell-readings when
the level probes are actuated will be noted. Using the probe locations
as baselines, the elevations of salt in the tank will be plotted vs
weight of salt added and weigh-cell readings. From these curves, the
weight and elevation of the salt in the coolant system can be determined
during fill operations. Details are given in Test Memo XI 2.3.L.3.

In order to determine how soon salt will freeze 1in the drain tank
if electric power is lost, the power will be turned off all heaters and
the salt will be allowed to cool approximately 200°F. The curve will
be extrapolated to determine the freezing time. Detalls are given in
Test Memo XTI 2.3.k.k.

2.3.5 Cover-Gas and Offgas Systems

The cover-gas system provides a helium supply to purge the salt
systems, to transfer salt by pressurization, and to provide an inert
atmosphere. The offgas system provides holdup for the fission gasses
from the fuel system and fuel-drain-tank system. It also includes vent
lines from the coolant system, coolant-drain-tank system, and lube-o0il
system to the stack.

Before purging the fuel and coolant systems, the cover-gas system
will be purged of air by evacuating and then pressurizing with cylinder
helium. The purge will be continued with cylinder helium during the

initial purge of fuel and coolant systems and then with treated helium
2-12

during the final purge of the fuel and coolant systems. The offgas system
is purged when the systems which it vents are purged. The detailed
procedure for evacuating and pressurizing the cover-gas system is given in
Test Memo XI 2.3.5.1.

2.3.5.2 Charcoal-Bed Retention Time

The helium purge from the fuel pump flows through the main charcoal
beds which are designed to hold up the assoclated fission-product gases
long enough tc allow them to decay sufficiently so that they can be dis-
charged safely from the stack to the atmosphere. The regquired holdup for
Kr at 10 Mw operation is 7 1/2 days at design carrier-gas flow rates.

To determine that the beds meet the design requirements, a burst of
85Ky will be charged into the bed inlet and the discharge stream will be
monitored for activity to determine the holdup time in the bed. Various
helium carrier-gas flow rates will be used during the test. Details are
given in Test Memo XI 2.3.5.2.
2.3.5.3 Charcoal-Bed Pressure Drop

The pressure drop across the main charcoal beds will be measured
over the expected flow range using installed instrumentation. This will
check the design calculations and assist in determining the bed condition
during reactor operations. Details of the test are given in Test Memo
XI 2.3.5.3.

Exact prototype units of both the oxygen-remover and dryer were
used in the development program for the cover-gas system. The performance
of the installed units will be determined by analyzing the inlet and out-
let helium for oxygen and moisture during system purging and precritical
testing. WNo loading tests are planned at the reactor site for either the
oxygen-removal unit or the dryer.

2.3.6 0il Systems

The two oil systems are auxiliaries of the fuel and cooclant salt
pumps. They provide both lubricating oil, for bearings and seals, and
cooling oil for the pump radiation shields. The test program will consist
of various tests and calibrations necessary for proper operation. Both

systems will be given a general shakedown before startup of the salt pumps.
2-13

2.3.6.1 Final Leak Test

The oil systems are parts of the secondary containment and will be
helium leak tested in conjunction with the fuel and coolant systems. For
details see Test Memo XTI 2.3.1.3.
2.3.6.2 (Calibration of Supply and 0il Catch Tenks

During normal operation there will be a small amount of oil leakage
through the salt pump seal. This is collected and measured in the oil
catch tank. A serious oil leak here or elsewhere in the system will show
up as a decrease in oil level in the supply tank. Therefore both the
supply-tank and catch~tank level indicators will be calibrated prior to
operation. Details are given in Test Memo XI 2.3.6.2
2.3.6.3 Emergency Supply |

It is important that the systems supplying oil to the salt pumps
e reliable. Details are given in the operating procedures for startup
of standby pumps under various operating conditions, operating with one
0il pump supplying both salt pumps, and adding oil during operation with-
out violating containment. The adequacy of the design will be tested by
simulating abnormal conditions and by operating the systems as detailed
in Section 9H, Part VIII, Operating Procedures.
2.3.6.4 Heat-Exchanger Test

The cooling systems on the oil supply tanks are designed with suf-
ficient capacity for the operation of both salt pumps at full reactor
power with one oil system. Fouling of the surfaces could reduce the heat
transfer capabiliiy below the design value. Therefore, periodic checks
will be made to detect changes in the overall heat transfer coefficient.
Shortly after startup of the system, tests will be made at several water
and o0il flows and heat-removal rates to obtain base information. Details
are given in Test Memo XI 2.3.6.L.
2.3.7 Chemical Processing System

The chemical processing system consists of the fuel storage tank,
transfer lines for moving salt from the drain-tank system and a gas
sparging system for removal of oxygen or uranium from the salt.

Equipment Preparation — The entire chemical processing system will

 

be helium leak checked during construction and component fabrication.
2-1h

After leak testing, the salt-handling part of the system will be purged
with nitrogen gas, to remove all moisture and oxygen. When the moisture
and O, have been removed, the system will be heated to 1200°F and pressure
tested.

Tank Calibrations — Prior to operation, the caustic neutralizer and

 

caustic-addition tanks will be calibrated using water. The fuel-storage-
tank weighing system and ultrasonic probe will be calibrated during the
initial salt addition from the fuel flush tank by comparing the weight

of salt in the two tanks.

O- Removal Calibration — Two methods of determining the amount of
oxygen removed from the salt as water vapor during H.-HF processing con-
sist of an electrolytic hygrometer to measure moisture in the offgas and
a syphon pot to measure increments of condensed water vapor. These will
be calibrated and compared during & test using a mixture of water vapor
and nitrogen.

Flush-balt Treatment — The final system test will consist of re-

 

moving the oxygen from the flush salt which is used during the precritical
operation of the fuel system. This test will determine the system
efficiency and indicate any modifications which are needed before the
system is contaminated by processing fuel salt.
2.3.8 ILeak-Detector System

This system consists of eight headers, each of which is connected
to a common helium supply on one end and by means of stainless steel
tubing to the ring grooves of in-cell flanges on the other end. Normally
the leak~detector system will be at a higher pressure than the systems
being monitored. The system will detect flange leaks, by a drop in system
pressure, and prevent outleakage by maintaining a buffer of helium. Pre-
critical testing will consist of the following items.
2.3.8.1 Calibration

In order to convert the pressure changes to volumetric leak rates,
it is necessary to know the volume of various segments of the system.
Fach header, leak-detector line, ete., will be calibrated by equalizing
pressure with a bomb of known volume. The volumes of the system will
be calculated from changes in pressure. For calibration details, see

Test Memo XI 2.3.8.1.
2-15

2.3.8.2 Purging

Before putting the leak detector system into operation, the headers
and lines will be purged of oxygen. The headers will be purged by
pressurizing to 100 psig with helium and venting three times. The lines
will be purged by bleeding gas through the leak detector lines to the
flanges before the flanges are sealed.
2.3.9 Cooling Water Systems

Potable water is supplied to the MSRE from the X-10 area. After
passing through a backflow preventer, it is called process water and is
used as makeup for the cooling tower and for other out-of-cell cooling and
process uses. Cooling-tower water which is circulated centrifugal pumps
provides cooling for the treated-water cooler and other out~of-cell
equipment. All in-cell cooling is done by treated water which is circu-
lated by centrifugal pumps. Precritical testing will consist of the
following items.
2.3.9.1 Potable-Water System

More than adequate supply is available to the area; however, this
will be verified during reactor startup.
2.3.9.2 Procegs-Water System

The backflow preventers will be tested in accordance with ORNL
Standard Practice Procedure No. 1h4. For details see Section U4C, Part VIIT,

Operating Procedures.

The cooling-tower system will be filled and operated to assure the
adequacy and reliability of the components. Details are given in Test
Memo XT 2.3.9.3.
2.3.9.4 [Treated-Water System

The treated-water system will be filled with condensate and all
equipment operated when the heat transfer coefficient of the treated-
water cooler is measured. Also, the condensate makeup rate will be
measured and calibrations of the system and tank volumes will be made.

Condensate Makeup — The condensate makeup rate will be determined

 

by measuring the time to produce a measured amount of condensate from the

normal steam supply. Condensate will be used to calibrate the various
2-16

tanks by adding or removing measured amounts of condensate. These
calibrations will be used to determine the rate of water usage. Details
of these tests are given in Test Memo XI 2.3.9.L4-A.

Volume Calibration — The system volume calibration will be made by

 

comparing the water analysis before and after addition of known amounts
of a corrosion inhibitor. The calibrations will be used in calculating
chemical additions required to maintain the proper water treatment.
Details are given in Test Memo XTI 2.3.9.L-B.

Treated-Water Cooler — The heat transfer coefficient of the treated

 

water cooler will be determined to provide a baseline figure which can be
used if tube fouling is suspected. Details are given in Test Memo
XT 2.3.9.k-C,

Effect of Water Flow on Cell Temperature and Pressure — The effect
of closing the radiation block valves on cell temperature and pressure
will be tested. After the cells are closed and sealed, the water to the
space coolers will be shut off and the rate of pressure and temperature
increase will be measured.

When the surge-tank vent valve closes on activity in the treated
water, the system becomes an unvented system. The effect on water makeup
and circulation will be tested with the vent closed. Details of testing
the block valve effects are given in Test Memo XI 2.3.9.L4-D.

Thermal Shield — The permissible pressure in the reactor thermal
shield is quite low. During the initial startup of the treated-water
system, tests will be conducted to insure that the thermal shield is
adequately protected against excessive pressures.

2.3.10 Component-Cooling Systems

The primary component-cooling-air system consists of a circulating
system in which gas from the reactor and drain-tank cells is compressed,
cooled, and reused to cool in-cell components (freeze valves, pump bowl,
control rods, reactor neck, and graphite-sampler neck). A secondary
cooling-air system supplies air for cooling the freeze valves located
outside the reactor and drain-tank cells. Precritical testing will con-

sist of the following items.
=17

2.3.10.1 ILeak Testing

The section of the primary component-cooling system which is outside
the main cells is part of the reactor containment enclosure. All flanged
Joints in this section will be soap checked for leaks and will be re-
checked during the containment leak test. Each component will be hydro-
statically tested during manufacture, and the entire system will be
pneumatically tested after installation. Details are given in Test Memo
XT 2.3.10.1.
2.3.10.2 Heat_ Transfer Coefficient of Heat Exchanger

A heat balance will be made on the heat exchanger, and the heat
transfer coefficient will be calculated. This will be used as a basis
for subsequent evaluation of the heat-exchanger performance. Details are
given in Test Memo XTI 2.3.10.2.

The effect of the cooling-air flow rate to the fuel-pump bowl on
the temperature distribution will be determined. This will allow selec-
tion of the optimum flow to minimize thermal stresses on the pump bowl.
Details are given in Test Memo XI 2.3.10.3.
2.3.10.4 Flow Adjustment

The flow to the freeze valves and other equipment served by both the
primary and secondary component-cooling-air systems will be set to give
the desired freeze valve operating characteristic and equipment cooling.
Details are given in Section LI, Part VIII, Operating Procedures.
2.3.10.5 Flow Stability

The stability of air flows will be checked by observing the system
pressure and monitored thermocouples on the equipment during periods when
air flow is being changed, such as during operation of freeze valves, or
the evacuation of the cell. Details are given in Test Memo XI 2.3.10.5.
2.3.11 Instrument-Air and Auxiliary-Air Systems

Clean, dry, compressed air is supplied to the MSRE instruments by
a reciprocating compressor and heatless ailr dryer with a spare compressor
and dryer in a standby. Cylinders of nitrogen provide emergency gas pres-

sure to the more important instruments. Auxiliary compressed air is
2-18

supplied by a third reciprocating air compressor for the operation of
pneumatic tools and other plant uses. Precritical testing will consist of
the following items.

In order to establish a base for comparing the future performance of
the compressors, the capacity of each instrument-alr compressor will be
checked while new. Tlow rates as indicated by the installed instruments
will be recorded with the system operating at steady~state design con-
ditions. The relative loading and unloading times will also be determined
at various flow rates. Details are given in Test Memo XTI 2.3.11.1.

To ensure proper operation of the pneumatic instruments, the dryers
must reduce the moisture content of the air to a dewpoint of —20°F. The
moisture content of the discharge air from each dryer will be monitored
with installed instruments. At design operating conditions the automatic
timing cycle of each dryer will also be checked. Details are given in
Test Memo XTI 2.3.11.2.
2.3.11.3 FEmergency Supply

oufficient emergency instrument-air capacity should be available to
assure an orderly shutdown of the reactor in case of loss of both instru-
ment-air compressors. Both compressors will be stopped and the rate of
pressure drop will be determined. Times will be noted when various
annunciations or instrument failures occur. Comparison of these with
the time necessary to drain the system should indicate whether the
emergency supply is adequate., Detailed procedures are given in Test Memo
XT 2.3.11.3.

£.3.12 Instrumentation

 

All instruments will be thoroughly checked prior to operation to
ensure that they function properly.

All instruments will be calibrated following fabrication. Each will
be checked to ensure that the transmitted signal covers the proper range.
The system recorders and indicators will be checked with a precision instru-

ment to ensure the proper value is read out,.
2-19

2.3.12.2 Temperature

Fach thermocouple will be checked for continuity and resistance.
Temperature readout instruments will be calibrated by feeding in milli-
volt signals which correspond to the temperature range to be covered.

A1l control circuits and control loops will be checked following
installation for continuity and proper control action. These are covered
by the instrument startup check list, Section LH, Part VITI, Operating
Procedures.
2.3.12.4 Computer

The computer to be used for scanning, recording, and processing
reactor data will be checked out and tested independently of the reactor
system. However, & complete checkout of the computer will require, not
only operation of the reactor system, but the production of nuclear power.
Therefore, it may be expected that some corrections and modifications
of the computer system and/or programs will be required after the system
is nominally in service.

The pre-operational testing of the computer will be performed in
two stages. The first stage will take place at the vendor's plant prior
to shipping and the second will take place at the reactor site after
installation of the equipment. Both stages will be a Jjoint effort by
members of the ORNL staff and representatives of the computer manufacturer.
After completion of the second stage of testing, the computer system
will be placed in normal operation and made available to the reactor
operating staff. OCubstantial on-line operating experience will be
accumulated so that complete reliance can be placed on all aspects of the
computer operation.

The first phase of the checkout will begin with the normal checks,
by the manufacturer, of quality, workmanship, and operability of the
computer itself and all the associated subsystems and peripheral devices.
This will be followed by assembly and checkout of all programs to be used
in the computer. Operation of the system will then be tested with fixed
values for all input parameters; values that are expected to be typical

for both full-power and low-power operation of the reactor will be used.
2-20

In cases where a computational program may follow any of several paths
depending on the input value, a variety of values will be used to allow
all possibilities to be checked. This phase will be concluded with a
series of customer-acceptance tests to demonstrate that the computer meets
all requirements.

The second phase of the checkout will be conducted after the computer
system has been installed at the reactor site and all of the reactor-
system signals have been connected to it. This will permit complete
testing of all aspects of the computer operation except those associlated
with nuclear-power operation of the reactor.

The details of the procedures used in checking the computer are the
joint responsibility of ORNL and the computer manufacturer. Since the
initial checkout is not directly associated with the reactor system and
does not involve operating personnel, it will be described separately.

The periodic checks to be performed by operating personnel will be
publicshed as part of the operating procedure for the computer.
2.3.13 Electrical System

2.3.13.1 TVA Feeder Switchover

The 7503 Area is supplied by two 13.8-kv feeders, a preferred line

 

(ORNL Circuit 234) and an alternate (ORNL Circuit 294). Low voltage on
Circuit 234 will initiate an automatic transfer to Circuit 294 after a
1- to 10-sec delay, providing there is voltage on Circuit 294 and no
fault between the two motor-operated switches. Details are given in
Test Memo XTI 2.3.13.1 and in 3A, Part VIII, Operating Procedures.

Diesel Cenerators 3 and L4 supply 480-v AC current after the loss of
both TVA feeders for operating motorized process equipment, some lighting,
and some instrument power. Generator 5 supplies 480-v AC power for
operating process electric heaters. Tests will be run to determine the
preoperational settings to bring the units to power safely from the remote
start. The actual operating load of each diesel will be compared to the
tabulated load. Parallel operation of Generators 3 and 4 with TVA will
also be tested. Details are given in Test Memo XI 2.3.13.2
2-21

— o — m— e ——

The 48-v system is used to supply uninterrupted power for critiecal
instrumentation. This power is normally supplied from either of two
3-kw, AC-IC M. G. sets with a battery "floating'" on-line to furnish
emergency power after a loss of the normal electrical supply. A test will
be made to determine the actual time during which the batteries will
supply adequate power to the instruments. Details are given in Test Memo
XTI 2.3.13.3. The two M. G. sets must be operated in parallel to charge
the battery. This procedure (Section 3A, Part VIII, Operating Procedures)
will be tested.
2.3.13.4 250-v Supply

The 250-v DC system supplies power for DC emergency lights, breaker
trip power, feeder transfer power, and a 25-kw IC-AC M. G. set. This
system is normally supplied from a 125-kw AC-DC M. G. set, and has a
battery capable of supplying emergency power for two hours under full
load. Low voltage from the 25-kw M. G. set will throw an automatic
switch, transferring the power for Instrument Panels 2 and 3 to Generator L.
The effective 1life of the battery will be tested under various loads,
and the automatic transfer of the instrument power will be tested as
outlined in Test Memo XI 2.3.13.L.
2.3.13.5 Emergency Lights

Emergency DC lights come on automatically on loss of AC power to
Lighting Panel H. The DC lights will be tested by opening the Lighting
Supply PBreaker and a check will be made to assure that all areas are
adequately lighted.
2.3.14 Shield and Containment
2.3.14.1 Initial Cell Testing

The construction of the reactor cell and drain-tank cell are part of
a major building-modification contract. Upon completion of these cells,
and prior to acceptance by ORNL, the cells will be sealed and hydro-
statically tested to 48 psig and then pneumatically leak tested at 20 psig
and —5 psig. Details of the test are outlined in Test Memo XI 2.3.1Lk.1.

The reactor cell, drain-tank cell, and certain appendages comprise

the secondary containment of the reactor system. All piping entering or
2=22

leaving the containment is protected against out-leakage of activity

during an accident vy check valves or automatic block valves., These
containment check valves and block valves will be leak tested prior to
operation. The secondary containment of the reactor will be leak tested

at 20 psig, 10 psig, 5 psig, 2 psig and —2 psig. The data will be
extrapolated to the expected leakage for the maximum credible accident and
must not exceed 1% of the cell volume in 24 hours at 39 psig. Details of
the leak test are covered by Shield and Containment Check Lists, (Section LE,
Part VIII, Operating Procedures).

— — — — om——

The vapor-condensing system is isolated from the reactor cell by two
rupture discs. This system is designed to limit the secondary-containment
pressure to 39 psig during the maximum credible accident. This system
will be leak tested at the same time as the reactor and drain-tank cells
and to the same specifications.

The compensated volume consists of several sealed pipe volumes in
the reactor and drain-tank cells into which the cell pressure can be
admitted. The compensating volume can then be isolated from the cells and
cell leak rate determined by measuring the differential pressure between
the cells and this volume using a sensitive gage. Since this leak-tight
volume is inside the cells and at thermal equilibrium, the leak rate
measured should be independent of changes in cell temperature. The compen-
sating volume will be leak tested during fabrication., The effectiveness
of the temperature compensation will be determined during the prepower
leak testing. Details of the test are given in Test Memo XTI 2.3.1k.k,
2.3.1k.5 Calibration of Sumps

oumps are provided in the reactor cell and drain-tank cell to collect
the leakage from any in-cell water-containing equipment. Both sump levels
are measured by bubbler-type level indicators. The sumps will be cali-

brated for volume and liquid depth versus instrument reading. Detaile are

given in Test Memo XI 2.3.14.5.
2.3.15 Ventilation System

The ventilation system is designed to ventilate all areas where the
potential hazard from radioactive contamination is high. Subatmospheric
pressures are maintained in these areas by the stack fan. The air
exhausted from these areas passes through an absolute filter before it is
discharged to the containment ventilation stack. The following tests
of this system will be performed.
2.3.15.1 Filter Test

A DOP smoke test will be performed on the absolute filters to de-
termine their efficiency. This test will be performed at the normal,
operating flow conditions. Briefly the DOP smoke test consists of intro-
ducing dioctyl phthalate smoke upstream of the filters, taking air samples
on both sides of the filters and determining the percentage of the smoke
removed by the filters. This test is described in Section 3F, Part VIIT,
Operating Procedures and in ORNL-3442, "Tests of High Efficiency Filters
and Filter Installations at ORNL."
2.3.15.2 Standby Fan Operation

The design of the stack-fan control system is such that if Fan No. 1
or its discharge damper should fail, resulting in a pressure rise (less
negative) in the suction line, Fan No. 2 will be started automatically
to maintain ventilation. The operation of the system will be thoroughly
tested and the controls set to start Fan No. 2 when the pressure in the
main ventilation header rises above limits. This test is described in
detail in Section L4F, Part VIII, Operating Procedures.
2.3.15.3 PStack-Flow Indicator Calibration

The stack-flow indicator will be calibrated so that the amount of
activity released can be determined. The flow indicator will also provide
information regarding the condition of the absolute filters. Details of
this calibration are given in Test Memo XTI 2.3.15.3.
2.3.15.4 Demper and Valve Settings

The dampers and ventilation-control valves will be set for normal
operation as described in Section LF, Part VIII, Operating Procedures.
All ventilated areas will be checked for adequate ventilation. The effect

on the ventilation in various areas caused by operating dampers or doors
D=2

will be checked. Installed and supplementary portable instruments will
be used in making these tests. Details of these tests are given in
Test Memo XI 2.3.15.L4.

During maintenance, when the reactor cell and drain-tank cell are
open, and during possible accidents, the high bay is considered as secondary
containment. ZExcessive leakage in the high bay could prevent keeping the
area at a negative pressure when the ventilating system is operating.
Also upon loss of the stack fans during an accident, excessive leakage
could cause contamination of the building and surrounding area. The
leakage into the high bay will be measured by closing all inlet vents and
doors and all exit vents except one which leads to the stack. The one
vent will be throttled to give the desired negative pressure in the high
bay and the flow measured with portable instruments. Detaills are given
in Test Memo XI 2.3.15.5.

2.3.16 Liquid-Waste System

The liquid-waste system 1s used to transfer and store aqueous waste

 

material which may contain radioactivity or beryllium. This liquid waste
is pumped periodically to the Melton Valley waste-handling system., The
liquid-waste system is also used for clarifying shielding water used in
the decontamination cell and tank. Precritical testing will consist of
the following items.,
2.3.16.1 Leak Testing

In order to assure that none of the tanks, cells, and piping leak,
they will be filled with water and physically observed or will be
physically checked during other tests described below. No pneumatic
or hydrostatic tests are projected. Details are given in Test Memo
XTI 12.3.16.1.
2.3.16.2 Tank Calibration and Waste-Pump Flow Rates

To facilitate future calculations of the amount of activity present
and the amount of caustic necessary for neutralization, the waste-tank
volume will be calibrated vs the level indicator. The waste pump will be
calipbrated to determine the flow rate vs discharge pressure. Details are

given in Test Memo XI 12.3.16.2
2-25

To determine the proper operating conditions for the waste filter,
shake down tests will be made. Details are given in Test Memo XI 2.3.16.3.
2.3.16.4 Transfer to Melton Valley Waste System

Several practice runs will be made to test the ability to transfer
waste to the Melton Valley waste system. Procedures given in Section 3J,

Part VIII, Operating Procedures will be followed.
2.3.17 Samplers

— — e — a—— — — — — - — — ———

The fuel and coolant salt samplers of the MSRE were developed and
tested by the Development Section of the Reactor Division and are
described in ORNL semiannual progress reports from 1961 to 1965. The
only testing on site will be the equipment leak check and operational
tests of the assembled unit during operator training.
2.3.17.2 Graphite Sampler

The removal of graphite and INOR-8 samples from the reactor core will
be tested by using remote procedures before going into power operation.
The procedure to be used is Remote Maintenance Procedure No. 21, Part X,
Maintenance Equipment and Procedures.*
2.3.17.3 Offgas_Sampling

The offgas sampling system consists of in-line conductivity measure-
ments and a chromatograph and also cells for liquifying and removing
samples in shielded containers. This system is being designed by the
Reactor-Division Development Section and will be bench tested before
installation. The system will also be checked after installation by
addition of known mixtures of gases for testing and calibration.

2.3.18 Control Rods

The three control rods used in the MSRE were designed and thoroughly
tested by the Reactor-Division Development Section before delivery to the
MSRE. After installation they will be tested to assure that they function

properly.

 

*
E. C. Hise and R. Blumberg, MSRE Design and Operationg Report, Part X,
Maintenance Eguipment and Procedures, USAEC Report ORNL-TM-910, Oak Ridge
National Iaboratory, in preparation.

 
— o — — —

The rate of fall of each control rod will be used as a guide to the
mechanical condition of the rod. The drop time, which is less than one
second, will be tested as described in Test Memo XI 2.3.18.1.

The position of each control rod is indicated by syncro position
indicators. The lower positions of the rods are also indicated by changes
in back pressure on the component coolant air as each rod passes through a
restriction., This 1s called the fiducial zero. The syncro position
indication will be checked against the fiducial zero for each rod. Future
periodic checks of this relationship will provide information about pos-
sible stretching of the rod mechanism or other difficulties.

2.3.19 Heaters

Electric heaters are provided for all parts of the systems, in order
to preheat all components before the addition of the salt and to maintain
the temperature of the salt during zero-power operation.

To permit proper operation and maintenance, the location of each
heater must be known as well as the routing of the power leads through
the breakers, controllers, junction boxes, and disconnects. Tests will
be made during construction and precritical testing to ascertain that
these are installed as designed. Measurements will be taken of the
heater-circuit resistance, the resistance to ground, and the thermocouple
response, The details of these tests for the reactor cell are given in
Test Memo XI 2.3.19.1-A, for the drain-tank cell in Test Memo XTI 2.3.19.1-3B,
and for the coolant cell in Test Memo XTI 2.3.19.1-C.

Control of the electrical input to the heaters is entirely manual,
in response to system temperature. Powerstat and induction-regulator
voltage controllers are used. Since the heaters have excess installed
capacity, the powerstats will be provided with mechanical stops and the
induection-regulator limit switches will be adjusted to limit the system

temperature to 1300°F or below. During the first few heatups of the
2-27

systems the settings and the ammeter readings of each controller will be
determined. The details of these tests are given in Test Memos XI 2.3.1.1-B,
XTI 2.3.2.2-B, and XI 2.3.3.1-B.

2.3.20 TFreeze Valves

The freeze valves were tested by the Development Section of the Re-
actor Division prior to installation. On-site testing will consist of
adjusting heater settings and air loadings to give the proper thawing and
freezing characteristics for each freeze valve.

2.3.21 Miscellaneous
2.3.21.1 Leak Check of Equipment and Piping

Prior to critical operation, all process piping and eguipment must be
essentially leak tight. Other piping not directly connected to salt lines
must not leak excesgssively. Testing will start with components as they are
completed and will continue throughout construction and early operation.
The following is a description of the tests which will be performed.

Salt Piping — All reactor-cell, drain-tank-cell, fuel-processing-cell,
and coolant-cell piping and equipment which is prefabricated and assembled
outside the cells will be evacuated and given a standard helium leak test
to <10 ® cec of helium per second. All salt piping and welds not tested
before installation will be pressurized with helium after installation,
and leak tested in the cell. To increase the sensitivity, the sections to
be leak checked will be sealed in plastic, and the inside of the plastic
will be surveyed with a helium leak detector. Any indicated leakage will
be located and repaired.

Auxiliary Systems — Both in-cell and out-of-cell equipment and

 

piping will be pressurized and soap checked or helium leak tested. In
addition to the pressure test the cover-gas system will be checked to
assure that there are no leaks which would allow back diffusion of moisture
into the cover-gas system. An increase in moisture between the treating
station and the cell penetrations would be an indication of this.

Both in-cell and out-of-cell treater-water piping will be hydro-
statically tested for leaks. The thermal shield will be valved off during
this test. All spool pieces between sections of the thermal shield will
be helium leak tested before installation, and all field welds which

connect to the thermal shield will be x-rayed to ensure that no leaks exist.
2-28

The leak-detector headers will be pressurized to 125 psig and ob-
served for time-dependent pressure drop. Leaks will be located by soap
testing. The leak-detector lines will be assumed to be leak tight unless
the closure monitored by any leak detector cannot meet satisfactory leak
rates. If the leak is not located at the flanged joint, the leak-detector
line will be inspected.

Lines and equipment in these systems will be tested by soap testing.

The MSRE will use an external neutron source of curium and americium.
This source will be located outside the reactor and on the opposite side
from the neutron-detecting instruments. The source has to be of sufficient
strength to give a finite reading on the wide-range counting channels
with the reactor empty. To properly size the neutron source to be used
during operation, an available source of known strength will be used to
obtain preliminary readings with the reactor empty before fabricating
a source for the reactor. Details are given in Test Memo XI 2.3.21.2.
2.3.22 Entire Plant

When construction is complete, the entire integrated plant will be
operated to test every aspect of the reactor except the nuclear behavior.
A number of the tests described above for the various systems require the
entire plant to be in operation. These will be completed at that time.
Oxygen will be purged from the equipment and the entire system tested
to assure that 1t is leak tight.

A number of normal startups and runs will be made to correct errors
in operating procedures, check adequacy of the integrated design, shake
down equipment, and train operating personnel.

When operation is reasonably stable and equipment and instrumen-
tation are functioning properly, tests will be made to determine the
effect of other operating conditions or modes. DBaseline data will be ob-
tained; adequacy of instrumentation will be checked; thermocouple biases
will be determined; and inventory methods will be evaluated. ©Several heat
balances will be taken and at least one routine pressure test will be made.
Puel- and coolant-system samples will be taken to test the operation of
the samplers and to investigate inleakage of oxygen and other contaminants
or changes in the composition of the salt. ©Salt additions using the

samplers will be tested.
2-29

At the end of the series of integrated runs, samples of the graphite
will be removed to determine salt permeation and physical damage.

Before adding the fuel, internal and external examinations will be
made for indications of excessive wear or corrosion. All in-cell modifi-
cations and repairs will be completed before criticality.

Details of these integrated runs will be given in the daily shift
instructions and in the run instructions.

Heat balances will be made on the MSRE system to determine the
thermal power that is generated and to provide a check on the other
methods of 1ndicating power.

The heat balance will be made by considering the reactor cell and
the drain-tank cell as an envelope and measuring all the energy that is
added to or taken from this envelope, the net energy removed being the
thermal power generated by the reactor.

There will be some heat sources and sinks that will be small and
therefore not evaluated, i.e. heat removed by the cover-gas system and
heat removed by keeping cell pressure below atmospheric are negligibly
small. There will be others, especially heat sinks, that cannot be
measured directly and are not individually included in the evaluation.
These terms are evaluated collectively in a correction term called "heat
losses."

At a time when the system is hot and circulating but no power is
being produced an evaluation of the heat-loss term can be made. This
may be accomplished by measuring the energy added to the envelope by
heaters, fuel-circulating pump, space-cooler motors, etc., and the energy
removed by cooling water, cooling salt, cooling oil, cooling air, etc.;
the difference between these two energy tabulations will be the term in
guestion. This term will be evaluated several times both before and
after power operation has begun. By the time this measurement has been
made several times the value should be known with good statistical

conf'idence.
2=-30

The heat balance will be calculated pericdically (normally, every

4 hours) by the on-line computer. Hand calculations of the heat balance

will be made for comparison with computer results. The details of the

method used for calculating a heat balance are essentially the same for

both the manuval and computerized approaches. These details are described

in Test Memo 2.3.22.k4.
SECTION 3
ZERO POWER EXPERIMENTS

3.1 OBJECTIVES

After the non-nuclear operation of the reactor system has been
adequately demonstrated, a program will be started whose ultimate ob-
jective is the operation of the reactor at full power. The initial
phase of this program is a series of experiments to establish the basic
nuclear and related characteristics of the system at essentially zero
power. The actual power level for these experiments can not be precisely
defined because accurate power calibrations will not be available until
after the system has operated at substantial powers. In general, the
power level during this phase of the operation will be a few watts with
a limit of about 10 kw. (The power will be kept low to minimize the
activity of the fuel in the shutdown preceding power operation.)

The first of these experiments is the initisl critical experiment.
During this experiment, enriched uranium in concentrated form will be
added to the fuel carrier salt in a carefully controlled manner to bring
the #3%y concentration up to the minimum required for criticality at
1200°F (no circulation, rods fully withdrawn). One purpose is to check
the caleulations of clean critical concentration. Preliminary informa-
tion on the concentration coefficient of reactivity and the effects of
circulation'will alsc be obtained. Finally, from the base point estab-
lished in this experiment, the 25U additions necessary to bring the
concentration up to the dperating levél can be made with confidence.

The additions of fuel to establish the operating concentration will
be used to compensate for control-rod insertion in experiments to es-
tablish rod worths as functions of position, temperature, uranium con-
centration, and pressure. In addition measurements will be made to
evaluate the reacfivity coefficients of the various perameters. Enough
enriched uranium will be added in these experiments to permit calibration
of one control rod over its entire range of travel.

Numerous samples of the fuel salt will be analyzed during the

course of the uranium additions. The primary purpose is verification
3-2

of the uranium concentration at each point in the experiment, but the
samples will also yield valuable data on salt composition and corrosion.

An extensive dynamics testing program will be started during this
‘phase of the operation. The purpose .of this program is to investigate
system stability and to obtain information about the reactor that is not
available otherwise from static measurements. The separate temperature
coefficients of reactivity of the fuel and the graphite, for exsmple,
could not be determined from static measurements. However, since the
transient temperature response characteristics of the fuel salt and the
graphite moderator are quite different (the graphite temperatuje respond-
ing much slower), a dynamic measurement may be analyzed such that the
two effects can be isolated. Determination of other important reactor
paremeters, such as coefficients of fuel-to-graphite heat transfer,
and heat exchanger, radiator, and piping heat transfer coefficients will
also be attempted by dynamic tests. Another function of these tests is
to determine the forms of mathematical models which adequately describe
the transient behavior of the system.

Noise analyses will also be made of the neutron flux to evaluate

the mechanisms causing random perturbations in reactivity.
3.2 PROCEIURES

3.2.1 Initial Critical Experiment

This experiment basically consists of adding increments of enriched
urenium concentrate to the fuel salt mixture and observing the progress
toward the critical concentration by the increased source multiplication.
Successive additions of kilogram quantities of 233y to the salt in the
drain tanks, followed each time by a fill of the core and multiplication
méasurements, will comprise about 98 percent of the critieal amount
(69 kg 235J), The remainder will be added in 85-g batches through the
sampler-enricher.

A removable external source (241Am-24ZCm-Be) emitting 10° n/sec will

be used. (In the later stages the alpha-n source in the fuel salt will
3-3

enter into the experimental procedure.) Four neutron counting chamnels
will be used: two fission chambers in the instrument shaft, a EFg
chamber in the instrument shaft, and another BFx chamber in the thermel
shield.

At the beginning of the experiment, drain tank 2 (FD-2) will contain
salt lacking only the addition of enriched uranium concentrate to reach
the specified operating composition. The enriched concentrate (molar
composition, T3% LiF-27% UF,, in which the ursnium is 93% 235U) can be
added directly to FD-2 from storage cans, each containing 15 kg of
235, The amount added from a can can be positively limited by adjust-
ment of a dip tube. , '

The temperature of the core will be maintained at 1200°F throughout
the experiment. Except for the Very last step, the count rates used in
predicting the critical condition will be taken with all 3 control rods
withdrawn to their upper limits. The rods will be partially inserted
while the salt level is rising in the core and while uranium is being
added through the sampler-enricher and will be fully inserted when the
fuel circulating pump is being started.

Reference count rates will be determined with the barren salt at L
levels in the core qnd with the reactor vessel full. During each fill of
the reactor following an addition of 235U in the drain tank, count rates
will be measured at these same levels. Count rates will also be measured
with barren salt circulating, when the core density is reduced by the
presence of entrained gsas.

The first addition of enriched concentrate to FD-2 will contain
45 kg of 235U, or 64% of the predicted critical amount. The sizes of
later additions through FD-2 will be specified on the basis of extrapo-
lation of plots of inverse count rates vs amount of =>5U in the salt.

The intention is to meke four additions through FD-2, bringing the &35y
concentration to 64, 87, 94, and 98% of the minimum critical value.

After count rates have been measured with the salt levels in fhe
reactor vessel, the loop will be filled and circulated. The purpose in
this is to insure complete mixing, to obtain samples for uranium analysis

from the pump bowl, and to observe the reactivity effects of circulation
3-k

(loss of delayed neutrons and entraimment of gas in the circulating
'salt). After the third and fourth additions, the external source will be
temporarily removed, to permit observation of the multiplication of the
internal source.

| When the count rate plots show that the 25U inventory is less than
1.5 kg below the critical loading (expected after the fourth addition
through FD-2), the remaeinder of the concentrate will be added through

the sampler-enricher.

During the addition of the remaining 35U in B5-g increments,
count rates will be detefimined at intervals with circulation stopped and
the rods fully withdrawn. Plots of inverse count rate vs 35U concen-
tration will be used to extrapclate to the critical point for these
conditions. As an aid to this extrapolation, coufit rates will be
measured with the salt circulating after each 85~g addition and plots
made. (The circulating points will be displaced in k from the non-
circulating points because of the delayed neutron and entrained gas
effects.)

Count rates will be measured both with the external source and
without it. The count rates without the external source will be used
later to evaluate the internal alpha-n source (after count rate-fission
rate correlations have been established).

When the extrapolations indicate that one more capsule will raise
the 223U concentration above the minimum critical value, the increment
will be added, circulation will be stopped, and the reactor made critical
by rod withdrawal.

3.2.2 Calibration of Control Rods

The  theoretically desirable objective of the control-rod calibra-
tion experiments is to measure the reactivity worth of each rod as a
function of the positions of the other two rods, the uranium concentra-
tion in the salt, the core temperature, fission-product poison distri-
bution, and the volume of entrained gas in the salt. In the practical
sense, separation of several of these effects iz difficult, and some
compromise has to be made in determining the effects most important to

the MSRE operation. Because of the fluid nature of the fuel and the fact
3-5

that, once added to the salt, removal of uranium is inconvenient, the
basis for the experimental work relating to rod calibration must be the
sequential addition of uranium leading to the amount required for full
power operation. In general, therefore, after a specified uranium
addition the experiments will be designed to measure'fihe effects of the
importent varisbles other than uranium concentration on reactivity and
control-rod worth and to provide experimental cross checks on the measured
worth at the new concentration. '

Several considerations enter into the specific design df the rod
calibration experiments. It is anticipated that during normal operation
of the reactor all three control rods will be partly inserted. The two
shim rods will be stationary, except for occasional édjustments following
power changes. The regulating rod may be driven up and down for short
distances rather frequently as the servo controller holds the power or
temperature at the set point. Present plans are to keep the shim rods
well above the regulating rod during normal operation. The reason 1s
that if the rods were kept at about the same level, there would be sharp
changes in regulating-rod sensitivity as the regulating rod moved into
and out of the shadow of the nearby rods. In addition to.operating the
regulating rod out of the shim rod shadow, it will be practical to
require that the shim rods always be held at nearly equal positions.

The maximum amount of uranium to be added to the fuel salt is that
amount required to attain full power operation, with maximum poisoning
and burnup occurring when all rods are withdrawn to the limits of their
operating ranges. OSince this amount depends on the xenon poisconing and
several other effects which will not be known with precision in advance
of the approach-to-power tests, the maximum uranium added to the salt
will be initially limited to the amount required to calibrate one rod
over its entire length with the other two rods fully withdrawn (~2.3%
‘Ak/k). More uranium will be added and the rod calibrations continued
if it proves necessary during the approach-to-power tests.

The general technigques which will be used in the calibration experi-

ments will be rod bump-stable period measurements to obtain differential
3-6 . ' ‘1!

worth data for a single rod with the other two rods held at fixed ot
positions, and rod drop-subcritical counting rate measurements to obtain
the integral worth of a specific rod configuration.

The sequence of measurements in the rod calibration experiments is
expected to be as follows. An initial series of uranium additions to
the fuel salt will be made to bring the reactor critical with the fuel
stationary at the nominal operating temperature of 1200°F, and with all
three rods fully withdrawn. In subseguent calibration tests, the ex-
ternal neutron source will be reinserted in order to provide significant
counting rates in subcritical measurements. With the initial condition
of rods fully withdrawn and the neutron level high enough to make the
contribution from the external source negligible (~10 watts), one of
the rods will be dropped with the other two held fixed, and the sub-
critical counting rate will be measured as a function of time. The
rod will then be withdrawn again and the measurement repeated by drop-
ping another of the three rods. After the individual worth of each of
the three rods has been measured, the rods will be dropped in succession
and then as a group to determine the cumulative worth.

Upon completion of the initial rod-drop tests, circulation will be
established and small additions of highly enriched uranium to the salt
will be made by dissolution in the pump bowl. The total amount added
will be that required to reattain criticality with one rod inserted a
small distance in the core (approximately 10% of total insertion). With
the core critical and the fuel stationary at 1200°F, rod-bump measure-
ments will be made by moving the rod upward a small distance from the
critical position and observing the stable reactor period which ensues.
It is anticipated that practical stable periods for these tests will
range from & lower limit of 30 seconds to an upper limit determined by
the uncertainties in the rod positions.

After the first rod-bump measurements are completed, the circulating
pump will be started and the change in rod position required to reattain
criticality will be determined. With the fuel circulating at the iso-
thermal temperature of 1200°F, the rod bump-stationary period measurement

will again be made. : ~
3-7

The sequence of'tésts involving uranium additidn, periocd measurement
with Puel stationsry, and period measurement with fuel circulating will
be'repeated for several additions of uranium uwntil the rod being cali-
brated is inserted approximately 25% of its total travel. At this
point, with the fuel stationary the rod bump-period measurement will be
repeated for the other two rods, each time with two of the rods fully
withdrawn and one rod inserted to attain criticality. Also at this
point, calibration méésurements for one rod at intermediate insertion
positions of the other two rods will be initiated. For example, &
critical configuration wlll be obtained with the two shim rods inserted
equal, small distances in the core, somewhat less than the rod being
calibrated, and thé rod bufip-period measurements will be repeated.

Upon completion of the above tests, rod drop measurements will be
repeated for this Intermediate uranium concentration. Firsfi, the
partially inserted rod will be dropped while the other two remain fully
wlthdrawn. Then the rods will again be dropped in successiocn.

For the next test in the series of the calibration experiments at
the specified intermediate uranium concentration, the isothermal tempera-
ture of the éysfiém will be varied by manipulation of the external heating
elements on the circulating loop. This experiment will be limiteéd to a
range of approximately 1100°F to 1250°F, which corresponds to a calcu-
lated reactivity change of about 1.4% Ak/k in the initial MSRE fuel.
Tempereture changes produced in this menner afe very slow, so that 1t
will be practiéal to heat the loop until eithef the rod is withdrawn
to its upper limit or the upper limiting temperature is reached. Then
the loop will be allowed to cool slowly, and the position of insertioh
of the rod as & function of the temperature will be recorded. At the
upper and lower limits of,the'temperature cycle, stable-period and rod-
drop measurements will again be made in order to determine the change in
reactivity worth of the rods produced by the change in reactor temperature.

The final phase in this series of tests will provide information
relating to the pressure coefficient of reactivity. In this connection,
the overpressure in the pump bowl will be increased by several.psi above

the normal value of 5 psig and the steady-state critical rod
3-8

configuration will be recorded. The vapor space in the fuel loop will
then be isolated from the drain tanks and the drain tanks vented to pro-
vide a pressure sink. Then, with the drain-tank vents closed, the inter-
connection to the fuel loop can be reopened to produce a rapid decrease

in overpressure. The control-rod motion required to maintain criti-
cality will provide some information about the prompt pressure coefficient
as well as a check on the long-term pressure effect.

The series of measurements described above can be considered to
comprise those tests which correspond to a substantial addition or
uranium to the salt. This series will be repeated for several sequences
of uranium additions. The terminationé of these sequences are antici-
pated to correspond to insertions of 40%, 50%, 60%, 75%, and 100% of the
rod being calibrated. |

The rod calibration program outlined here has the advantages of
requiring no special equipment and of providing a maximum amount of
calibration data at each stage of fhe tests. As the data is accumu-
lated, the influence of the important variables affecting the reactivity
can be separated and the rod worth for a fixed set of core conditions
can be determined by integration of the data.

3.2.3 Evaluation of Nuclear Parameters

Part of the task in the zero-power control-rod callbration experi-
ments will be to determine the separate reactivity coefficients corre-
sponding to changes in uranium concentration, in isothermal temperature
of the core, in effective delayed neutron fraction and bubble entrainment
when steady-state circulation is established, and in system overpressure.
As these changes are introduced, the quantity directly measured is the
change in rod position which compensates for the associated reactivity
increment and results in a new critical configuration. The reactivity
coefficients of each variable must be determined by correcting the
measured reactivity increment for the change in the total worth of the
rod. Thus the complete analysis of these experiments requires an
"unwinding" of the data obtained from both rod bump-period measurements
and the rod-drop measurements. For the important case of uranium

addition in the range required to calibrate the rods, theoretical
3-9

calculations indicate that this coupling effect will be very small.
Hence the equivalent fuel addition corresponding to a given insertion
of the rods should provide a reliable standard of comparison for the
direct measurement techniques used in obtaining the rod worths.

3.2.4 Preliminary Studies of Dynamics

A variety of dynamic tests is planned for the zero-power run for
the purpose of determining system parameters, verifying mathematical
models describing the reactor kinetic behavior, and measuring the
characteristics of the reactivity-perturbing functions. Estensive use
will be made of the on-line computer in these tests (and throughout
the dynamic testing program) for the acquisition of data. Some vari-
ations from the normal mode of operation of the computer will be required
in some cases to provide the data required.
3.2.4.1 Non-Nuclear Tests

A number of non-nuclear tests will be performed during the flushing
operations prior to the initial critical experiment. These tests are
designed to produce background and reference information to be used in
the analysis of subseguent experiments. ©OSpecifically, tests will be run
to:

1. determine transient salt flowrates for pump startup and coastdown;

2. determine the effects of the loop heaters on loop temperatures and
on nearby thermocouple readings;

3. measure the transient thermal-response characteristics of various
locp components and thermocouples; and

L. measure the non-nuclear background nolse in the detection channels
to be used for flux-noise measurements.

Transient Flow Rate Measurements — Since the coolant-loop flow
rate is monitored by a venturi meter with two readout devices, a direct
megsurement of flow startup and coastdown can be made. Since the
startup transient will be fast, however, it will be necessary to monitor
the output of the flow transmitters directly, because there are two

magnetic-amplifier devices between each transmitter and the computer.
3-10

For the coastdown test, it is necessary to have the loop isothermal (as
will be the case at zero power) so as to avoid thermal-convection flow.
The pump speed transients will also be recorded.

To determine the fuel-loop flowrate transients, it will be necessary
to extrapolate from coolant-loop measurements. It is expected that salt
flowrate and pump speed will coast down "in unison"” in both loops; if
this is verified by coolant-loop measurements, then the fuel flow coast-
down can be determined from fuel-pump speed. If not, more sophisticated
caiculations may be reguired.

A gamma-ray densitometer, which will be installed about 8 feet
upstream of the reactor vessel inlet, will be used to monitor the
transient effects of pump startup and coastdown on the behavior of
bubbles entrained in the fuel salt. The densitometer output will depend
on how much gas is held up in the loop (both static and circulating)
at the start of the transients. There is a possibility of bubbles
agglomerating in the loop after a flow stoppage; this may be detected
by monitoring the densitometer after flow is restarted.

Effects of Loop Heaters — Steady-state heat loss data (abt several

 

temperatures) may be useful for correcting low-power test results. The
transient response of circulating-loop salt temperstures (with the other
loop static) to a change in power of a group of heaters will indicate
the time response of coupling between heaters and salt. This will be
done ir both the primary and secondary loops, since the heaters are
differernt. At the same time, the transient effects of the heater power
changes on nearby thermocouple readings may be megsursad.

Temperature Response Measurements — Temperaturs response measure-

 

ments of various components in the fuel and coolant loops and of thermo-
couples will bs made by introducing temperature pulses into the loops.

A hot-slug pulse will be introduced in the fuel loop as follows:
achieve thermal equilibrium (as nearly as practical) with the fuel
loop stagnant and the coolant loop circulating at a higher temperature.
The fuel in the heat exchanger will then be heated to coolant-100p

temperature. Since the heat exchanger is high in the fusl loop, convection
3-11

flow should be negligible. When the fuel pump is started, this hotter
fuel slug will pass through the cooler sections of the loop giving the
temperature pulse. By measuring reactor inlet and outlet temperatures,
its thermal transfer function can be determined. Comparison of the
response of the thermocouple in the well at the reactor outlet with
nearby thermocouples on the outside of the piping will indicate the
response time of the pipe thermocouples.

A temperature pulse in the coolant loop will be introduced in the
same manner; here, however, the static coolant salt in the heat exchanger
should be cooler than the rest of the coolant loop, since the heat ex-
changer is approximately at the low point of the coolant loop. Since
there 1s some piping below the heat exchanger, there may be convection
flow problems. This cold-slug test will be used to measure the radiator
salt-side transfer function and to check the response of the thermo-
couples on the pipes by comparison with those in the wells at the
radiator inlet and outlet.

Background Noise Measurements -— Preliminary tests will be made to

 

determine background levels for the low-power flux-noise measurements.
A special flux-measurement channel that was designed especially for
sensitive noise measurements will be used. The detector will be inserted
in a spare hole in the nuclear instrument shaft. Analog tape recordings
of the channel output noise will be taken both with and without fuel
circulation. Analysis of this data will reveal any effects of vibration
or electromagnetic radiation that will be present in later nuclear tests.
The noise output of the regular flux-measurement channels will be compared
with that of the special noise detector channel.
3.2.4.2 Tests in Critical Reactor

Various nuclear characteristics will be measured by dynamic tests
performed with the reactor critical. The neutron kinetic behavior will
be affécted by fuel circulation due to both the effects of a loss of
delayed-neutron precursors from the core (i.e. a reduction in B) and of
the entrained gas in the fuel salt. The tests will be designed to
determine the separate effects of each. The effects of fuel-loop

overpressure on reactivity via changes in fuel density will also be
3-1z

measured. Determination of the zero-power neubron=kinetics transfer
function and of separate temperature coefficients of reactivity for the
fuel and graphite will be attempted.

Effects of Flow Transients on Neutron Kinetics — One test will be

 

made with the reactor critical, fuel and coolant loops isothermal,
fuel loop static, coolant loop circulating, and the flux servo controller
on. The fuel pump will then be started, and the densitometer reading
and rod reactivity addition required to keep the flux level constant
will be monitored. Assuming that the flux controller keeps the reactor
critical, the rod reactivity changes will be equal (and opposite) to
the reactivity effects of circulation. Since the effects of delayed-
neutron precursor losses will appear quickly compared to the more gradual
buildup of voids in the loop, the two effects may be separable. After
density equilibrium is attained, the fuel pump will be shut off again,
and the rod motion monitored. The effects on reactivity of the increase
in B and of the bubbles floéting up out of the core will then be measured.
The steady-state high-frequency fluctuations in density (as measured
by the densitometer) are not expected to affect the steady-state neutron-
level fluctuations due to the lon (i.e. 8-sec) fuel residence time in
the core, which will "average out" these higher-frequency fluctuations.
Neutron-level fluctuations with hydrodynamic-pressure fluctuations in
the core are expected to be much greater since they would modulate the
entire core gas volume; however, the fuel-salt pressure cannot be
monitored. A cross power-spectral-density analysis will be made to
see 1f there 1s any correlation between the densitometer output and the
neutron level.
Should pockets of entrained voids be present in the static fuel
loop, starting the fuel pump may sweep falrly large vold volumes into
the core. This would show up in the densitometer signal, and a few
seconds later as a negative reactivity pulse. 1If this occurs, it may
be possible to determine a reactivity-to-void transfer function.

Effect of Fuel Loop Overpressure on Reactivity — A rapld increase in

 

pressure in the core will compress the entrained gas, increase the fuel

density, and thus increase the reactivity. A slow increase in pump-bowl
3-13

pressure, however, will increage the density of the gas entrained in the
fuel salt, and (assuming the rate of the gas volume entraimment in the
bowl is constant) will result in a net increase in core void volume,
hence a decrease in reactivity. The transient characteristics of these
effects will be studied by slowly building up the pressure in the fuel-
pump bowl, then guickly venting the bowl to a previously vented drain
tank through the by-pass line.

Zero-Power Neutron-Kinetics Measurements — Dynemic tests will be

 

made to give information about the neutron-kinetics transient-response
characteristics and transfer functions. These experiments will involve
transients induced by small changes in control-rod position. The tests
will be made both for stagnant fuel and flowing fuel.

The first test will be a control-rod pulse. A total reactivity
insertion of no greater than 0.02% Ak/k is considered adequate unless
the flux noise level is high. This will require a rod motion of 1/2
to 2 inches, depending on the location of the rod tip. Pulses ranging
from 5 seconds duration to 30 seconds duration will be used. At the
end of the pulse, the rod will be returned to its initial location.
Tests on the control-rod mock-up indicate that rods can be positioned
with sufficient accuracy. An avtomatic timing device will control the
rod drive motor for this test.

Another set of tests will use pseudo-random binary input. The
control. rod will move a preset distance using the automatic timing
device. The test will involve a series of rod motions in which the rod
is moved in and out around the critical position. The pattern of pulsing
will have special properties that will facilitate the fregquency-response
analysis. The tests will use pulses of 1 to 60 seconds with reactivity
insertions of less than 0.02%.

Flux noise spectrum measurements will also be made. These should
give good information in the high frequency range (1—10 cycles per
second) where the transfer function roll-off occurs. Since this
roll-off is determined by the valve of B/fl*, it will furnish additional
information for separating the relative effects of bubble circulation

and precursor circulation.
3-14

Determination of Separate Fuel and Graphite Temperature Coefficients
of Reactivity — By introducing a hot-fuel-slug pulse into the core
(as described previously), but with the reactor critical and on flux-servo
control, the effects of fuel and graphite temperature on reactivity may
be determined. The reactivity added by the rods would be equal (and
opposite) to that due to the temperature changes, and to the effects of
a reduction in P and vold entraimment. Since these last two effects
will have been measured previously, the temperature effects may be
separable.
3+2.5 Evaluation of Neutron Sources and Future Requirements of the

External Source

 

The external source is one in which alpha particles from americium-241
and curium-242 interact with beryllium to produce neutrons. The source
was fabricated by encapsulating a mixture of beryllium with 2 curies of
24lpm (462-y half-life). It was then irradiated for 21 days in the ORR
to bulld up about 380 curies of 24%Cm (163-day half-life). The resulting
source emitted about 10° n/sec Just after irradiation. Because most of
the neutrons come from 242Cnm alphas, the source initially decays with
the 163 day half-life and is expected to become inadequate for MSRE
startup requirements in about one year if it is not reirradiated. The
source will be exposed to a flux of approximately 1 x 1012 n/em® sec
when the MSRE is at 10 Mw, but this 1s not high enough to maintain an
adequate source when the target is only 2 curies of 24lim. The intention
is to install a new source after one year, containing enough Z4%Am
target so that the flux in the source tube can keep an adeguate amount
of 242Cm built up.

Since the reliability of the flux calculation is limited for positions
far away from the core (~ 20" from the core, ~ 50" from the center of
the core), the amount of Am needed for an adequate source cannot be
specified accurately. The high cost of Am (~ $1600/g) makes it desirable
to specify the smallest practical amount of Am for the permanent source.
Therefore, 1t will be necessary to make measurements of the flux in the
source tube at low power © that the useful 1life of the initial source

and the future requirements may be calculated. This will be accomplished
3-15

by withdrawing the source while the reactor is operating at ~ 10 Kw and
inserting an array of gold and copper foils into the source tube. This
will determine the flux at the specified power. Since the flui is pro-
portional to power, the degree of activation of the Am can be calculated
for any given period of operation. With this information we can determine
the amount of Am needed to make & source that wiil have a minimum cost
and a maximum practical lifEe

Another experiment plamned for this time is the establishment of
the intensity of the internal (inherent) neutron source of the clean
salt in the core. This will be accamplished by measufing count rates at
various detectors when the reactor is only slightly subcritiecal, both
with and withbut the external source present, The relations between
counting rates of the detectofs-and fission rate (neutron production in
the core) will be determined, at power levels where the source contri-
bution is negliglble, by-heat balance and other calibration techniques.
These relations will bé applied to the subceritical data (because the
spatial flux distribution and neutron leakage probabilities 1n the core
do not change much between the high—multiplicatioh, gubceritical condition
and the critical condition) to evaluate the effective, in-core neutron
sources from both the external asnd the inherent sources.

3.2.6 Chemical Analyses

| During the course of the zero-power experiments numerous samples
of the fuel salt will be anslyzed for uranium. The analytical data will
be uged to Suppdrt and supplement the calculations of uranium.concen—
trations from inventory considerations.

The fuel solvent will be sampled prior to the initial critical
experiment both while 1t 1s in the drain tank and while it is circu-
lating‘in the fuel system. AIn addition, many of the fuel samples taken
during the experiments will be analyzed for many other constituents
besides uranium, ' The coolant system will also be sampled. The purpose
of this progfamiwill be to esf&blifih analytical bage levels of constit-
uents and contaminants to compare with analyses in subsequent stages

of the MSR experiment.
3-16

Analyses will be obtained for the component metal fluorides and
also for the diggolved metals which would result from torrosion — iron,
chromium, nickel, and molybdenum. Chemical analyses will also be
obtained for the oxide content &nd the redueing power of the salt. It
must be pointed out that in the beginning stages of the experiment,
"reducing power" of the salt should not be inferred to represent con-
centration of reduced urenium species because finely divided iron and
nickel, probably present as impurities, can also give rise te."feducing
power" values. Petrographic examinations will be made of fuel solvent
and'fuei'Spetimens in the early steges of the zero~powe£_teét period to
afferd_fi bageline for optical data since the petrographic methed may
have unigue application in later stages of the Molten-Salt Reactor
Experiment.
SECTION 4
LOW POWER MEASUREMENTS
L.1 OBJECTLVES

 

After the zero-power experiments, the reactor system will be shut
down to make the final preparations for operation at significant power
levels. These preparations will include the hermetic sealing and testing
of the secondary containment, installation of all shielding known to be
required, final modification and adjustment of the heat-rejection system,
and any maintenance work which may be required. This work will be
followed by the next, or low-power, phase of the test program.

The power level of the reactor will be limited to sbout 1 Mw
during this phase of the program to avoid most of the effects of power
and still permit the determination of the required information. This
power level is the point, in routine operation, at which a transition
is made from automatic control of the neutron flux (with independent,
manual control of temperature) to control of both temperature and flux.

This phase of the operation will afford the first opportunity to
evaluate many of the power-associated characteristics of the system.

A1l of these characteristics will be studied in more detail and evalu-
ated more accurately as the power level is increased but the measurements
at low power provide the information on which the power increases are
based. 1In this connection: |

1. The biological shielding and containment will be surveyed for
adequacy and to locate area which may be of questionable adequacy at
higher powers.

2. The nuclear power instruments will be calibrated and adjusted to

provide the desired ranges of applicability.

3. Preliminary values will be obtained for the power coefficient of
reactivity.
L. The behavior of the noble gases will be observed and xenon poison-

ing will be measured.
In addition to the items mentioned above, an extensive program will

be put in effect to evaluate the nuclear, thermal, and mechanical
L.2

performance of the system. Much of this work will make use of the on-
line computer for the acquisition and processing of data. The various
computer programs will have been checked out, but this operation will
permit evaluation of the mathematical treatments used in the programs.

Of particular interest will be the programs which calculate the reactivity
balance, heat balance, and salt inventories.

The experimental analysis of the kinetics of the system, which was
started at zero power will be continued in all areas where useful infor-
mation can be gained. The objectives will be essentially the same as
those described for the zero-power operation, but it is expected that
the earlier parsmeter and model estimates can be upgraded.

Samples of the circulating salts will be removed periodically during
this, as well as all other, phases of the program. These operations will
provide an opportunity to evaluate and modify, if necessary, the tech-
niques and procedures for handling radicactive samples. The results of
the sample analyses will be used in conjunction with earlier results to

establish baselines for the study of the effects of power operation.

L.2 PROCEDURES

L.2.1 Shielding and Containment Surveys

Surveys will be made in all areas that are accessible during re-
actor operation; these surveys will be carried out at all power levels
including the highest expected during any run.

Except for the top plugs, the shielding around the reactor cell is
essentially that which was installed for the ART. Calculations indicate
that before the MSRE is operated at full power, additional shielding
will be required in some areas. Space was left (on the outside of the
shield) for supplementary shielding, but none was installed because diffi-
cult source and shield geometries made it impossible to predict accurately
the requirements. Instead, radiation levels measured in low-power
operation will be extrapolated to high power to determine the amount of

additional shielding actually required.
b-3

One ares where stacked block shielding will be added is on the
southwest of the reactor cell.  There the originél ‘shielding is thinner
in the vieinity of the coolant piping penetrations and high dose rates
are expected in the coolant cell, the blower housg,and outside the blower
house. Another area which will be given special attention is the north
electric service srea. There penetrstions in the wall of the south |
electric service aree will be monitored to determine if local shielding
is required, »(The south electric service ares is shieldéd from the
reactor cell only by the cell annulus, and entry will befprohibited
during power operation.)

Surveys will also cover ell other areas, for example: the water
roem, the special equipment room, the sump room, transmitter room,
service room, and service tunnel, The fuel system éampler-enricher will
receive speciel attention in this survey because it not only is a shield-
ing problem but one of contalnment as well. Shielding surveys will be
made &uring the sampling operation and in the manipulation of the sample
after it has been witndrawn.

The top of the reactor cell will not be occupied during power
operation, but radistion levels will be measured there to complete the
survey of the shield and also to provide data which can be used for
simple checks of the shielding calculations. |

The primary accident containment, i.e. the Reactor and Drain-Tank
Cells, will be hermetically sealed and mnintained at a pressure of 2 psi
below atmospheric (12.7 psia). Containment integrity will be monitored
by a system of reference vessels installed in the cells to measure
pressure changes, by monitoring the oxygen content of the containment
atmosphere, and by determining the amount of exhaust from the vacuum
pumps.

Since the cell atmosphere will be kept at some. low oxygen concen-
tration (< 5%) by injection of nitrngen at a mgé,éuivéa _réte, it should
be practiéable to analyze for oxygen and relate the fherease in oxygen
concentration to cell inleakage. If a continuous pump down of the cell
is necessary to maintain the desired negati&e pressure, then that wnich

is exhausted from the vacuum pumps may be measured and this, when corrected
L-y

for nitrogen injection, will represent the inleakage. The reference-
vessel techfiique for measuring containment leakage has been used quite
extensively and is described in ORNL-CF-64-11~-31. All of the above
monitoring methods will be used until good confidence is established in
our ability to continuously determine the cell leakage.

The high bey of Bullding 7503 will be another containment barrier
in case of an activity release. The building will be kept at & negative
pressure Gw O.l“ Hz0) with reference to atmosphere. The magnitude of
this négntive pressure will be measured in the high bay as well as in
individusl cells and areams that are serviced by the building ventilation
system. Tt should be sssured that air movement will be from the less
contaminated to the more contsminated parts of the building.

4,2.2 Calibration of Nuclear-Power Instruments

The neutron-sgensitive instruments which give signals proportional
to nuclear power are the two wide-range counting channels, the two
linear power channels and the thfee safety channels. The objective in
the calibrations is to establish relstionships between reactor power,
ingtrument output, and chember location. Preliminary values will have
been established prior to and during the zero power tests, but as the
povwer is raised to 1 Mw the greater brecision with which reactor power
can be measured will permit more accurate determinations. The final
relations will be determined later at high power, when heat balances
will be more accurate, but the calibrations at low power are necessary
to make the neutron instruments useful during the approach to full
power. | ,

Calibration date will be obtained on the various instruments at
several power levels up to 1000 kw. Below about 200 kw the most useful
nuclear-power data will probably be obtained from changes in the power
to the electric heaters. Rates of change of system temperature will
yield good information at powers above about 100 kw. System heat balances
will give good results above about 500 kw. |
4.2.2.1 Power Measurements |

Measurements of the change in instrument output as a function of
power will be made with the sensing elements in fixed locations. In '
-5

this cofidition, the relation between the two parameters will be essen-
tially linear. However, only the slopes of the relations will be ob-

' tained becauses changes in power, rather than absolute power, will be
measured, particularly at low levels. Gince thé ihstrument signals are
(ideally, at least) zero at zero power; the curves can be translated

| parallel.to themselves until they pass through the origin to produce
absolute calibrations. Once these absolute felationships have been
established, the chambers can be located to prbduce the required signal
sfrength for a given power. | _

Changes in Heater Power —The energy input to any hesater or group
of heaters can be determined from the current flow and the known re-
sistance of the heater elements. if all other conditions are held
constant, any decrease in heater power input must be accompanied by an
éqfiivalent increase 1in nuclear power to maintain steady temperatures.
Thus, the chenge in heater input 1s a direct measure of the change in
nuclear power.: Various amounts of electric heat, up to about 200 kw,
will be shut off to obtain data. (The limit of 200 kw is imposed on
_.this method by the fact that this is the normel heater requirement for
steady-state operation at zero power.) |
Measurements involving only changes in heater power require that
'the system temperature remain constant with time. If'this is not the
case; corrections based on the rate of change of temperature and the
system heat capacity will be applied. The direct correspondence between
changes in heater power and nuclear power is valid only if the heat
" losses from the circulating loops are independent of the heater status,
This independence is expected to be an adequate epproximation for the
-early callbrations. o

Changes in System Temperature —If the heat removal from (or addition
to) the circulating loops remains constant in time, any variation in. the
~ rate of change of system temperasture is related to & change in nuclear
power through the heat capacity of the circulating systems. A serieé of
arbitrary changes in nuclear power will be made and the rates of change
'of system temperature will be observed to ¢btain power calibration data.

Significant variations in the rate of change of temperature will result
4-6

Trom power changes of 100 kw or more. These data will be used in con-
junction with the other measurements to establish the instrument calibra-
tions. It is anticipated that adequate data will be obtained with no
more than 20°F change in system temperature so that the system heat
losses will not change significantly.

Heat Balances — Heat balances will be calculated at all power levels
(including zero power) to determine absolute powers for comparison with
other measurements. These calculations will give the absolute nuclear
power when the system is &t steady state. However, since any heat
balance contains some errors which are not power dependent, the accuracy
of this method of power measurement will improve with Increasing power.

It is expected that the accuracy of the heat balance will become comparable
to that of the other methods of power measurement at 0.5 to 1 Mw.
L.,2.2.2 Position Correlations

Since the safety chambers and the linear-power chambers will remain
in fixed positions during normal operation, it will be necessary only to
locate the position for each chamber that gives the desired ratio of
output current to nuclear power. This is the product of the ratio of
the chamber current to neutron flux and the ratio of neutron flux to
nuclear power. The latter ratio is a function of position and is
practically constant at any position. On the cother hand, the current/flux
ratio will probably not be linear over the entire range of fluxes to be
encountered in reactor operation. Therefore, these chambers will be
positioned to give maximum fidelity in the normal power range, 1 to 10 Mw.
Measurements will be made to determine the extent of any deviations from
linearity at lower powers.

The principle of operation of the wide-range counting channels
derives from the approximately exponential decrease in neutron flux along
the instrument shaft. Because the flux/power ratio does not decrease
perfectly exponentially, an electronic function generator is included
in each wide-range counting channel to produce a function of chamber po-
sition that is linearly related to the log of the flux. The initial
gettings of parameters in the function generators will be based on calcu-

lations, but additional adjustments will be required after the actual
b7

relations between flux and chamber position in the reactor installation
have been determined. These relations will be determined at constant
power by measuring count rates on each of the fission chambers as func-
tions of chamber position. Two or more power levels differing by about
three orders of magnitude may be required to cover the entire range of
travel of the fission chambers.

Since the fission chambers will be moved during normal operation,
any flux perturbations (which may be caused by other chambers) through
which the fission chambers must move must also be compensated for by the
function generators. If the other chambers produce significant effects
along the paths of the fission chambers, the function generators will be
readjusted as necessary each time the other chambers are moved. 1In
this same connection efforts will be made to identify any effects on
other chamber signals as the fission chambers move past them.

L.2.2.3 OQOther Correlations_

It is not anticipated that conditions in the reactor, such as
control-rod configuration or system temperature within the normal opera-
ting range, will signhificantly affect the relations between nuclear
power and instrument signal. However, the data will be analyzed for
evidence of any relations that may exist.

The degree of compensation of the compensated ion chambers in the
linear power channels can be adjusted if necessary. The data taken
when the power is lowered and then raised (changing the ratio of gammas
to neutrons) will be analyzed to determine the need for such adjustments.
h,2.3 Power Coefficient of Reactivity

If the reactor outlet temperature is held constant as the power is
raised, the temperature distributions in the core result in effective,
or nuclear-average, temperatures in both the fuel and graphite which
differ from the isothermal temperature of the zero-power system. The
net reactivity effect of these changes in temperature is such that the
control rods must be withdrawn slightly as the power is raised in order
to maintain the desired outlet temperature. This reactivity effect is
expected to be approximately linear with power level and will be described

in system analyses in terms of a power coefficient of reactivity.
4-8

The most precise measurements of the power coefficient of reac- -
tivity will be made later in the program when rapid power changes of
several megawatts can be imposed on the system. At 1 Mw, the effect of
the power coeffilclent of reactivity is expected to approach the lower
limit of detection capability, so some preliminary measurements will
be made. The calculated power coefficient for constant outlet tempera-
ture is —0.006 (% 6k/k)fiw and the maximum differential worth of a single
control rod is 0.08 (% Bk/k)én., implying a chenge in critical rod
position of only about 0.1 in. between zero power and 1 Mw. Therefore,
measurements of this parameter at 1 Mw will be quite crude, but wide
deviations from the expected value should be detectable.

‘The achievement of steady-state temperatures is much more rapid
than other effects of power operation (such as buildup of xenon and
other fission products or fuel burnup). There the power coefficient
can be determined from short-term changes in criticel control-rod posi-
tion associated with changes in power. These changes will be converted
to reactivity with the aid of control-rod calibration data.

L.2.4 Xenon Poisoning

The Xenon-135 poisoning effect is of particular interest in this
reactor because of the presence of the unclad graphite moderator. Infor-
mation is required about the distribution of 125%e in the primary system
as well aé the total poisoning; part of the xenon that contributes to the
poisoning will be circulating with the fuel salt, while the remainder
is absorbed in the graphite} The xenon concentration in the fuel salt
depends strongly on the efficiency of the stripping mechanism in the
pump bowl. This, in turn; is influenced by circulating gas bubbles in
the fuel system and the lével of the salt in the pump bowl. The xenon
concentration in the graphite depends on the mass-transfer coefficiefit for
xenon between the fuel salt and the graphite, the concentration in the i
circulating salt, and the diffusion of xenon through graphite.

As described in section 2.3.1.3 (Krypton Stripping), experiments
were performed to measure the stripping efficiency in the pump bowl and
the mass transfer coefficient for gas transport from the graphite to the

circulating salt using ®5Kr. These results were used to estimate the -
k-9

behavior of xenon in the reactor system; i.e. theIStripping of xenon in
- the pump bowl and the mass transfer of this gas-from-the salt to the
graphite. On this basis, the xenon poisoning is expected to be of the
order of o.Oosnto 0.05 (% 8k/k)/Mw (compared to ~ 0.2 (% 8k/k)/Mw for a
stationary-fuel reactor with the same core composition). This means
that during the low-power tests, the reactivity effects of xenon may be
too small for significant measurement. Névertheless, the reactivity
effects 6f changes in power will be analyzed for evidence of xenon
polisoning. - |

Tn addition to the direct observation of reactivity effects, the
determination of the 136ye — 134%e patio in the fuel system offgas gives
information on the poisoning by l?SXeQ The provisions and the tech-
nigues for sampling the offgas and analyzing for the xenon isotopic
ratios will first be put to use during the low-power operation.
4,2.5 On-Line Analysis of Operation

An important function of the on-line computer will be the reduction
and analysis, on a real-time or current basis, of reactor data as they
are accumulatéd from the operating system. These operations will be
performed in addition to, and concurrent with, the routine signal-monitor-
ing and data-acquisition functions of the computer.

The operation of the computer and the mechanical performance of
the various computafiions in the programs will have been checked out
during earlier phases of the program. (See 2.3.12.2). However, many of
the programs are designed to evaluate the performance of the reactor
system during power operation. The mathematical treatments and the
values of paremeters in the initial versions of these programs are based
partly on theoretical considerations and partly on.empirical results of
earlier operations. Therefore, the adequacy with which these calcula-
tions describe the operating reactor can he determined only under
operating conditions. It is expected that operation at power levels
approaching 1 Mw will provide the first opportunity to check out the
calculations that will be used later to monitor reactor behavior. (It
is quite likely that full-power operation for some time will be required

before the final versions of some programs are established.)
L-10

4.2.5.1 Reactivity Balance
The ultimate function of the reactivity balance is to reveal any
deviations or anomalies in the reactivity behavior of the critical re-
actor. Such balances will be calculated every 5 minutes (and on demand)
and excessgive deviations will be called to the attention of the operators.
Under normal circumstances, the results of the reactivity balance are
printed out once an hour and all other results are stored on magnetic
tape.
The reactivity balance sums all the reactivity changes, both posi-
tive and negative, from a reference condition and compares the result
to the expected result, namely zero. The reference condition for the
MSRE is the Jjust-critical, clean, zero-power reactor at 1200°F with all
control rods fully withdrawn. With this basis, the only positive re-
activity term (provided the reactor outlet temperature is lZOOOE or greater)
is that due to excess uranium concentration. The program computes the
sum of all uranium additions after the achievement of initial criticality
and corrects this for burnup to obtain the current concentration. The
following negative contributions to the reactivity balance are considered:
1. power effect, computed from power level and the empirically
determined power coefficient of reactivity,
2. temperature effect, computed from the reactor outlet temperature
and the isothermal temperature coefficient of reactivity,
3. control-rod poisoning, computed from rod-position data and empirical
calibrations,
L. xenon poisoning, computed from the power history of the reactor,
5. samarium poisoning, computed from the power history, and
6. poisoning due to other fission products, computed from the power
history.
The only parameters that will be known with any degree of certainty
at the start of low-power operation are the temperature coefficient and
the control-rod worth. All the other parameters and, in the cases of

Xe and Sm, the models for describing transient behavior must be verified
4-11

or adjusted on the basis of operating experience. If there is nc other
evidence of anomalous behavior at powers up to 1 Mw, the reactivity
balance will be adjusted to give zero net reactivity for all conditions.
Since most of the reactivity effects will be small at 1 Mw, the adjust-
ments made during this phase of operation will certainly have to be
refined when higher powers are achieved.
L.2.5.2 Heat_ Balance

A heat balance will be calculated and printed out by the computer
every 4 hours (and on demand). It is expected that heat-balance calcu-
lations at zero power will have resulted in a good value for unmeasurable
heat-loss terms (see 2.3.22.4). This term will then be used in heat
balances at power to evaluate the net nuclear power. The heat-balance
result will be compared with manual calculations and other power cali-
brations to demonstrate its adequacy. Term-by-term comparisons will
permit modification of the computer program in any areas in which it is
inadequate.
4.2.5.3 Salt Inventory

Inventories of fuel, flush, and coolant salts will be calculated
every 8 hours (and on demand). The purpose of these calculations is to
reveal any changes 1in inventory which may indicate salt losses. There-
fore, it is essential that the calculations properly account for the
changes in bulk-average temperature in the non-isothermel loops when
the reactor 1s at power. The adequacy of these calculations will be
checked under no-loss conditions (which can be verified by returning
to the isothermal state) and modified to give the required results.
L.2,5.4 Qther On-ILine Calculations

The computer will also perform a number of other calculations whose
results may depend on the reactor power level. These include:
1. tabulation of the number of power-induced thermal cycles on the

fuel pump tank,
2. the temperature difference between the fuel inlet and the lower

head of the reactor vessel,
3. the temperature difference between the fuel inlet and the core-

support flange,
y-12

L. celli-air average temperatures, and
5. nuclear average temperatures of the fuel and graphite.

All these calculations will be examined at low power, first to
determine the adequacy with which they reflect actual occurrences, and
second to determine their potential value in revealing operating
anomalies.

In addition to the reduction of current data in real time, the
computer has the capability for retrieving and processing previously
recorded information. This work, including preparation of the necessary
programs, can be carried out while the computer is on-line without
interfering with the on-line functions. (The possibility also exists
for adding tothe on-line functions of the machine.) Since the nature
of the data processing that will be required depends on the operating
experience with the reactor, it is not possible to specify the calcu-
lations that will be performed. In cases where advance specification
was possible, the calculations were included in the on-line functions.

L.2.6 Establishment of Bageline for Chemical Analyses

 

During the low-power operation the program of sampling and analyzing
the fuel salt, started during the zero-power experiments, will be ex-
tended and broadened. At all power levels, information concerning the
intrinsic stability of the fuel as well as that vis-a-vis INOR-8,
graphite, and fission products, will be furnished by the results of
chemical analyses. One important goal, therefore, will be to prepare for
the study of power effects by establishing with maximum confidence the
concentrations of those constituents which will be of interest during
later operation at higher power.

A fuel salt sample will be taken routinely once a day for analysis
in the High Radiation Level Analytical Facility (HRLAF). Analyses will
include the primary salt constituents, corrosion products, oxygen, and
the reducing powver of the sait.

Growth of fission products into the fuel during reactor operation
will be followed by measuring the characteristic activity spectra of salt

specimens. This type of analysis will be instituted during the
413

low-power operation and the results.will be related to calculations of
burnup and fission product growth.

The operation at low power will also be used to further evaluate
and prove all of the other techniques and procedures involved in hand-
ling radicactive samples. |
4.2.7 Intermediate Dynamics Studies

Dynamic tests will be made &t each power level during low power
testing. These tests will use control-rod perturbations to give reac-
tivity pulses and pseudo-random binary reactlvity Inputs. The procedure
will be the same as for the zero-power tests discussed in 3.2.3.

The cbserved transient response will be compared with the calcu-
lated transient response. Methods are being developed to automatically
adjust the persmeters in the theoretical model to agree with the experi-
mental results.

The transient response results will be Fourier analyzed to give the
frequency feéponse. The frequency response as well as the transient
response will be used in the automatic parameter‘addustment routine,

The effect of reactivity feedback due to temperature changes will begin
to appear at these low power levels. Analysis of the frequency response
data will give some indication of the validity of the feedback model
used in the theoretical calculations. |

The inherent fluctuations in the flux level will be analyzed and the
spectral density of these fluctuations will be determined. The spectral
density gives hhe product of the square of the system transfer function
and the spectral density of the input, Since the system transfer function
will be known from the pulse tests and pseudo-random binary input tests,
it may‘be”possible to isolate the spectral density of the input. This
will help in determining the cause of the fluctuations.

These tasts will indicate the stability performence of the gystem at
8 particulaf:power level. Also, the trends in the stabllity performance
with power level will be determined by ccmparison with previous results.
These results, along with theoretical results using a model updated with
latest information, will be used to predict system stability at the next

power level before the power is changed.
SECTION 5
REACTOR CAPABTLITY INVESTIGATIONS -
APPRCACH TO FULL POWER

5,1 OBTECTIVES

This phase of the test program is the logical extension of the
operation at low powers. The primary objective is to raise the power
level, in steps, to the design power of 10 Mw.

The principal difference between the operation in this phase and
earlier operations is in the mode of power control. The locad on the
reactor will be established by adjusting the rate of heat extraction at
the radiator while both the nuclear power and the primary-system tempera-
ture are controlled automatically.

- Although the primary objective is to increase the power, the ob-
jective of each of the individual tests is to determine whether or not
there is any aspect of the operation that might restrict the attainment
of the primary objective. Therefore, experiments will be performed at
each power step to establish the mechanical, thermal, nuclear, and
chemical performance of the system. This performance will be analyzed
in the light of theoretical and design predictions, observations at
lower powers, and extrapolations of earlier results. Any unexpected or
anomalous behavior that may be observed will be resolved before the next
power increase. If any limitations are approached that can be relieved
(for example, by improving shielding, cooling, or heating), the necessary
corrective measures will be taken at the same time.

In addition to permitting evaluation of the reactor system, opera-
tion at intermediate powers will permit the refinement of calculational
models, techniques, and parameters to be used in predicting the behavior
at successively higher powers. Since a number of factors, and hence the
accuracy with which they can be evaluated, depend on power, time at
power, or just time, it is expected that improvements in the ability to
evaluate the system will continue throughout the operation.

The anticipated steps in the approach to 10 Mw are: 1.5, 3.0, 5.0,

and 7.5 Mw. Before the power is raised above 5 Mw, the reactor will be
5-2

operated. at that power long enough tc discern any short-term effects on
the fuel salt. This will require abcut 15 days from the beginning of
5 Mw operation until the power is raised to 7.5 Mw. At 1.5, 3.0, and
7.5 Mw about 5 days will elapse between power 1acreasas. These times
are, of course, based on the assumption of nc difficulties or anomalous

behavior to be investigated. If such appear, the schedule will be re-

vised as necessary.

5.2 PROCEDURES

5.2.1 Performance cof Control Systems

The reactor power level, above 1 Mw, is determined by the heat re-
moval rate at the radiator which is set by a combiznaticn of the
radiator-door and bypass-damper positions, and the number of main blowers
operating. The autcmatic temperature-contreol servo then adjusts the
reactor power To meet this load demand while maintaining the reactor
outlet temperature constant at some preselected value.

The contrcller has been tested and the required adjustments have
been made using an analog ccmputer. The test program, therefore, con-
sists of insuring that the coatrol system will maintain the proper con-
trol of the reactor power and ocutlet temperature at various power levels,
and that the control system has adequate response to cover the normal
transients which may occur.

With the system operating cn temperature servo and with the tempera-
ture set point at 1200°F, the system will be cperated at a constant load
demard &t the radiator for a period of time sufficient to determine 1if
the cutlet temperature remains constant or if there is a long-term
temperature drift. The system will also be cbserved for outlet-temperature
or flux cycling and for excessive control-rod "hunting'. The steady-
state operating teste will be conducted at each power level. Steady-
state operation will also be reviewed to determine if the small changes
in radiator load demend resulting from changes in ambient alr temperature

can be detected by changes in the neutron flux level.
2-3

A series of tests will also be run at the same power levels to
determine the control system performance under transient conditions.

The control ac%ion and the reactor system response will be observed
during transients introduced in the following manner to determine if the
transients are satisfactorily controlled and to determine if the "computed-
flux-demand" limitation of 1/2 to 11 Mw is adequate. The temperature set
point will be changed from 1200 to 1225°F at the normal motor-driven rate
of 5°F per minute and held there until the reactor system reaches steady
state. The set point will then be returned to 1200°F. A rapid load-
demand change of about 2 Mw will be made at the radiator by changing

door position or by changing the bypass-damper position. With the re-
actor operating at steady-state conditions under manval control, the
system will be switched to temperature servo with both a positive and a
negative 5°F differential between the reactor outlet temperature and the
set-point temperature. The test will then be repeated using a 25°F
temperature differential.

Adjfistments will be made on the controller to correct any areas of
poor performance which are found by the above tests, and the test program
will be repeated.

If unsatisfactory operation is found at certain power levels which
cannot be corrected by adjustment of the controller, further tests will

be performed to define the range of satisfactory performance.

5.2.2 Shielding and Containment Adequacy

The shielding surveys that were carried out at low power (4.2.2)
will be continued at all power levels up through 10 Mw. Particular
attention will be given to the two areas mentioned in 4.2.1; the north
electric-service area and the south-west side of the facility building.
Continued intense surveillience will be carried out for the purpose of
locating points at which dose rates are higher than expected.

As reactor power is increased the radiation source is increased
proportionally because of the fission process. In'addition, operation
for long periocds of time will result in an increase in the radiation

source because of fission-product buildup in the salt. Gince both of
5=k

these factors affect the dose rate outside of the biological shield, it
may be difficult to extrapclate dose-rste mezzurements taken with clean
salt at low power (<1 Mw) to the fission-prodact-contamigated salt at
high power. Therefore, careful dose~rate measurements will be made in
all areas to verify the extrapolations that were made from the low-power
measurements. As the reactor operates, accumilating fission products,
the charcoal beds will become a progressively greater radiation source
which will produce measurable dose rates only after the resctor has
operated for some period of time. Continued monitoring of those beds
will be carried out in the approach to full power.

For the reasons outlined in the previous paragraphs, the sampler-
enricher will come under close scrutiny as power is increased. Dose
rates during the sampling operation may iacrease with {ime and power
level; this will be monitored.

Continued monitoring of the contaimment in-leakage and of the build-

ing ventilation system will be carried on as cutlined in L4.2.1.

5.2.3 Calibration of Power Instruments

Calibration measurements on all the nuclear power instruments will
be continued throughout the approach to full power. The purpose cf
these measurements is to determine the final adjustments in chamber
position required to produce the optimum correlation between indicated
power and actual thermal power. Measurements will alsc be made to
demonstrate the adequacy of the wide-range countiang channels over the
entire power range.

The data toc be obtained is egserntially the same as that cutiined
in 4.2.2. However, the exclusive calibration standard in this phase cf
the operation will be the system heat balance. Therefore, the heat
balances will be carefully checked tc ensure that no avoldable errors
are introduced, either in the data cr the calculaticus.

The largest single term in the heat balance is the heat removal
by the coolarnt salt. This term accounts for more than 95% cf the tctal
nuclear power, A small, specilal-purpose, anslog device is Installed

to continuously compute and record the ccolant-sslt heat removal from
5-5

the measured salt flow rate and temperature drop at the radiator.
Operation at powers above 1 Mw will enable us to check the output

of this device against the heat balance and to adjust its fixed para-
meters to produce agreement between it and the standard. Ancther check
on the heat removal at the radiator is the heat absorbed by the air
flowing through the radiator enclosure. Since the coolant-alr stack
will be calibrated, the product of air flow and temperature rise (coupled
with the heat capacity of the air) will be compared with the other

calculations.

5.2.4 Xenon Poisoning

The behavior of Xenon-135 will be studied during the approach to
full power by observation of reactivity effects and by isotopic analysis
of the xenon in the offgas.

Reactivity changes will be observed after the power is stepped up
or down between a few kilowatts and several megawatts. The steady-state
change in reactivity will be corrected for the temperature effects (which
occur more rapidly than the xenon effects) to get the net poisoning
effect of the '®Xe. The xenon poisoning transients will be analyzed
to determine the best mathematical representation. If this is a signifi-
cant improvement over the xenon computation programmed in the computer,
the computer program will be modified.

The ratio of 13%Xe to 12%Xe will be determined in samples of the
offgas. Comparison with the fission yields of these isotopes will show
the increase in 2®Xe due to captures in 17°Xe. Steady-state informetion
will be compared with the net reactivity effect observed independently.
Offgas samples during transients may give information on transfer between
the salt and graphite. Ratios of radioisotopes of xenon (and of krypton)
will give a measure of the "age" of the gases or how quickly they are

removed from the resactor.

5.2.5 On-Line Analysis of Operation
The on-line, computerized analysis of the reactor operaticn that was
started at low power (see Sec. L4.2.5) will be continued throughout this

and all subsequent phases of the program. The principal purpose during
5-6

the early part of the approach to full power is the refinement of the
calculations and the establishment, if possible, of their final forms.
However, since the possibility of anomalous or undesired behavior of the
reactor exists at all power levels, any differences between observed
reactivity behavior and that predicted by the computer program will be
examined with great care. If such disagreements occur and can definitely
be attributed to inadequacies in the calculations of known reactivity
factors, the calculations will be modified or compensated for before the
next power increase. If, on the other hand, the evidence is inconclusive
or points to anomalous behavior of the reactor system itself, the experi-
mental operation will aim at resolving the anomaly before the power is

raised further.

5.2.6 Thermal Effects of Power Operation

As the reactor power is increased from a low level to the design
power of 10 Mw, significant changes in temperature will oeccur throughout
the system. These temperature changes should follow the predictable
changes in fluid temperature which result from heat generation in the
fuel in the reactor vessel, heat transfer to the coolant system in the
heat exchanger, and heat removal from the coolant salt in the radiator.
However, if abnormal conditions develop, certain fuel-system temperatures,
especially in the reactor vessel, may deviate from the expected trends.

The increase in power from 1 Mw to 10 Mw will be made in several
increments, and a complete set of thermocouple readings will be taken
at each power level after the system has reached steady-state conditions.
The temperatures will be compared with the values from the previous power
level, and the change in temperature will be compared with the expected
change. Particular attention will be pald to areas where solids might
possibly accumulate (if they form). The core-support ring and the lower
head of the reactor vessel are two such regions; these have sufficient
thermocouples attached to the walls so that overheating caused by solids
deposition should be detectable. If any excessively high temperatures

are encountered or if any trends are observed which would lead to
2=

excessively high temperatures at full power, the reactor power level will
be held at/or below that particular power until a more detailed investi-
gation of the specific problem can be completed.

One of the areas most sensitive to the normal thermal effects of
power operation is the fuel-pump tank. Detailed temperature-distribution
and thermal-stress calculations have been completed for zero and full
power operation and a cooling air flow has been selected to maintain
the thermal stress within acceptable limits. The pump-tank temperatures
will be observed at each power level, and adjustments will be made on the
cooling-air flow rate as required.

5.2.7 Capability and Performance of Heat Transfer Systems

1t is the purpose of this program to determine whether or not any
heat removal system will limit the power level or in any other way cur-
tail the success of this experiment. The primary heat exchanger (fuel
to coolant) and the radiator (coolant to air) are the most important heat
transfer gystems in the plant. The proper performance of the reactor
depends directly upon these two components. Because of this, careful
evaluation of the initial heat transfer capability and long term per-
formance are planned. Evaluation of some auxiliary systems is also
planned.
5.2.7.1 Primary Heat Bxchanger

Use of the on-line computer allows rapid accumulation of data from
which the value for the overall heat transfer coefficient can be calcu-
lated. The calculation will utilize a large number of thermocouple
readings on the fuel and coolant piping at several different power levels
to reduce the effects of individual thermocouple bias which would prevent
meaningful calculation of heat transfer coefficient from a single set
of readings. This program will be continued throughout the operating
life of the MSRE. The data taken during the initial approach to power
will be used to calculate a value for the overall coefficient which will
indicate the performance of the heat exchanger at the outset and will

provide base-line data for the program discussed in 6.2.k.
5-8

5.2.7.2 The Radiator_

Following the changes in heat transfer characteristics that might
occur in the radiator will be more difficult than with the primary heat
exchanger. At various power coaditions, the heat transfer area and the
alr velocity across the radiator will change, making it impractical to
compute the cverali heat traasfer coefficient. OSince this is true, a
value for the product of the overall heat transfer coefficient and the
surface area will be used for analysis. This value will be computed at
various times with radiator door position (which gcverns the heat transfer
area) and air velocity across the tubes specified. It is expected that
the most meaningful data will be taken at/or near maximum power because
under these conditions the overall heat transfer coefficlent can e
calculated. The on-line computer will be used in the analysis of the
radiator data.
5.2.7.3 Auxiliary Coolers

Cocling for in-cell components is provided by three auxiliary
cooling systems.

1. The componernt-cooling-air system provides cooling for the fuel pump
bowl and for freeze valves in the reactor and drain tank cells.

2. The cell air coolers will be expected tc maintain cell ampbient
temperature below 150°F at all pcower levels.

3. The treated water cooler provides a heat sink for several in-cell
components including the thermal shield, cell air coclers, ccmponent-
cooling-air cooler, and cther smaller items.

Iinlet and outlet temperatures aund flow rates to the cecoliers wiil be
measured and this together with a value for the heat transfer surface
area will be used to evaluate the cverall heat transfer coefficient.

It is planned to cbtain data for initial evaluation of the per-
formance of these coolers to prouvide a basils for future surveillaance.
5.2.8 Chemical Effects of Power Operation

The evidence of the in-pils testing programs indicates that the
thermal and radiation environment attesding power operation of the MSRE
will have nc deletericus effect on the fueli salt or the compatibility
of the salt, graphite, and TNUE-8. Nevertheless, the fuel salt will

be sampled frequently durirng the approach to full power and the analytical
~

5-9

results will be studied and tested statistically to determine if there
are discernible effects of power level or integrated power.
5:.2.9 Dynamics‘Studies

The tests made during low-power operation (see 4.2.7) will be con-

 

tinued throughout the approach to full power. As in the low-power tests,
these tests will serve to determine system stability and to furnish
information on system parameters and the validity of calculational methods.

The power level during this phase will be high enough that the
effect of power level on system temperatures, and thus reactivity, will
become important. Analysis of these tests will furnish an improved
characterization of the feedback effects in the theoretical model.

The increased power level will also make it possible to introduce
system perturbations by load changes as well as by control-rod positioning.
This will be accomplished by adjustment of position of the radiator doors
and the observed temperature transients will be compared with predictions.
The procedure for automatic adjustment of appropriate system parameters
to cause agreement between theoretical and experimental results will be
used. This can be used to give direct information on the feedback trans-

fer function.
SECTION 6
SYSTEM CAPABITITY INVESTIGATIONS — EXTENDED OPERATION

6.1 OBJECTIVES

Once the operability and stability of the reactor éystem have been
demonstrated at full power, the next logical step is sustained operation
at high power levels. The general objectives of this phase of the opera-
tion are to demonstrate the durability of the reactor system and to permit
investigation of any effects of long-term operation.

Samples of the fuel and coolant salts will be removed at regular
- intervals for detailed analysis. These will be used to study the behavior
of the major constituents of the salts as well as the effects of fission
products, corrosion products, and any other contaminants that may be intro-
duced. Specimens of graphite and INOR-8 will be removed from the core
periodically to provide further data about the compatibility of these
materials with the salts during long-term irradiation. These specimens
will also aid in projecting the operating life of the system.

Many of the components in use at the MSRE were developed specifically
for this application and are similar to components that would be used on
larger molten-salt reactors. Therefore, component performance will be
carefully monitored and evaluated to provide reference data for continulng
| development.

An important aspect of the long-term operation of the reactor system
is the ability to perform maintenance on the equipment. We expect that
component failures will occur and that remote maintenance techniques will
be required to repalr or replace the failed items. Demonstration that
remote repalrs can be made without excessive expenditures of time and
effort 1s a necessary part of the overall feasibility demonstration.

The fuel to be used in the initial operation of the MSRE contains
235 as the fissionable material with some diluent Z38U. However, the
salts for breeder reactors will contain 233U and thorium. Therefore, con-
sideration will be given to operating the MSRE with a fuel mixture con-
taining 2337 and Th to demonstrate the compatibility of these constituents
and to provide develcpment informaticon. Operation with the modified fuel

mixture would follow successful operation with the initial fuel charge.
6-2

6.2 PROCEDURES

6.2.1 Fuel Chemistry

A substantial amount of information about the behavior of molten
fluoride-salt mixtures in an enviromment like that of the MSRE has bazn
obtained from out-cof-pile loop experimerts and frem short-term, high-
intensity in-pile tests. All of this informaticn indicates that the
chemistry of the MSRE salt mixtures 1s satisfactory for loag-term copera-
tion. The extended cperation of the reactor will provide an cpportunity
to study the effects, if any, of irradiaticn exposure for loang pericds of
time. It will also test the adequacy of the operating procedures with
regard to keeping the salts free from external contaminants.

Detailed chemical analyses will be performed on all the salt samples
that are obtained. Analytical informaticn will be scught in the followlirg
areas:

1. concentrations of major constituents {(Li, Be, Zr, U, Fl,

2. concentrations of fission products,

3. concentrations of corrosion products,

4. oxide contamination,

5. contamination by other foreign species, and

6. reducing power of the melt.

We do not anticipate that the fuel chemistry will impose any operating
iimitations on the reactor but all the results will be closely examined
for any unexpected behavior.

6.2.2 Materials Compatability

The demenstration of suitable compatapility of the fuel salt with
graphite and with INCR-8 is one of the more important chjectives of the
MSRE. It is alsc desirable to determine the fission anpd corrosicon product
distribution within the reactor system and tc determine any changes which
occur in the mechanical properties of the graphite cr INCR-8 as a result of
the long-term exposure to tne fuel salt and te a high aneutron fiux.

A sample assembly containiag INOR-8 and graphite specimens is in-
stalled near the vertical center line of the reactor core., The ITNOR-8 and
graphite samples extend the fu1ll length of the cors so that the various

specimens will be exposed to different levels of neutron flux. A pertion
6-3

of the samples will be removed at various times throughout the entire life
of the reactor, with the first samples being removed after a period of
six months to a year. New specimens will be installed to replace those
that are removed.

A series of control samples, identical to the INOR-8 and graphite
samples in the reactor core, will be exposed to fuel salt in the absence
of radiation. The control samples will be subjected to the reactor thermal
and salt—exposure history concurrently with the reactor. Control samples
will be removed and examined on the same schedule as the reactor samples.
6.2.2.1  Graphite

The graphite samples will be examined for shrinkage effects, any
tendency toward salt permeation, changes in physical or mechanical proper-
ties, and for deposition or absorption of fission or corrosion products.

Carefully machined, measured, and weighed specimens are included from
both the axial and transverse direction of a graphite stringer. These
will be checked for weight gain or loss and for dimensional changes.
Changes in the mechanical properties will be evaluated by bend tests on
premachined specimens. Electrical conductivity measurements will also
be made to indicate changes in the physical properties.

Metallographic examinations will be made to determine if there is
any evidence of salt penetration into the graphite or if there is any
surface deposition of fission or corrosion products. Autoradiography
willl also be used to 1indicate if salt penetration has occurred.

Sample drillings will be analyzed spectographically and chemically to
determine the identity and quantity of any fission products which may be
present either as a surface deposit or as a result of diffusion into the
graphite,
6.2.2.2 INOR-8

There are six INOR-8 rods in the sample assembly, and each rod has
been machined to form 32 tensile specimens when cut at the proper loca-
tions. A portion of these specimens will be tensile tested at various
temperatures to determine the ultimate and yield strengths and the duc-
tility. Creep tests at 1250°F will be run on the other specimens to deter-

mine the creep and stress-rupture properties and to determine the creep
6l

duetility. There are four specimens in the sample assembly made from weld-
deposited metal which will also be used in tensile and creep tests.

The specimens will be examined metallographically to determine if
any changes in microstructure have occurred and to evaluate any evidence
of corrogsion by subsurface leaching. The specimens will also be examined
for fission-product deposition by spectographic and chemical analysis.
6.2.2.3 Fuel and Coolant_Salt

The fuel and coolant salt samples will be chemically analyzed for the
build-up corrosion products and oxides. The fuel samples will alsc be
examined by gamma spectroscopy for the build-up of certain fission products.
Chemical separations are possible for various fission products, but thils
type of analysis will not be done uniess a specific need develops.

6.2.2.%  Fuel System Offgas_

The helium offgas stream from the fuel system may contain a variety
of contaminants ranging from the rare-gas fission products (xenons and
kryptons) to decomposition products of the lubricating oil used in the
fuel circulating pump. Facilities will be provided for removing samples
of this gas both at the operating concentration and after concentrating
the contaminants on molecular sieve material. The samples will be sub-
Jjected to mass- and gamma-spectrographic analyses to identify the contami-
nants and measure their concentrations. Measurements of the 124Xe to 12%Xe
ratio in the offgas will provide independent data on the °>Xe poisoning
in the reactor as well as some information on the dynamice of rare-gas
stripping. The analyses may alsc provide information about other volatile
fission products. The exact nature of the information that is obtained
will depend on the identity of the species that appear in the offgas.

If a sufficient quantity of fission products 1s deposited on the gas-
exposed surfaces of the fuel pump, their presence could be detected by an
increase in pump-tank temperature during shutdown periods.

The equilibrium temperatures of the upper pump-tank surface will be
recorded with the coocling air off and with the system heated and drained
during each shutdown period. The temperatures after power operation will
then be compared with the original values for evidence of an additional

energy source. The fuel pump must be drained and the pump-tank helium
6-5

purged to eliminate the heating from the fission products in the gas and
in the fuel salt which would tend to mask the heating by surface deposits.
6.2.3 Changes in Dynamics

The tests made during the low-power phase (see h.2.7) and intermediate
power phase (see 5.2.9) will be repeated at the start of full-power opera-
tion.

In addition, tests will be made periodically during full-power opera-
tion. These will be primarily for the purpose of detecting any possible
unexpected changes in the dynamic characteristics. Measurement of the
noise spectrum will be used for this 1f statistically reliable information
can be obtained around the resonant frequency (~ 0.01 cycles/sec). It is
expected, however, that it will be necessary to use control-rod pulse
experiments as the primary means for these tests,

6.2.4 Performance of Components and Eguipment

The components and equipment exposed to fuel and coolant salt are
almost all of a unigue design and are constructed of a unique material
(INOR-8). Meny of the component designs have been tested in test loops
at or near reactor conditions, but there is no actual long-term reactor
operating experience of either the components or the material of construc-
tion. It is therefore important to monitor the performance of these
components to reveal any unforeseen changes in performance or incipient
mechanical failure. It is also important to monitor the performance of
some of the more conventional equipment so that the failure of these com-
ponents during important phases of reactor operation can be avoided as
nearly as possible., The areas of particular interest are outlined below.
6.2.4.1 Heat Transfer Equipment

Several components in the reactor system and in the auxiliary systems
must provide adequate heat transfer for continued operation of the reactor
at full power. The heat-removal rate from the reactor depends directly
on the heat-transfer characteristics of the primary fuel-to-coolant heat
exchanger and the cooiant-system radiator. Any loss in heat transfer,
either by the bulldup of scale, loss of flow, or by blockage of tubes,
will result in a decrease in the maximum power level at which the reactor

can be operated for a given reactor outlet temperature. (The ultimate
6-6

limitation 1s the minimum safe temperature of the coolant salt at the
radiator outlet, >900°F.) The heat-transfer characteristics of the heat
exchanger will be determined periodically by taking temperature data at
several different power levels. A computerized prccedure is available
which eliminates any constant bias in the thermocouple readings and yields
an averaged, overall heat-transfer coefficient for all the power levels.
By observing these heat-transfer coefficients over a period cof time, any
tendency toward fouling or plugging of the heat exchanger can be detected.

Since the radiator has an almost infirite variety of combinations of
air flow and tube exposure for any given power level, a true heat-transfer
coefficilent cannot be determined. However an effective coefficient can
be defined and calculated using standardized procedures. This effective
coefficient will be evaluated at several power levels using preselected
docr and bypass-damper positions, and the radiator performance over a
period of time can be compared to the original performance.

The fuel-drain-tank coolers are required to remove the fission-product
afterheat from the fuel salt after a sustained run at relatively high
power. There is not a sufficient heat source available to permit the
cooling system to be tested at full load prior to nuclear operation at
full power. However, a transient test using the heat capacity of the salt
as a heat source indicated a heat-removal capability of 140 kw, which is
more than adequate. A heat-removal test will be conducted after the re-
actor has operated for a pericd cf time at part load to determine the
steady-state heat-removal capability.

Other pieces of heat-exchange equipment which are required fcr the
continued cperation of the reactor are the reactor-, coolant-, and drain-
cell space coclers, the treated-water coocler, the cooling tower, the
component-cooling-pump oil coolers, the component-coolling-pump gas cocler,
and the lube oil coolers. The performance of this equipment will be
observed throughout the life of the reactor to insure that adequate per-
formance is maintained for the continued operation of the reactor.
6.2.4.2  Thermocouples

Thermocouples are used throughout the system as control signals, to

monitor the operating temperature of all the high-temperature components.
&=T

Thermocouple errors may be introduced by actual failure of the thermocouple,
by long-term drift caused by the high temperatures, by changes in heat-
transfer characteristics, or by nuclear-radiation effects. Some of the
thermocouples may also be influenced by local nuclear-radiation heating
which would cause them to give an apparently high reading. Although
separation of the various types of error would be difficult, overall
changes in the thermocouple performance throughout the reactor system can
be determined by statistical methods.

A complete set of readings at isothermal (zero power) conditions for
all thermocouples reading fuel- or cooclant-salt temperature will be taken,
and a statistical analysis can be made to determine the mean temperature
and the standard deviation. Similar analyses will be run during the
operating life of the reactor, and the mean temperatures and standard
deviations will be observed for changes in thermocouple performance.
Probability plots will also bhe made to determine if significant new
mechanisms occur during the operating life which cause statistical varia-
tion of the readings. The thermocouple readings will also be reviewed to
determine whether the individual thermocouples maintain their same posi-
tions relative to the mean.

Preliminary data has indicated that the thermocouples on the radiator
should be grouped and analyzed separately from those on the remainder of
the fuel and coolant systems. Additional groupings will be made as re-
quired to obtain a maximum of information. For example, it may be reason-
able to separate those in high radiation fields from those in lower fields
to assess the radiation-exposure effect.

Records will also be maintained of the actual failures and the causes
of the failures where these are known.
6.2.4.3 Heaters and Insulation

All the reactor piping and components which are exposed to either
fuel or coolant salt are heated electrically to insure that the metal
surfaces in contact with the molten salt remain above the freezing point
of the salt. Three general types of insulation are used in the system.
The main piping inside the reactor cell and the heat exchanger use remov-

able heater units with integral reflective insulation. The other components
6-8

im the reactor and drain-tank cells are enclosed in furances which are
insulated with more conventional low-conductivity insulation. Some of the
piping, particularly in the ccolant cell where direct maintenance is
feasible, hag permanently installed Calrod heaters and conventional insu-
lation.

The heat loss of the reflective insulation is expected to increase
about 10% because of changes in emissivity. The performance of the low-
conductivity insulation, and also the reflective type, may change because
of cracks and crevices developing through the insulation and through the
Joints of the component furnaces. Any increases in heat loss from either
type insulation will be determined during shutdown periods by setting all
the heaters back to their original 1200°F settings and recording the
equilibrium temperatures with the system drained. The resulting tempera-
tures will then be compared to the original reading, and any decrease in
temperature is an indication of increased heat loss. While it would be
possible to determine a quantitative number for the increase in heat loss
by setting the heaters to obtain the original temperatures, the time re-
quired to make these adjustments throughout the system is prohibitively
long. Some guantitative information about the overall performance may
also be gained from observation of the heat loads on the space coolers in
the cells,

Records will also be kept of heater or power-lead failures and of
the conditions and causes of these failures.

The fuel circulating pumps may be subject to five types of slow deteri-
oration of performance. These are (1) a slow wearing of the bearings ang
0il seals, (2) a buildup of scale in the cooling-oil passages in the
shield plug and the shaft, (3) fatigue damage due to temperature cycling,
(k) a slow plugging of the xenon-stripper jets, and (5) radiation damage
effects to the drive-motor electrical insulation and to the motor-bearing
lubricant. The coolant pump is susceptible to only the first three of
the above items. Unless a completely unprecedented corrosive or erogive
attach occurs on the hydraulic parts, a loss in hydraulic performance is

not anticipated.
6-9

The lubricating-oil packages associated with each of the main circu-
lating pumps may also be subject to a slow deterioration of performance by
a decrease in heat transfer at the cocler or by a plugging of the filter
to an extent that it cannot be cleared by standard procedures. There are
standby oil pumps with independent power supplies on each oll package so
that the electrical or mechanical failure of one pump does not require
a reactor shutdown.

The circulating-pump-motor power input and speed, the pump tempera-
tures, and the rate of seal-oil leakage will be monitored for long-term
changes in the circulating-pump performance. A thermal-cycle history of
the pumps will also be maintained to insure that the calculated safe life
of the pumps will not be exceeded and to determine if there are any prema-
ture failures.

Although all the damaging and wear processes are generally continuous,
most of them have threshold values, and the probability of detecting an
incipient failure with a significantly long advance warning is relatively
small. The wearing of the lower oil seal and the decrease in heat trans-
fer from the shield plug are possible exceptions where a trend could be
followed.
6.2.4k.5 Freeze Valves

Freeze valves are used to prevent the undesired transfer of salt from
one part of the system to another. Freeze valves 103, 204, and 206, which
hold the fuel and coolant salt in their respective systems, are also
required to melt within specified times. Each freeze valve has a series
of control modules which maintain the valve parts at the proper tempera-
tures.

The performance of the freeze valves and modules will be evaluated
by maintaining a history of the freeze and thaw cycles and a history of
the adjustments and replacements of the control modules. A history will
also be maintained of any unintentional thaws and any other malfunctions
which might occur.
6.2.4.6  Freeze Flanges_

The main fuel- and coolant-system piping inside the reactor cell is

connected with a total of five pairs freeze flanges so that the major
6-10

components can be removed and replaced. The flanges have a buffer-gas

and leak-detector system which insures that any leakage is gas leakage
into the system and also permits the measurement of the total leakage rate
of buffer gas from the flange joint.

The long-term performance of the freeze flanges will be evaluated
by maintaining a history cof the buffer-gas leakage rates and a history
of the opening and closing of the flanges. A thermal-cycle history will
also be maintained to insure that the calculated safe life is not exceeded
and to determine if there are any premature failures.
6.2.4.7 Radiator and Radiator Enclosure

The radiator and radiator enclosure provide the heat-load demand and
the load control of the reactor. The load demand is determined by the
combination of door position, bypass-damper position, and the number of
main blowers operating. The load may be set by eilther manually setting
the door and damper positions and by manually energizing the proper blowers,
or by using an automatic load-demand switch that actuates the doors, the
damper, and the blowers in a preprogrammed sequence. The radiator doors
may also be closed relatively rapidly to remove the heat load from the
reactor.

The performance of the radiator is partially covered in Sections
6.2.4,1, .2, and .3, and therefore the mechanical performance of the doors,
the door-actuating mechanism, the bypass damper, the main blowers, and the
automatic loading system will be considered in this section. The per-
formance of these components will be evaluated by periodically checking
that the doors and the damper actuate smoothly and at the prescribed rates
under both normal and scram conditions and by checking to insure that the
position indicators agree with the actual positions. The automatic
loading system will be checked to insure a relatively smooth increase and
decrease in load over the range of operation. A detalled history will be
maintained of any failures which occur and of any adjustments which are
required on any of the radiator components.
6.2.4.8 Instrumentation

A large variety of both nuclear and process instruments and con-

trols are present in the reactor and the auxiliary systems. These
6-11

instruments and controls will be calilbrated and adjusted periocdically during
shut-down periods and also at any time an instrument appears to be ex-
cessively in error.

A detailed record will be maintained on each instrument of the changes
that occur from calibration to calibration, the adjustments that are made
during the calibration, and of any actual failures which occur.
6.2.4.9 Control Rods and Drives

The muclear characteristies of the reactor are such that fast-acting
control rods were not required. A flexible design was chosen so that a
single vessel penetration could be used and the rod thimbles could be
bent to avoid interference with the graphite sampler. The rods are a
series of short, cylindrical poison elements slipped over a flexible core.
Development tests have shown that the rods may stretch during the early
life of the control-rod assemblies. A position calibration point was
built into the control-rod thimbles which will be used to pericdically
check the rods for elongation. The rod position indicators will also be
calibrated at the same time.

The flexible design with two bends prevents a free gravity rod drop
which is the common method for scramming a reactor. However, the rods
have been specified to, and do, drop with a minimum acceleration of
12 ft/sec®, Tests will be conducted throughout the reactor life to insure
that this acceleration is maintained and to observe any changes from the
original value., The rods will also be observed for erratic motion which
would be an indication of sticking or binding.

When the reactor is operated at power for sustained periods of time,
the control-rod worth will eventually begin to decrease because of the
"burnup" of the poison material. Rod-drop tests to evaluate the control-
rod worth will be conducted at appropriate intervals. A more accurate
calibration of the control-rod worth will be made, if the results of the
rod-drop tests indicate that this is needed.

The control-rod drives operate in a high radiation field inside the
biological shielding and have been specified to coperate in a field of
10° rad/hr. The drives will be observed for radiation damage effects on
the electrical insulation, the lubricants, and the switches. The drives

will also be observed for mechanical faillure or other malfunction.
6-12

— e S e meme G b e L me

The normal surveillance of the reactor vessel and core graphite is
done by the exemination of the graphite-and-INOR-8 sample assembly as
outlined in Section 6.2.,2. However, should the need arise, the 10-inch
access flange can be removed which, will allow five of the graphite
stringers to be removed and will permit a visual examination of parts of
the vessel interior using remote viewing techniques.

The capability alsc exists for making a more limited visual inspection
each time the graphite sample assembly is removed. These inspections
will be made at times when deemed necessary or desirable and not on a

scheduled basis.
29.
30.

32.
33.

35.
36.

38,
39.
4o.
b1,

Internal Distribution

MSRP Director's Office
Rm. 219, 920L4-1
. Adams
. Adamson
Affel
Alexander
Apple
Baes
. Bker
. Ball
Rarthold
F. Bauman
. K. Beall
. Bender
S. Bettis
F. Blankenship
Blumberg
. G. Bohlmann
. J. Borkowskil
. H. Burger
Cantor
. Carter
Cathers
Compere
Cook
Corbin
Crowley
Culler
Dale
Davis
Ditto
Donnelly
Doss
Engel
. Epler
Ferguson
. I'ray
Fraas
Friedman
Frye, Jr.
Gabbard
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ORNL-TM-911

. Grimes
. Grindell
. Guymon

Harley
Haubenreich
Hebert
Heddleson
Herndon
Holt

Hudson
Kasten
Kedl

. Kelly

Kennedy
Kerlin
Kerr

. Kirslis
. Knowles
. Krakoviak

Krewson
Iane

. Lindauer

Lundin
Lyon
MacPherson
MacPherson

. Martin

Mathews
MeCoy
McCurdy

. Mcbhbuffie

. McGlothlan
. McNeesge

. Meyer

Miller
Moore
triarca
Payne
Perry

. Piper

Prince
Redford
93.
k.

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98.
99-

100.

101.

102.

103.

104,

105.

106.

107.

108.

109.

137-
139-

M. Richardson
R. C. Robertson
H. C. Roller
M. W. Rosenthal
H. W. Savage
A. W. Savolainen
D. Scott
H. E. Seagren
J. H. Shaffer
M. J. Skinner
A. N. Smith
P. G. Smith
W. F. Spencer
I. Spiewak
R. C. Steffy
H. H. Stone
Jd. R. Tallackson

131. A

132. C. L.

133, T. W.

134, H. M.

135. W. L.

136. R. F.

138.

153%.

Internal Distribution

 

(continued)

110.
111.
112,
113.
11k,
115.
116.
117.
118.
119.
120.
121.
122,

123-12Lk,

125-126.

127-129,
130.

External Distribution

. Giambusso, AEC-Washington
Matthews, AEC ORO
McIntosh, AEC Washington
Roth, Division of Research and Development, AEC-ORO
Smalley, Reactor Division, AEC-ORO

Sweek, AEC Washington
Reactor Division, AEC ORO
Division of Technical Information Extension

?:C4ca:§ NAarw=sgQ

SPepREEREEQEEH

ORNL-TM-911

Thoma
Tolson
Trauger

. Ulrich

Webster
Weinberg
weir, Jr.
West
Whatley
Whitman
White
Wills
Wilson

Central Research Library
Document Reference Section
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