ORNL-1234.txt

 

ORNL 1234

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OAK RIDGE NATIONAL LABORATORY
OPERATED BY

CARBIDE AND CARBON CHEMICALS COMF’ANY

y
:/ A DIVISION OF UNION CARBIDE AND CARBON CORPORATION
L OCAK RIDGE, IENNESSEE
N5 :JP#

  

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* Index No. ORNL-1234
| This document contains 138 pages.
This is copy f,’g of 216, Series A.

 

Subject Category: Reactors-Research
and Power

REACTOR PROGRAM

of the

AIRCRAFT NUCLEAR PROPULSION PROJECT

by

Members of the -
Aircraft Nuclear Propulsion Project

Wm. B. Cottrell, Editor

June 2, 1952

  

OAK RIDGE NATIONAL LABORATORY
Operated by
CARBIDE AND CARBON CHEMICALS COMPANY
A Division of Union Carbide and Carbon Corporation
Oak Ridge, Tennessee

Contract No. W-7405-eng-26

o, i

3 4455 03L058Y L

LY

 

 

 
Internal Distributiol:

 
 

1. G. M. Adams§n
2. C. J. Barton

3. E. S. Bettis

4. D. S. Billingto
5. F. F. Blankensl@p
6. E. P. Blizard

7. A. Brasunas

8. R. C. Briant

9. F. R. Bruce

10. J. H. Buck

11. T. H. J. Burnett
12. A. D. Callihan
13. C. E. Center
14, G. H. Clewett
15, C. E. Clifford
16. W. B. Cottrell
17. W. K. Eister

18. L. B. Emlet

19. W. K. Ergen

20. G. T. Felbeck
21. A. P. Fraas

22. W. R. Grimes J
23. L. F. Hemphillf
24. A. Hollaender f
25. W. B. Humegf
26. R. J. Jonegj

27. G. W. Keilibliz
28. C. P. Keiy

29. F. Kertesj

30. J. A. Lghe

31. C. E. Lfirson
32. R. 8. JNivingston

Issuing Offfte
Technical
Date Issued: J{}i 7 - 857

ormation Departmen

 

 

 

 

Index No. ORNL-1234
Reactors-Research and

Power

33. R a/N. Lyon
34. " D. Manly
35. F. B. McDonald
36. fil. L. Meem
37. ‘A, J. Miller
38. C. B. Mills
39. K. Z. Morgan
40, E. J. Murphy
4 H. F. Poppendiek
4 H. W. Savage

. R. W. Schroeder
44. E. D. Shipley
45. O. Sisman
46. L. P. Smith
47. A. H. Snell
48. F. L. Steahly
49. C. D. Susano
50. J. A. Swartout
51. E. H. Taylor
52. F. C. Uffelman
53. F. C. Vonderlage
54. A. M. Weinberg
55. E. P. Wigner
56. C. E. Winters
57. Biology Library
58. Chemistry Library
593-60. Health Physics and Metallurgy

Library

61. Physics Library
62. Training School Library
63-72. ANP Library
73-77. X-10 Central Files
78-83. Reports Office, TID
, ¥Y-12 Area

 

-,

g *
 

&t

 

i Index No. ORNL-1234
Reactors-Research and
Power (Special)

  

External Distributiorng

84-94.
95.
96-~103.
104.
105-109.
110.
111-112.
113,
114,
115-119.
120-139.
140-143.
144.
145-151.
152.
153-156.
157.
158-160.
161.
162.
163-165,
166-168,
169.
170-171.
172-173.
174.
175,
176.
177.
178.
179.
180-181.
182.
183-186,
187-201.
202-216,

 
 

ArgonneWational Laboglftory

Armed FRrces SpecialfNeapons Project {Sandia)
Atomic ERergy CommfEsion, Washington
Battelle morial Ingitute

Brookhavel} National.aboratory

Bureau of SRips
California Rysearclf and Development Company
Chicago Pate{yt Grgup
Chief of NavaRRefearch
duPont Compalg
General Electr §lfCompany (ANPP)
General Electrjil Company, Richland
Hanford Operafids Office

Idaho Operatighs @ifice

Towa State Cgilegd
Knolls Atomjif Powgr Laboratory
Liockland Agea Offie ‘
Los Alamo '.

Massachug®tts Instifite of Technology (Kaufmann)
Massachuffetts Instit§fe of Technology (Benedict)

Mound Lgporatory

Nationalfhdvisory Comittee for Aeronautics, Cleveland
NationajAdvisory Co ittee for Aeronautics, Washington
New Yk Operations (¥fice

North @merican Aviatide, Inc.

Nucleffr Development Afsociates (NDA)

Patej Branch, Washingfn

RangfCorporation

Savianah River Operation} Office (Augusta)

SagBnnah River OperationWOffice (Wilmington)

UA F-Headquarters-Office §f Assistant for Atomic Energy
iversity of California RadWgtion L.aboratory

tro Corporation of Americd

estinghouse Electric Corporgion

right Air Development Cente
Technical Information Service, Oak Ridge

    
  
     
  
   
   
   
 
 
 
 
  

   
 
   
   
  
 

 

 
   
  
 
   
  
    
 

TABLE OF C

FOREWORD . . . .

PART I. CIRCULATING- L AIRCRAFT REACTORS
INTRODUCTION
CHARACTERISTICS OF CIRCULANING-FUEL REACTORS

uel Reactor Systems
eactors

Heat Transfer in Circulating
Physics of Circulating-Fuel
Other Considerations .

DESIGN STUDIES OF AIRCRAFF REACTOR SYSTEM

Turbojet Radiators

Circulation of Fuel Throu .
rated Circulating-Fuel Reactor

Beryllium Oxide-Mo

Engines and Radiato

Shielding .

Accessory Systems '
Circulation of Fuel Through Intermedlate Heat Exchangers

Tandem Heat Exchanger and Shield Design

Annular Heat Exchanger and Shield Design

Spherical Heat Exchanger and Shield Design .

PART II.

Page
No.

 

10

13
15

16
17
19

21

23
25
27
29
31
31
33
36
39

THE CIRCULATING-FUEL AIRCRAFT REACTOR EXPERIMENT

INTRODUCTION
ENGINEERING DESIGN .

General Description

Core Design . .
Primary Coolmg Clrcult
Secondary Cooling Circuit
Structural Assembly :
Core Temperature Distribution

|

 

43

47

47
49
52,
54
54
54

<y
™

 

 

Page

No.

External Fluid Circuit . . . . . . . . . . . . . . . . . 55
Primary Cooling System . . . . . . . . . . . . . . 58
Secondary Cooling System . . . . . . . . . . . . . . 58
Monitoring Circuit . . . . . . . . . . . . . . . .. 58
Pumps . . . C e e e e e e e e e e e e e e e 58
Heat Exchanger C e e e e e e e e e e e e e e 62
Preheating System . . . . . . . . . . . . . oo L 63
Reactor Assembly . . . . . . . . . . . . . . . .. 63
Piping . . C e e e e e e e e e e e e e 63
Heat Exchangers Coe e e e e e e e e e e e e 63
Reactor Controls . . . . . . . . . . . . « . « o . . .. 64
Shim Control . . . . . . . . . . . .« .« .« . . .. 65
Regulating System . . . . . . . . . . . . . . . .. 67
Safety System . . . . . . . . . . . .. o0 .. 68
Electrical Power Circuits . . . . . . . . . . . . . . . . 71
Instrumentation . . . . . . . . . . o o .o .0 L L 73
Flow Measurement . . . . . . . . . . .« . « « « « . 73
Pressure Measurement . . . . . . . . . . . . . . . 77
Temperature Measurement . . . e e e e e e e 77
Liquid Level Indicators and Controls Coe e e e e e 78
REACTOR PHYSICS . . . . . . . « « « « o v v v & « v v . 79
Statics . . . C e e e e e e e e e e e 79
Reactor Volume Fractmns .o C e e e 81
Critical Mass and Total Uranium Investment Coe e 81
Reactivity Coefficients . . . . . . . . . . . . . . . 84
Power Distribution . . . e e e e e e e 84
Neutron Flux and Leakage Spectra e 86
Kinetics . . . . . . . . . . . . . .00 86
Control . . . . e e e e e e e e e e 91
Shim Control Requlrements e e e e e e e e e e 91
Regulating Rod . . . . . . . . . . . . . . . . .. 93
Safety Rods . . . . . . . . .« .« . . . o 0 e e 93
REACTOR MATERIALS . . . . . . . . . « « « v « v v « . 94
ngh Temperature Liquid Fuels . . . . . . . . . . . . . 94
Homogeneous Fuels . . . . ., . . . . . . . . . . . 95
Nonmoderating Fuels . . . . . . . . . . . . . . . . 95
Chemical Properties . . . . . . . . . . . . . . . . 97
Physical Properties E 100
Chemical Prpecessing . . . . . . . . . =0 o o . .. 101

cnmsii—
6 ! .

Structural Metals

Static Corrosion
Dynamic Corrosion .
Control Rod Fabrication
Stress-Corrosion Tests
Radiation Damage

CRITICAL EXPERIMENTS

Experimental Materials
Experimental Assembly
Experimental Program

REACTOR OPERATION .

Pretesting .

Loading and Start- Up ,

Power Operation .

Operation of the Servo- Controls

Shut Down .
ARE BUILDING
Building Schedule

Arrangement
- Utilities .

CONSTRUCTION SCHEDULE

CONCLUSIONS .

BIBLIOGRAPHY

 

PART III.

APPENDIXES

 

Page
No.

 

103
104
106
110
111
112

116
116
116
117
118
119
120
121
122
123
125
125
125
129
131

133

137

il

»
«

10
11
12
13
14
15
16
17
18
19
20

21

 

 

LIST OF FIGURES

Title

 

Uranium Investment as a Function of Various Reactor
Parameters

Mach 1.5 Aircraft with a Circulating-Fuel Reactor
BeO-Moderated Circulating-Fuel Aircraft Reactor
Power-Plant Arrangement

Hs0O-Moderated, Circulating-Fuel Reactor with a Tandem
Heat Exchanger

NaOH-Moderated, Circulating-Fuel Reactor with an
Annular Heat Exchanger

Reactor with a Spherical Heat Exchanger
Schematic Arrangement of the Aircraft Reactor Equipment
Elevation Section of the Experimental Reactor
Plan Section of the Experimental Reactor
Temperature Profile in Fuel Tube and Moderator
Fluid Circuit Equipment Arrangement

Primary Coolant Circuit

Secondary Coolant Circuit

ARE Centrifugal Pump

Fuel Injection System

Reactor Control Elements

Reactor Control Mechanism

Safety Interlock and Scrarfi Circuit

Electrical Power Circuit

Fluid Circuit Instrum%;tation A&

 

 

Page
No.

 

18

24
26

28

35

37
40
48
50
51
56

57
59
60
61
66
69
70
72
74
75
8  aml——

 

 

 

Page

Fig. Title No. *
22 Radial Power Distribution in the Experimental Reactor 85 ¥
23 Axial Power Distribution in the Experimental Reactor 87
24 Leakage Spectrum from the Reflector of the Experimental

Reactor 88
25 Leakage Spectrum from the Open Ends of the Experimental

Reactor 88
26 Neutron Flux Specira at Various Locations in the Experi-

mental Reactor 89
27 Radial Neutron Flux Distribution in the Experimental Reactor 90
28 The System NaF-BeF,-UF, 98
29 The System NaF-KF-ZrF; 99
30 Density of NaF-KF-ZrF,-UF, as a Function of Temperature 102 .

31 Viscosityof NaF-KF-ZrF4-UF4 as a Functionof Temperature 102

~ -

32 Effect of NaK Additative on the Corrosion of Inconel
Thermal Convection Loops by NaF-LiF-KF (11.5-42. 0-

46. 5-mol %) 107
33 Corrosion of Inconel Thermal Convection Loops by

NaF-KF-ZrFs-UF4 (4. 8-50.1-41.3-3.8 mol %) After 500

Hours at 15000F 108
34 Effect of Neutron Irradiation on the Static Corrosion of

Inconel by NaF-KF-UF, (46.5-26.0-27.5 mol %) at 15000F 114
35 '~ The ARE Building Facility {Looking Northwest) 126
36  Plan of the Building Facility 127
37 Elevation of the Building Facility 128

¥

 

ey
u?

Table
No.

 

e

10

11

I2

13

14

15

16

17

LIST OF TABLES

Title

 

Weight Breakdown of the Mach 1.5 Airplane

Design Specifications of the Circulating-Fuel Aircraft
Reactor ) |

Engine Performance at Mach 1.5 and 45,000 Ft

Allowed Contributions to the Total fiadiation Dose in
Crew Compartment

The Principle Shield Dimensions

Shield Weights for Circulating-Fuel Reactors with
Various Intermediate Heat Exchangers

Aircraft Reactor Expériment Dés{gn Data

Summary of Design Data of the E};perimental Reactor
Reactiv.ity Effects Provided for by the Criticall. Mass
Uranium Requirements of the Experimental Reactor
Reactivity Coefficients of the Experimental Reactor
Shim Control Requirements

Melting-Point Data for Potential Fuel Systems

- Static Corrosion of Structural Metals by Potential Fuels

Chemical Analysis of the NaF-KF-ZrF,-UF, Fuel for
Ccrrosion Products

Inconel Stress-Rupture Data

Pile'Irradiation Tests on Fused Fluoride Fuels in Inconel

 

Page
No.

 

23

25

27

30

30

34
45
82
83
83
84
92
96

105

109
111

113
10 & - |

FOREWORD

In addition to fundamental research on problems pertinent to the air-
craft applications of nuclear power, the ultimate objective of the Aircraft
Nuclear Propulsion Project at the Oak Ridge National Laboratory is the de-
sign, construction, and operation of a nuclear power plant capable of super-
sonic aircraft propulsion. Toward this latter end the Laboratory has in-
vestigated all the promising types of reactors proposed in the three years
of the existence of the projectat ORNL. Duringthistime it hasbeen guided
by the results of the Technical Advisory Board!l (a joint NEPA-ORNL ad-
visory committee), the work of the NEPA project, ¢ and the study by the
Lexington Project.3 These earlier surveys suggested the feasibility of
aircraft nuclear propulsion by any one of several systems, but it was left
for subsequent investigations to extend the necessary research and ana-
lytical studies to permit a detailed evaluation of the various systems. As
a result of its studies, the Laboratory has selected the circulating-fuel
reactor, employing a molten mixture of metal fluorides, as the type of
reactor with which the objective of supersonic propulsion will be achieved
most readily.

Since for any type of aircraft reactor the extrapolation of present knowl-
edge is too uncertain to permit the immediate construction of that reactor,
a low-power experimental reactor was recognized as a necessary step
toward the ultimate goal. This report outlines the practicality and poten-
tialities of the circulating-fuel aircraft reactor {(Part I) and then describes
in detail the ARE —Aircraft Reactor Experiment (Part II) —with which per-
tinent and reliable data can most economically be obtained to permit the
design. development, and construction of the actual aircraft reactor.

 

1. "Report of the Technical Advisory Board to the Technical Committee

of the Aircraft Nuclear Propulsion Program, " ANP-52, Aug. 4, 1950.

"Final Status Report of the Fairchild NEPA Project, " NEPA-1830, 1951.

3. "Nuclear Powered Flight, A Report to the Atomic Energy Commission
by the Lexington Project," Lex P-1, Sept. 30, 1948.

4N

 

 

Ny
ot

 

PART 1

CIRCULATING-FUEL AIRCRAFT REACTORS

 
 

 

 
K ¥

i

GRS | 13

 

L

INTRODUCTION

The development of a supersonic aircraft is so extremely difficult that
to date few have been built and these have achieved supersonic speeds for
only a few minutes. The concurrent development of a light, compact nu-
clear power plant to propel a supersonic aircraft would appear to be vastly
more difficult, and yet the resulting weapon would be so potent as to war-
rant extensive effort. One objective of the Oak Ridge National Laboratory
in the National Aircraft Nuclear Propulsion Program is the development of
a nuclear power reactor with which supersonic aircraft propulsion can be
obtained. |

The ultimate reactor must attain the highest possible output per pound
of machinery and the greatest possible thrust per square foot of frontal
area. In fact, unless very good values for these two parameters areat-
tainable, an operational supersonic airplane is simply not feasible. Ob-
viously, the better the performance of the power plant over and above the
acceptable minimum, the easier and more reasonable the airplane design
and development problems become. Practicallyall the weight and harmful
frontal area of a complete aircraft with a nuclear power plant are in the
engines and in the reactor-shield assembly. (The part of the frontal area
that consists of air duct inlets need not cause much drag.) The engine
thrust per unit of both weight and frontal area increases with turbine air
inlet temperature irrespective of whether a turbojet, turboprop, or a
ducted-fan type of engine is used. Since the product of the thermodynamic
and propulsive efficiencies alsoincreases though less rapidly than engine
thrust, an increase in operatingtemperature also means a reduction in the
reactor power required and, hence a smaller and lighter reactor-shield
assemblyfor a given total thrust. Possibly an even more important factor
is that the frontal area and the weight of the reactor-shield assembly in-
crease rapidly with reactor diameter for a given power output. The per-
formance of the airplane will therefore improve with increases in either

the t¥mperature and/or the power density obtainable in the reactor pro-

vided, of course, that the temperature increase can be accomplished with-
out impairing the reliability of the power plant.

In surveying the reactor systems that have been proposed for super-
sonic aircraft propulsion, it has seemed to the ANP group at ORNL that
the circulating-fuel type promisesthe highest temperature and power den-
sity=—and hence the highest over -all performance—of any power plant whose

n—

 
14

construction curretftly appears feasible. Inthis connection itis to be noted
that the Technical Advisory Board" concluded that a homogeneous reactor
has outstanding advantages for nuclear aircraft propulsion. (although con-
tingent upon the containment of hydroxides at high—around  15000F —tem-
peratures). Furthermore, the circulating-fuel reactor, as presently con-
ceived, has incorporated essentially all of the advantages attributed to the
homogeneous type of reactor (without having to resort to the use of hydrox-
ides). The mutual advantages of these two reactor systems are (1) sepa-
ration of the heat exchange function from the reactor core, (2) a simple,
rugged core structure, (3) self-stabilization associated with a liquid fuel,
(4) simplified fuel removal and reprocessing, and (5) an arrangement
whereby the highest system temperature occurs in the working fluid.

 

1. "Report of the Technical Advisory Board to the Technical Committee of
the Aircraft Nuclear Propulsion Program,' ANP-52, Aug. 4, 1950,

%

Xt
wF

15

CHARACTERISTICS OF CIRCULATING-FUEL REACTORS

The established requirementslsZ2 of a high-performance reactor for

supersonic aircraft propulsion are high operating temperature----as re-
quired by the thermodynamics of aircraft engines, and high power den-
sity----as required to minimize the size of the shielded volume. These

requirements imply the use of liquid cooling systems that not only permit
small reactor cores but whichalso have good heat transfer characteristics.
For some time the specific objective of the ORNL-ANP project has been
the development of a liquid-cooled reactor employing a nonoxidative high-
temperature liquid.

However, in addition to the above criteria of high temperature and
high power density, two fundamental concepts constitute the real justifica-
tion of the circulating-fuel reactor now being developed by the Laboratory.
These are (1) inherent self-stabilization of the reactor against random
transients and outright power change, and (2) elimination of the heat ex-
changer from the active lattice of the reactor. The first conceptagain im-
plies a liquid fuel; the second, a system in which the fuel itself is the me-
dium that carries the heat out of the core.

In their scope these concepts include both the circulating-fuel and ho-
mogeneous types of reactors. However, at the present stage of materials
development satisfactoryhigh-temperature homogeneous fuels donot exist.
(A solution of UQO;5 in a hydroxide now appears to be the most promising
homogeneous fuel, but the hydroxides cannot be contained at the required
temperatures without excessive corrosion. 3) On the other hand, molten
fluoride salt mixtures containing the uranium salt have not only been con-
tained but also circulated, pumped, and otherwise manipulated for long pe-
riods of time at temperatures of 15009F. They do not, however, possess
a substantial capacity for moderation. Consequently most of the design
work to date onthe molten-fluoride circulating-fuel reactor, has employed
beryllium oxide as the moderator. 4

 

2. C. B. Ellis, 'The Technical Problems of Aircraft Reactors,’ ANP-64,
Parts 1 and II, June 1, 1951 and Jan. 1, 1952.

3. See sections on ''Corrosion Research' in the last several Quarterly
Progress Reports of the Aircraft Nuclear Propulsion Project at the Oak
Ridge National Laboratory, ORNL-1154, -1170 and -1227.

4. A similar reactor using the molten fluorides as a homogeneous fuel,
i. e., no moderator employed in the reactor, is feasible and constitutes
a more advanced reactor design upon which the second ARE may be
based.
16

The engineering separation of the reactor core and the heat exchanger
greatly simplifies the design and fabrication and improves the efficiency
of each. However, this separation is achieved at the expense of a larger
shielded volume since the intensely radioactive fuel solution must be car-
ried outside the reactor for heat removal. The radiator, or intermediate
heat exchanger, is then so radioactive that it requires roughly two-thirds
as much shielding as the reactor core. Since the shielded volume and the
shield weight thus increase beyond that required for the reactor alone, it
is not obvious that the circulating-fuel system is advantageous compared
to the several fixed-fuel proposals. This point has been examined with
great care and it was only after it became clear that the shield weights
were reasonable that the present path was chosen.

The circulating-fuel reactors described here embody all the concepts
discussed above and lead to the design of rugged, high-flux power sources.
The reactors are designed with a fuel exittemperature of 1500°F and should
permit very high power densities. Furthermore, the only heat transfer
within the core is a relatively small amount from the moderator, because
the circulating-fuel carries the heat out of the core to external heat ex-
changers. The reactors should be stable because the rise in fueltempera-
ture accompanying any increase in power causes some of the liquid fuel to
be expelled from the core as aconsequence of simple thermal expansion of
the fuel.

Heat Transfer in Circulating-Fuel Reactor Systems

The advantage of separating the heat exchanger and core is readily ap-
parent from a consideration of heat transfer in fixed-fuel reactors. Since
the heat flux to achieve high power density in any static fuelreactor is very
large, it is necessary tohave alarge heat transfer surface to keep the heat
flux to an attainable magnitude. However, because of the nuclear poison
introduced by the available metalheat transfer surfaces when these are in-
corporated in the active lattice of the reactor, there is a limitto the amount
of metal surface that can be tolerated within the reactor. Even if the poi-
son were of no concern and unlimited fuel material were available, there
is still quite a finite limit to the surface-to-volume ratio. This imposes
an upper limit—that is not high—to the power that can be extracted from
a given volume of a fixed-fuel reactor. This constraint is avoided in the
circulating-fuelreactor by removing the heat from the reactor by circulat-
ing the fuel through heat exchangers external to the core where nuclear
poison considerations no longer pertain.
17

The heat thus developed in the circulating-fuel reactor can be trans-
ferred to the engine air stream either directly (by circulating the fuel
through the air radiators) or through the use of an intermediate heat ex-
changer and fluid. In either case, this reactor design has removed the
heat transfer problem from the reactor core so that extremely high power
densities may be achieved therein. At the same time, it becomes possible
to make the fuel temperature atthe reactor outlet approach that of the tem-
perature of any structure in the system, and provision can even be made
to cool the structure through the use of a circulating liquid moderator.
Furthermore, it is possible to reduce the temperature differential between
the peak metal temperature in the reactor and the air temperature in the »,
engine turbine to a value even less than that obtainable in the 'directair cycle, -
In addition, the upper temperature limit of the system is imposed by the
strength and corrosion characteristics of the structural metal, which are
better in a molten salt environment than in air. Therefore, since turbojet
performance improves rapidly with an increase in turbine air inlet tem-
perature, the circulating-fuel reactor promises to give the highest per-
formance—in this respect—of any nuclear power plant proposed to date.

It may be emphasized that the problems avoided here—heat transfer
and hot spots within the reactor core—are critical problems for both the
air and supercritical-water cycles. The circulating-fuel reactor is tech-
nologically distinct. Similar difficulties will not beset it.

Physics of Circulating-Fuel Reactors

 

Surveys of the circulating-fuel reactor system have been conducted
for several parametric variations. For a given moderator these param-
eters were core diameter and the ratio of the volume of circulating-fuel
to that of the moderator. A simplified calculation method was used, and
the assumed fuel probably differed somewhat in compositionfrom that which
will ultimately be used. Thus, the absolute values of the results shown in
Figure 1 are not exact, although the important parametric relationships
are applicable to any circulating-fuel system. If the core diameter is
fixed, the critical mass increases as the fuel-coolant volume fraction in-
creases, but the uranium density in the fuel-coolant at firstdecreases and
then increases. Since from 60 to 80% of the liquid fuel is in the heat ex-
changer, pumps, or piping, the minimum uranium investment inthe overall
system is obtained with a more dilute fuel solution than that for minimum
critical mass. This minimum total uranium investment usually results in
a reactor with about half thermal fissions, i.e., a reactor on the border-
line between the epithermal and the intermediate. Although the data for
ANP-PHY-223

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

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o el ke
20 |+~ ~7 3-ft DIA CORE
——
0 -
0 0.40 0.20 0.30 0.40 0.50 0.60 0.70

- VOLUME FRACTION FUEL-COOLANT IN CORE

FIGURE I. URANIUM INVESTMENT AS A FUNCTION OF VARIOUS
REACTOR PARAMETERS

8l
 

19

Figure 1 were prepared for a beryllium oxide-moderated reactor, other
calculations indicate that if water were substituted for the beryllium oxide
the curves would be shifted less than 10%, and if sodium hydroxide were
substituted for the beryllium oxide the curves would be shifted upward a
little and considerably to the left.

The enormous power density required in an aircraft reactor coupled
with the demand for accurate and dependable control essential for aircraft
use make it imperative that the reactor possess an inherent stability a-
gainst anomalous transients. This requirement is most easily met if the
fuel consists of a liquid with a relatively large volume increase with in-
creasing temperature. Sincepower is producedin the fuel, this volumetric
change should lag the power production very little and the desired self-
stabilization should result. The use of circulating liquid fuel (as well as
static liquid fuels) results in reactors with self-stabilizing characteristics.
On the other hand the loss of delayed neutrons from the active lattice,
caused by the circulation of the fuel, had been viewed with concernbecause
these neutrons have a damping effect on oscillations in reactor power. It
now appears that the loss of these delayed neutrons may be compensated
by a damping mechanism associated with the circulation of the fuel.

Other Considerations

 

Although the mostvital considerations in the selectionof a nuclear air-
craft power plant are weight and frontal area per pound of thrust—whichare
both a functionof turbine inlet temperature—other factors are also impor-
tant. The amount of fissionable material required per airplane; the time,
cost, and inventory of fissionable material involved in reprocessing the
fuel; the composition of the fuel; and the ease of maintenance of the com-
plete aircraft power plant are also pertinent factors. Fortunately, the
proposed circulating fluoride fuel should be relatively easy to reprocess,
and the fissionable material installed in each airplane should run less than
100 pounds. Both of these factors combine to give what appear to be the
lowest total fissionable material investment per operational airplane {in-
cluding material in reprocessing, inventory, etc.) for any aircraft reactor
yet proposed.

The possibility of draining the liquid fuel from the reactor after a
flight can greatly simplify the airplane maintenance problem. The refuel-
ing problem is obviously vastly easier than changing solid fuel elements.
In addition, design studies have shown that it is possible to design the re-
actor so that specially shielded or remotely controlled ground handling
equipment need be provided only for the fuel filling and draining operations
20

and yet keep the weight of a 400, 000-kw reactor, intermediate heat ex-
changer, and shield combination under 120, 000 pounds. Even if a lighter,
divided shield were used, once the fuel was drained radiation should im-
pose practicallyno limitations on maintenance work. The combined bene-
fits of these factors should decrease ground time, and very possibly make
the in-service time of an airplane using a circulating-fuel reactor a high
multiple of that of one with a solid fuel. |

ko
21

DESIGN STUDIES OF AIRCRAFT REACTOR SYSTEMS

The circulation of fissionable material external to the core of the re-
actor directs attention to design studies of components of the entire system,

‘such as the reactor, radiator, and engine. Although the circulating-fuel

system has not been studied sufficiently to define the performance of the
resulting aircraft, enough work has been done to appreciate the more out-
standing problems. The preliminary design studies—both with and without
an intermediate heat exchanger—of a supersonic airplane application of a
circulating-fuel reactor are encouraging. These studies were not intended
to establish the finaldesign for such an aircraft but to resolve some gen-
eral questions regarding the feasibility of supersonic aircraft with circulat-
ing-fuel reactor power plants.

The decision to employ a circulating-fuel type of reactor did not nec-
essarily resolve the alternative choices that exist in the several other
features of the reactor system, namely, the type of moderator, the type
of heat exchanger and shield system, or even the nature of the fuel. At
this time, however, it.is considered that the fuel will be a mixture of metal
fluorides, including uranium fluoride. Certain of these salt mixtures have
particularly advantageous combinations of desirable physical properties,
including high heat capacity, good thermal conductivity, adequate thermal
expansion, satisfactory viscosity, sufficiently low melting point, low nu-
clear cross sections, excellent corrosioncharacteristics, and satisfactory
radiation stability. It has not been possible, however, to make an irrevo-
cable decision at the present time regarding either the type of moderator
or the type of heat exchanger and shield system that will ultimately prove
most desirable for supersonic aircraft application.

Several broad classes of moderator were reviewed for applicability
to the circulating-fuel concept, including:

1. low-temperature hydrogenous liquids (such as water),

2. intermediate-temperature hydrogenous liquids (such as hy-
droxides),

3. high-temperature solid moderators (such as beryllium oxide).

Each of these possible moderator arrangements appears to offer some
unique advantages and some disadvantages. Consequently, each of the
reactor-shield combinations presented below employs a different type of
moderator. No attempt is made in this report to evaluate the relative
merits of the various moderator arrangements, nor is it clear that any
one is superior to the others.
22

There exists a fundamental choice regarding the heat transfer system
in the airplane, i. e., whether to circulate the fuel directlythrough the en-
gine radiator or toemploy an intermediate heat exchanger and asecondary
heat transfer fluid. The former technique would appear ito require the
simplest mechanical system, but it is resiricted to the use of a divided
shield and only certain special types of airplane configuration because of
the intense source of radiation external to the core; the latter technique
would be readily adaptable to a wide range of airplane configurations and
would permit the use of near-unit shields, which would greatly facilitate
servicing and maintenance of the aircraft. At this stage in the design of a
supersonic nuclear aircraft the system that willultimately prove superior
is largely a matter of opinion. A fairly complete outline of a design of a
supersonic airplane in which the fuel is circulated to the turbojet radiator -
is presented together with severalreactor and intermediate heat exchanger
arrangements employing various approximations to a unit shield to give
some notion of the possibilities of liquid-fuel aircraft reactors.

In the design of an aircraft reactor it is most important to know the
effects of various parameters on the weight of the aircraft reactor and
shield, but it is very difficult to determine these effects quantitatively be-
cause a shield having anything like the minimum weight is a very closely
integrated, complex unit. The only course seems to be to make quite a
number of different detail designs on essentially the same bases and then
compare them. Unfortunately, the bases cannot be identical because any
engineered shield involves many compromises and, at present, many es-
timates and assumptions. As a result, even for comparative purposes
weight data on engineered shields probably cannot be estimated closer than
+ 5% if a wide variety of shield types is considered. This follows from the
fact that there is of necessity a considerable variation in conditions, e. g.,
activation of sodium in a secondary circuitis a muchmore serious problem
with a circulating fuelthan with sodium as a primary coolant. These con-
ditions should be kept in mind in comparing the shield weights presented
in this report. In all cases the shielding was designed to give 1 r/hr in
the crew compartment at full power output. ’£hese studies have been re-
ported in greater detail in separate reports. 3 One of these reports con-
cerns the design in which the fuel is circulated through the engine radiator;
the other, the reactor and shield design studies in which the fuel is circu-
lated through an intermediate heat exchanger.

 

5. R. W. Schroeder, gg_eil:,, . "_Preliminary Investigation of a Circulating-
Fuel Reactor System for Supersonic Aircraft Propulsion, " ORNL-1287
(to be issued).

6. A. P. Fraas and G, F. Wislicenus, "Several Circulating-Fuel Re-
actor —Heat Exchanger—Shield Designs for Supersonic Aircraft," Y-
F15-10 (to be issued).
Circulation of Fuel Through Turbojet Radiators

The circulation of fuel through the turbojet radiator permits the direct
transfer of nuclear heat to the air stream. Thus, as in the air-cycle re-
actor, there is no temperature loss associated with an intermediate heat
exchanger. The circulating-fuel reactor, however, can have a less com-
plex design and more rugged structure than an air-cycle reactor because
it does not also function as a heat exchanger. The shielding problem
changes from one of shielding a large reactor with air ducts to one of shield-
ing the crew from the radioactive fuel in the radiator and other plumbing
external to the core. This, of course, necessitates the use of a divided
shield, with the major portion of the shield around the crew because the
size and number of the air radiators makes it quite impractical to shield
them.

An airplane withadesignpoint of Mach 1. 5 at 45, 000 ft, incorporating
a circulating-fuel reactor in which the fuel is circulated through the engine
radiators has been outlined. This aircraft, Figure 2, has a gross weight
of approximately 350, 000 1b, a L./D ratio of approximately 6. 5, and a wing
loading of approximately 70 1b/ft2. A divided shield is employed with
hydrogenous shielding about the reactor and with hydrogenous and lead
shielding about the crew compartment. By confining the six turbojet en-
gines to the space in the fuselage behind the reactor, by using a 120-ft
reactor to crew separation distance, and by employing a small crew com-
partment (4.5 by 5.5 by 12 ft) it appears possible to keep the weight of the
reactor-shield-crew combination down to around 105, 000 pounds. The
weights of various components of the aircraft are given in Table 1.

TABLE 1

WEIGHT BREAKDOWN OF THE MACH 1.5 AIRPLANE

Component Weight (1b)
Airframe 105,700
Engines 59,900
Radiators 23,800
Auxiliary equipment (pumps, ducting,

etc.) 15,300
Reactor and reactor shield 48,400
Crew shield 56,600
Payload 20,000
Contingencies 20,300

Total 350,000
 

FIGURE 2.

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

MAGH

1.5

 

AIRCRAFT WITH A CIRCULATING-FUEL REACTOR

pbll=Ld4-A il

vt
Tz

25

The reactor and fuel system has sufficient overload capacity to pro-
vide approximately 155,000 pounds thrust for take-off and landing, with
approximately double design point reactor power. The radiation dose re-
ceived by the crew could be permitted to double {from 1 r/hr to 2 r/hr) for
the short periods involved in these operations as well as during the initial
climb. As will all divided shield reactor arrangements, special provision
would have to be made for ground personnel.

Beryllium Oxide-Moderated Circulating-Fuel Reactor

The reactor designed for use withthe airplane outlined above employed
beryllium oxide for both moderator and reflector. The core consisis of a
number of parallel tubes, arranged in two-series passes, to convey circu-
lating fuel through the beryllium oxide lattice. The beryllium oxide reflector
is provided adjacent to all core surfaces except the fluid inlet and outletend.
Moderator heat is removed by the circulating fuel and is employed usefully
in the propulsion cycle. By proper allocation of coolant flows, coolant-
tube wall temperatures should run only slightly inexcess of the circulating-
fuel maximum temperature. To permit removing reflector heat with a high
radiator temperature difference, the design under study postulates cooling
the reflector with a non-fuel-bearing fluoride mixture. Thisreactor, Figure
3, conforms to the specifications given in Table 2.

TABLE 2

DESIGN SPECIFICATIONS OF THE CIRCULATING-FUEL AIRCRAFT REACTOR

Size 3.5-ft sphere
Uranium Investment (Total System ) 120 1b
Core Constituents
Moderator BeO
Fuel Fluoride Fuel Mixture
Structure Inconel
Core Volume Fractions
Moderator ~) .65
Fuel ~0.33
Structure ~0.02
Design Point Characteristics
Power ~320,000 Btu/sec
Fuel inlet temperature ~1000°F

Fuel outlet temperature ~1500°F
26

   

 
       

FUEL OQUT ===t
FUEL OUT

SALT ouUT
=~S5ALT. IN -G PLACES

*

 

5

FIGURE 3. Be O-MODERATED CIRCULATING-FUEL AIRCRAFT REACTOR i’
27

Engines and Radiators

A circuMating-fuel type of reactor should be inherently capable of sup-
plying propulsive power by means of a turbojet cycle or any of several
vapor cycles. The relative performance of these cycles has not yet been
clearly established, but, pending comparisons, the use of the turbojet cycle
has been presumed. Preliminary optimization studies have indicated that
a compressionratio in the regionof 6:1 with aturbine air inlet'temperature
of approximately 1250°F is desirable. Further increases inthe compression
ratio would diminish the radiator weight at the expense of engine weight,
and further increases in turbine inlet temperature (with fixed reactor con-
ditions) would favor turbojet weight to the detriment of the radiator log
mean temperature difference and radiator weight.

Having established the total air flow requirements, the compressor
inlet area needed was established on the basis of NACA developmental ex-
perience. The number of engines necessaryto accomodate the total air flow
(or to provide the total inlet area needed) will depend on the size engine
that can be made available when an airplane of the type described will re-
quire engines. Consequently the specification of the actual number of en-
gines at this time would be very arbritary. The use of any number of en-
gines, within reason, would have only secondary effects on the overall
airplane weight. However, since the use of six engines appears to be con-
venient from an installation standpoint, the use of six engines, Figure 4,
conforming to the general specifications listed in Table 3, has been assumed.

TABLE 3

ENGINE PERFORMANCE AT MACH 1.5 AND 45,000 FT

Thrust per engine 8100 1b
Compression ratio 6:1
Turbine inlet temperature 1250°F
Equivalent sea-level airflow 665 1b/sec

The total thrust provided by the six engines plus that from the auxil-
iary systems was found to be approximately 54, 000 lbs. Calculations have
indicated a specific impulse of approximately 28 and an over-all thermal
efficiency of 29%. This latter is defined as the ratio of net thrust horse-
power to reactor thermal horsepower.

The radiator design involves a radial grouping of rectangular-shaped
banks about the engine centerline between the compressor and turbine.
Liquid-conveying tubes are passed through, and normal to, closely spaced
 

 

 

 

 

 

 

 

 

§—4

 

 

 

 

 

 

 

 

 

 

 

REACTOR SHIELD ASSEMBLY
MAIN TURBOJET ENGINE

MAIN ENGINE COMPRESSOR
FUEL-TO—-AIR RADIATORS

MAIN ENGINE TURBINE

SHIELD WATER RADIATOR
REFLEGCTOR GCOLANT RADIATOR

REFLECTOR COOLANT PUMP
WATER PUMP AIR TURBINE

ELECTRIGC GENERATOR
ELECTRIC GENERATOR AIR TURBINE

L M
Qi
A
B.
G
D.
E.
F.
G.
H. FUEL PUMP
\ll WATER PUMP
K.
L.
:,- REACTOR
FIGURE 4 POWER PLANT

ARRANGEMENT

vgsl-~-24A (ENEUR

8¢
29

sheet fins. The flow pattern contemplated is countercurrent, as dictated
by the temperatures of the two fluids. The arrangement entails the use of
dividing baffles between adjacent banks, and the installation of by-pass
valves in these baffles will permit controlling turbine inlet temperature
while substantially constant liquid mean temperature is maintained.

Shielding

The circulating fuel introduces new problems in shielding, due to the
delayed neutrons and fission product gamma rays which must be shiélded
out at the crew position. These radiations specify the crew shield thick-
nesses within rather narrow limits, thus removing much of the freedom
usually enjoyed in the design of a divided shield. Furthermore, this shield
differs from other divided shields in that the reactor shield is of uniform
thickness throughout, and all the lead shielding is located at the crew com-
partment inside the plastic layer.

The method of design is much like that of ANP-53, and is described
in another report.” However, considerable uncertainty exists in the strength
of the fission product gamma source, which is the dominant component. As
a consequence, it was necessary to incorporate considerable conservatism
in the design. The total tolerance dose was taken to be 1 r/hr, divided as
shown in Table 4. The resulting principle shield dimensions are shown in
Table 5.

 

7.R. W. Schroeder, Preliminary Investigation of A Circulating Fuel Re-
actor System for Supersonic Aircraft Propulsion,’ ORNL-1287, to be
issued.
- 30

TABLE 4

ALLOWED CONTRIBUTIONS TO THE TOTAL RADIA'I"ION, DOSE
IN CREW COMPARTMENT

 

Component

A. Neutrons

O A0 O

. Radiators to rear
. Radiators to sides
. Radiators to front
. Reactor to rear

. Reactor to sides

Reactor to front

Total

B.. Gamma Rays

-

m 0 a0 T

Radiators to rear
Radiators to sides
Radiators to front
Reactor to rear

- Reactor to sides

Reactor to front

Total

Radiation Dose
rem/hr rep/hr

Flux

 

0,002 29.4/cm? sec _
0.002 - 1/4(29.4/cm?2 sec) per side

 

0.250 5 x 104 hard gamma/cm? sec

0.025 1.38 x 10% mev/cm? sec

TABLE 5

0.02
0.02
0.005 0.0005 7.25/cm? sec
0.10  0.010
0.10 0,010
0.005  0.005
0.25 0,025
0.250
0.300 0.300 See ANP-53
0.025
0.100 0,100
0,06  0.05
0.025 0.025
0.75  0.75

THE PRINCIPLE SHIELD DIMENSIONS

 

Shield

Reactor (all sides)

Crew Shield
Rear
"~ Sides

‘Vertical Front
Slanting Front

Material
Water

Lead and Plastic
Lead and Plastic
Lead and Plastic
Lead and Plastic

Thickness (cm)
78

22.4 and 33.4
2.75 and 39.0
0.62 and 37.3
1,50 and 38.0
31

Accessory Systems

 

Accessory circuits would be required to permit cooling the reflector
and shield and to provide power for pumps and general aircraft accessory
demands. Since it appears that many of the accessories are favored by the
use of variable-speed drives, the over-all accessory system comtemplated
involves the use of individual pneumatic turbines for power supply. The
energy is supplied, and reflector cooling is achieved, by passing com-
pressor bleed-off air through the reflector radiators and then to the ac-
cessory turbines and to parallel propulsive nozzles. The accessory tur-
bines are controlled by means of variable-area discharge nozzles, which
provide some net thrust after imparting energy to the accessory turbines.
Preliminary analysis of this system indicates that it not only provides re-
flector cooling and accessorypower adequately, but that the reflector cool-
ing system, when considered as an cpen, Brayton-cycle power plant, has a
favorable cycle efficiency and power-to-weight ratio. The use of individual
accessory-drive turbines permits supplying the accessories with power and
speed control without special mechanical, hydraulic, or electrical trans-
mission systems.

The radiator for the shield water is cooled by low-pressure compressor
bleed-off air, and the air is then discharged through a propulsive nozzle.

Circulation of Fuel Through Intermediate Heat Exchangers

The limitations imposed by partially shielded nuclear power plants on
ground handling, servicing, and maintenance of nuclear aircraft are quite
serious. The circulation of the reactor fuel to the turbojet radiators im-
poses a further constraint on possible aircraft serviceability, In addition
to serviceability, however, significant limitations are also imposed on
crew compartment size, airplane configuration, and possibly on aerody-
ramic stability and control in nuclear aircraft that employ divided shields.
It is therefore apparent that strong incentives exist for developing reactor
arrangements which are amenable tc a i iype of shield.

- It is notable that for the range of sizes of reactor and crew compart-
ments under consideration, and with the separation distances involved, the
divided shield results in the lowest shield weight; unit shields are appreci-
ably heavier.® Obviously, a type of shield can exist that is intermediate
in weight and concept, between these two extremes which would permit a

 

8 'Report of the Shielding Board for the Aircraft Nuclear Propulsion
Program,' ANP-53, October 16, 1950,
32

mechanic without special shielding to service the engine and reactor—al-
beit for short periods of time. An upper limit on the radiation does from
such an intermediate shield was established as 1 r/hr at 50 ft from the
reactor operating at 5% full power.

A circulating-fuel reactor system in which the ground handling and
servicing are greatly facilitated may be effected by incorporating a com-
pact reactor design that permits easy and complete fuel drainage with a
quasi-unit shield such as that defined above. The reactor heat would then
have to be transferred from the circulating fuel to a secondary fluid in an
intermediate heat exchanger located within the reactor shield.

Since the use of intermediate heat exchangers enclosed within the re-
actor shield substantially increases the size of the shielded volume, con-
siderable effort has beendevoted todetermining the optimum configuration
and location of the heat exchangers. Three heat exchanger and reactor
arrangements have been investigated; each was somewhat better than its
predecessor. Thethree arrangements, in chronological order of analysis,
are: tandem, annular, and spherical. -  Each arrangement is distinguished
by the location of the heat exchanger withrespect to the reactor core. Iden-
tical assumptions were made in the design of the shield for each arrange-
ment. In the arrangements discussed, the use of both low-temperature
(water) and high-temperature (sodium hydroxide) moderators inthe reactor
core are described. No attempt has been made at this time to select the
optimum moderator from among these and the previously discussed beryl-
lium oxide moderator.

In all cases the heat exchanger was separated from the reactor only
by enough material to reducethe reactor neutron-leakage flux entering the
heat exchanger to the level of the delayed-neutron flux. This required the
equivalent of about 10 in. of water between the active lattice and the heat
exchanger matrix. It was found that by lacing the intermediate heat ex-
changer with 8.5 vol % boron carbide it should be possible to utilize prac-
tically any material in the secondary circuit without encountering an intol-
erable amount of activation, because most of the neutrons absorbed in the
heat exchanger would be absorbed by the boron. Probably no coolant likely
to be used would be inclined to become much more radioactive than NakK,
yet evenwith NaK thisheat exchanger arrangement should make it possible
to keep the maximum radiation dose down to less than about 20 r/hr at 5
ft from the engine radiators immediately after a long flight. With fresh
NaK (or NaK that had been allowed to decay for 5 days) the corresponding
dose after 1.5 hr of warm-up at 5% power wouldbe only .1 r/hr. Estimates
of the activation of other coolants such as natural lithium, water, a eutectic
melt of LiCi-MgCl,-KCl, and a noninflammable sodium-lead alloy were
found to give activities from 1 to 50% of that of NaK.
33

It is gratifying to find that it is possible to design something like a unit
shield of about 120, 000 1b that makes possible servicing and tune-up work
with circulating-fuel reactors even with the reactor running and permits
ground personnel to approach a nuclear airplane for short periods without
special shielding in the event of a minor landing accident or the like. A
summary of the various arrangement and design conditions together with
the resulting shield weights are given in Table 6,

Tandem Heat Exchanger and Shield Design

The first arrangement considered was one in which the heat exchangers
were grouped behind the reactor core (Figure 5). This reactor design was
‘based on the use of water as a moderator although it is readily adaptable
to sodium hydroxide or some other liquid moderator. The reactor core is
composed of tubes through which the liquid fluoride fuel flows at about 10
ft/sec for operation at 400, 000 kw with a 400°F temperature rise through
the reactor. The fuel enters at the top aft end of the reactor, makes a
complete loop through the fuel tubes in the core, and discharges to the heat
exchanger. A divided shield and two quasi-unit shields, as summarized in
Table 6, were designed for this heat exchanger arrangement. The shield
design features and specifications are discussed for each case.

The first shield designed for the tandem heat exchanger arrangement
was a divided shield somewhat similar to the NEPA divided shield.}
The reactor shield was designedto provide no heavy material except in its
structure. For simplicity, the shielding material was considered to be
principally borated water containing two atoms of boron for every 100 atoms
of hydrogen. The thickness of the reactor shield was taken as being such
as to give approximately the same dose at the rear of the crew shield as in
the NEPA shield, except that the lead shadow-shield disks at the reactor
in the NEPA shield were moved to the rear of the crew compartment. The
thick layer of steel (3. 75 in) in the form of the core pressure shell and
shield inner shells coupled with a layer of lead shielding around the heat
exchanger made it possible to remove the lead from the sides of the NEPA
crew compartment. The total weight of reactor, intermediate heat ex-
changer, reactor shield, and crew shield was found to be 111, 000 pounds.

A second shield for the same reactor was designed to provide some-
thing as nearly like a unit shield as possible. In this design enough shield-
ing was provided at the reactor to bring the radiation leakage from the
reactor shield down to the same level as that from the sodium in the sec-
ondary circuit. This was done by introducing a layer of lead, 5.4 in. in
thickness, around the reactor and a layer, 10.4 in. in thickness around
the heat exchanger. The thicknesses of the water layers around both the
TABLE 6

PARTIALLY DIVIDED SHIELDS FOR CIRCULATING-FUEL REACTORS WITH VARIOUS INTERMEDIATE HEAT EXCHANGERS

 

 

Tandem Heat Exchanger

Annular Heat Exchanger

Spherical Heat Exchanger

 

 

 

1 2 3 4 1 2 3 1 2 3

Reactor shield diametar {in.} 150 150 121 121 148 118 118 148 118 118
Crew shield weight (1b} 5,000 11,000 36,000 14, 000 5,000 36,000 14,000 5,000 36,000 14,000
Weight of reactor, intermedi-

ate heat exchanger, and re-

actor shield {lb) 151, 000 130, 000 75,000 75,000 123,000 62,000 62,000 115,000 54,000 54,000
Total weight of reactor, in-

termediate heat exchanger,

and all shielding (includ-

ing crew shield) (1b) 156, 000 141, 000 111,000 89, 00C 128,000 98,000 76,000 120,000 90,000 68,000
Reactor power (kw) 4000, 000 400, 000 400, 000 400, 000 400, 000 400, 000 400,000 400, 000 400, 000 400, 600
Diameter of reactor core (in.) 32 32 32 32 32 32 32 32 32 3z
Liquids in primary and sec-

ondary circuits Fluoride-NaK Fluoride-NaK Fluoride-NaK Fluoride-NaK Fluoride-NaK Fluoride-NaK Fluoride-NaK Fluoride-NaK Fluoride-NaK Fluoride-NaK
Temperature loss in inter-

mediate heat exchanger(0F) 100 100 100 100 100 100 100 100 100 100
Pressure loss in intermedi-

ate heat exchanger (psi) 160 100 100 100 100 100 100 100 100 100

i i 1 1 1 i 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

Crew shield size (ft) 6z x7 xlz2 6-2'):7235122 6 x?lez 5x5x122 62x72x122 6Zx'llez2 5x5x122 62:':7'2);122 6Zx72x122 5x5xl123
Reactor-crew separation

distance (ft) 50 50 50 120 50 50 120 50 50 50
Radiation inside crew com-

partment (r/hr) 1 1 1 1 1 1 1 1 1 I
Radiation 5 ft from center

of reactor (r/hr) 300 2,400 380,000 380,000 300 380,000 330,000 300 380,000 380, 000
Radiation 50 ft from center

of reactor (r/hr) 7 36 5,600 5,600 7 5,600 5,600 7 5,500 5,600
Radiation 300 ft from center

of reactor (r/hr) 1 6 156 156 1 156 156 1 156 156

 

 

 

 

 

 

 

 

 

 

 

 

Ve
 
  

U
DWG. E-Y-F 7-474R{

APPROX. 12 in. H,0 AROUND HEAT
EXCHANGER AND PUMP HOUSINGS

 

   
 

H,0 MODERATOR
INLET PIPE

 

Y, -in. STEEL PLATE FOR REACTOR AND

SHIELD SUPPORT, 2 REQUIRED
HEAT EXCHANGER

 
  

 
 

H,0 MODERATOR l
PIPE OUTLET

  
   
  

   
  

ACTIVE 1
LATTICE

 

 
 

SECONDARY
COOLANT
LINES

  
 

 
   
  

    

  
 
     
 
  
  
   

    
  

FUEL PUMP
{-in. THICK SHELL HOUSING
CIRCULATING AIR AND
THERMAL EXPANSION GAP
{ .
FUEL EXPANSION /4-in. THICK
TANK, 4 REQ'D
3g-in. THICK ALUMINUM
¥ -in. MINIMUM H,O LAYER AT 300-350 psi AROUND
ENTIRE REACTOR AND HEAT EXCHANGER CASTINGS
FIGURE 5. H,0—-MODERATED CIRCULATING —-FUEL REACTOR WITH A TANDEM HEAT

EXCHANGER

Gt
36

reactor and heat exchanger were also increased. A light shield at the
crew was then used o bring the total dose in the crew compartment down
to thetolerance of 1 r/hr at full power output. The total weight of reactor,
heat exchanger, and shield was found to be 156, 000 pounds.

Since the above shield seemed to be excessively heavy, an attempt
was made to reduce the thickness of lead in the reactor shield and increase
the amount of shielding at the crew. It was felt that even if a mechanic
could not remain in the vicinity of the reactor for an extended length of
time it would be a tremendous advantage to be able to carry out a limited
amount of airplane maintenance work. The bulk of the maintenance work
would then be carried out after the fuel and secondary coolant had been
drained, and yet there would be an opportunity for checks, adjustments,
and other tune-up work as the power plant was being warmed up. Accord-
ingly, a shield was designed that permitted a dose of something like 1 r/hr

at 50 ft from the center of the reactor at 5% of full power output. In this

design some of the lead and water was removed from the reactor and heat
exchanger shield evolved for the preceding design and the shield thicknesses
at the sides andrear of the crew compartmentwere increased accordingly.
The total weight of this reactor, heat exchanger, and shield was found to
be 141, 000 pounds.

The design study covering an airplane in which the circulating fluoride
fuel was carried directly from the reactor to the engine air radiators uti-
lized a reactor-to-crew separation distance of 120 ft and a smaller crew
compartment than that employed in the shields of this study. To get some
idea of these effects a new crew shield was designed to go with the first
reactor shield design. A greater reactor-to-crew separationdistance re-
duced the dose at the crew by a factor of about 6, and the smaller size of
the crew compartment gave another factor of nearly 2 in crew-shield sur-
face area. This made possible a saving of 22,000 lb in the weight of the
crew shield.

Annular Heat Exchanger and Shield Design

 

Since much of the weight in the tandem heat exchanger arrangement
seemed to arise from the large surface area of the long pressure shell of
the reactor-heat exchanger combination, the more compact arrangement
inherent in the annular heat exchanger arrangement seemed promising.
'The layout shown in Figure 6 was prepared for a sodium hydroxide moder-
ated and reflected reactor. Critical mass calculations indicated that even
with only 25 vol % sodium hydroxide in the core the critical mass of an
unpoisoned 25 in. dia sphere would be about 35 pounds. In the design
shown, a series of concentric helical coils of tubing were provided to
carry the sodium hydroxide through the reactor core with a circulation
NaK INLET

NaOH -NaK HEAT EXCHANGER

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CIRCULATING—FUEL REACTOR WITH AN ANULAR HEAT EXCHANGER

g
38

rate high enough through both core and reflector to keep the outlet temper-
ature to 12500F with an inlet temperature of 1050°F. A special small
heat exchanger was placed in the inlet manifold for the NaK of the sécond-
ary circuit to provide cooling of the sodium hydroxide, The reflector was
made 11 in thick and was separated from the heat exchanger by a 1 in. thick
layer of boron carbide to attenuate the neutron leakage flux from the reactor
to the level of the delayed-neutron flux from the fuel in the heat exchanger.
As in all these heat exchanger designs, the center tube of every square
cluster of nine was a '"dead' tube filled with boron carbide and sealed off

at the ends.

The shield designs were similar to those prepared for the tandem
heat exchanger except for their shape and reduced volume. It was found
that both the delayed neutrons and the decay gammas from the fuel in the
heat exchanger were attenuated more rapidly in passing through the shield
than the prompt neutrons and gammas released in the core. Thus, the bulk
of the radiation leakage from the shield surface came from the core instead
of from the fuel in the heat exchanger. In fact, the material in the heat ex-
changer was moderately effective as shielding material (roughly 60% as
effective as water); hence, the heat exchanger actually at least partly ''paid
its own way'' instead of being completely parasitic.

Two by-products of the design study with the 11 in,thick reflector fol-
lowed by 1 in. of boron carbide appear to be so important that they alone
would justify the use of a thick reflector. If a considerable amount of boron
is provided in the shield water to suppress secondary gammas from neutron
captures inhydrogen, the indications are that with a thin reflector the steel
pressure shell becomes a more important source of gammas than the re-
actor core, so far as leakage from the outer surface of the shield is con-
cerned. Closely associated with this is the fact that with a thin reflector
the problem of gamma heating of the pressure shell is quite serious, where-
as with the arrangement used it was not. Calculations indicate that a re-
flector thickness equivalent to 10 in. of water followed by 1 in.of boron car-
bide is the minimum that should be provided from these two standpoints.
The layer of boron carbide would, of course, require careful and thorough
cooling by passing sodium hydroxide or some other coolant through tubes
imbedded in the canned boron carbide.

The weights of the reactor-annular heat exchanger-shield combinations
were foundtobe 98, 000 and 128, 000 1b for the divided and near -unit shields,
respectively, a saving of 13, 000 lb over the tandem arrangement for the
divided shield and 28, 000 1b for the near-unit shield. Fairly complete
specifications are presented in Table 6.
39

v,

Spherical Heat Exchanger and Shield Design

 

The large saving in shield weight effected by going from the tandem
to the annular heat exchanger arrangement was most encouraging. It im-
mediately suggested that a further substantial reduction in weight might
be realized by going to a more nearly spherical configuration. A novel
heat exchanger in which the tube matrix was laid out in the form of a spher-
ical shell was devised toform a basis for further designs. The basic con-
figuration envisioned is shown in Figure 7.

The heat exchanger tubes were grouped in bundles. Each bundle had a
rectangular cross section except where the bundles terminated in circular
disk headers. If the two pumps are taken as being at the North and South
poles respectively, each tube bundle would lie on a variable pitch helix
running from approximately the Arctic to the Antarctic Circle. The angle
between tube centerlines and the Equator would be about 40 deg, whereas
the corresponding angle at the Arctic circle would be about 90 degrees.
This would make it possible to keep the center-to-center spacing of the
tubes independent of latitude. Although the tube bundle shape is quite
unusual, fabrication of the heat exchanger should not be too difficult. After
jigs had been prepared for building up a tube bundle, one tube bundle after
another could be assembled, brazed and/or welded into its header disks,
pressure tested, inspected, and assembled in the spherical steel shell. A
simple fillet weld between the header disk and the pressure shellwould suf-
fice to complete the installation.

In other respects the heat exchanger and shield design of Figure 7 is
essentially similar to the tandem and annular arrangements described pre-
viously. The 11 in. thick reflector of the annular heat exchanger design was
used again, as was the same type of shield construction employed in both
of the other designs. One difference was that boron carbide was placed in
the baffles between tube bundles instead of in ''dead' tubes.

The reactor design is suggestive of a homogeneous reactor, which it
really approaches quite closely. The calculatedcritical mass proved to be
reasonably low—only 20 pounds. Credit for this must go to the thick
beryllium oxide reflector. The latter would be cooled by a nonuranium
bearing fluoride salt. It should be possible through careful design of the
moderator cooling passages and the diffusing vanes at the reactor core inlet
to keep the metal temperature at all points below the average temperature
of the fuel leaving the reactor. Thus, the hottest metal anywhere in the
system would be the tube walls at the fuel inlet end of the heat exchanger.

As shown by the data in Table 6, the sphericalheat exchanger designre-
duced the over-all weight of the reactor-heat exchanger-shield combination
for the design conditions to only 90, 000 1b for the combination using a divided
shield and 120, 000 1b for the near-unit shield, a savings of roughly 8000 lb
in both instances over the annular heat exchanger arrangement.
Pt Lo Saarr

   
 

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182~Ld-A

AEQUT
PART 11

THE CIRCULATING-FUEL AIRCRAFT REACTOR EXPERIMENT
43

INTRODUCTION

The ARE is areactor experiment whichhas as its fundamental objective
the acquisition of useful and reliable data that cannot be obtained faster and
cheaper by any other means. Accordingly, these data must be of such
nature that they will expedite the development and fabrication of the aircraft
reactor in such a manner as toreduce the aircraft reactor cost, in time and
money, below whatthis cost would be without the ARE. It is the opinion of
the members of the ANP Project of the ORNL that the aircraft reactor will
- most economically be attained through the information to be obtained from
this Aircraft Reactor Experiment. Indeed, as the task of flying a super-
sonic airplane on nuclear energy is so exceedingly complex and the nec-
essary data are so meager, more than one experimental reactor may be
required before sufficient information and experience are obtained to per-
mit the design and construction of a reactor for a useful airplane.

The experimental reactor is, as nearly as is practical, a prototype of
350-megawatt, BeO-moderated, circulating-fuel aircraft reactor..The
core is approximately a 3-ft-square beryllium oxide cylinder in an Inconel
shell. The fuel is circulated through the central portion of the beryllium
oxide moderator and thence to anexternal heat exchanger, pump, and back
to the core. While in the core, the fissionprocesses raise the temperature
of the fuel by several hundred degrees to 1500°F. The reflector and pres-
sure shell are cooled by a secondary system independent of the circulating
fuel. This basic nuclear design of the ARE is capable of developing many
times the power output at which it will be run in the experiment.

As the primarypurpose of the ARE is to obtain experience in the high-
temperature range with a nuclear power source, the reactor will be oper-
ated at design-point power for as much of its lifetime as operating diffi-
culties permit. However, it is planned to limit the operational life of the
reactor to about 1000 hours, but with as much time at design power of 3
megawatts as possible., Other critical parameters of the experimental
reactor are as nearly in one-to-one correspondence with aircraft reactor
specifications, as practical considerations permit. For example, the
maximum temperature in each is 1500°F and the temperature drop across
the ARE is 350°F, while that in the aircraft reactor would be around 500°0F.
No attempt, however, will be made to make ausefuldisposition of the output
power which will be dumped directly as heat energy. It is ultimately car-
ried away by a waste water system.

Although most of the thermodynamics of the fuel can be evaluated with-
out the use of a chain-reacting system, there is no way of simulating the
heat production of the fission process. Therefore, the power-production
44

and heat-removal functions of the experimental reactor will be as thor-
oughly studied as the operation time of the experiment permits. Another
most important aspect of the ARE is the information to be obtained on the
control and stability of such circulating-fuel reactors. Answers tosuch
questions as effecton reactivity of temperature fluctuations in the fuel and
effect on stability of the "lost' delayed neutrons may then be ascertained.

Obvicusly, to build a circulating-fuel reactor requires the existence
~of a liguid fuel and structural metal, both capable of prolonged operation
in a mutual environment at high temperatures and radiation fluxes. In
addition, these materials must both possess certain ratherrestrictive
physical, chemical, and nuclear characteristics. In particular, the fuel
must have a low viscosity and a reasonably high heat capacity to effect the
desired heat removal from the cores. A fuel which seems to possess the
minimum requirements in these categories is the system of fluorides of
sodium, potassium, zirconium, and uranium. This fuel has been studied
to a considerable exient, and the data on the fuel characteristics indicate
that it can be satisfactorily employed in the ARE.

The fuel tubes and structural members of the reactor will be fabri-
cated of Inconel, as this metal appears to possess the best combination of
the required characteristics. Even Inconel, however, does not possess
entirely satisfactory corrosion resistance, although there is every indica-
tion that the dynamic corrosion of this metal by the fluoride fuel may be
improved. Otherwise, the metal, the fuel, and the beryllium oxide mod-
erator possess the desired high~temperature characteristics.

Experiments have shown that the proposed materials suffer no radiation
instability at the intended operating flux levels of the ARE. However, there
is some uncertainty regarding the existence and extent of the radiation-
induced corrosion at neutron fluxes approaching that to be expected in the
aircraft reactor. |

Alrcraft reactors would require minimum structural thicknesses within
the reactor, as well as componcnts with the minimum size and weight. At
the present time, however, information on corrosion rates and properties
~of structural materials under the proposed operating conditions, is too
meager to permit the incorporation of aircraft-type mechanical features
into the ARE without jeopardizing the entire experiment. Accordingly, in
instances where the relevance of the experiment is riot unduly compromised,
the design of ARE components has emphasized durability and reliability.

In general, therefore, the design of the ARE is analogous to that of an
engine dynamometer test cell, in that in both instances the relevance of the
item under test is preserved, whereas the power absorption equipment and
other accessories are not intended to simulate those of the ultimate applica-
45

tion. Thus, insofar as possible, the experimental reactor is a low-power
3-megawatt prototype of the 350-megawatt circulating-fuel reactor proposed
for supersonic aircraft propulsion.

Summarily, the ARE project stands at a crucial point in the ANP pro-
gram where some five years of analytical work have failed to establish the
impossibility of achieving nuclear-powered flight, but where gaps in avail-
able data do not permit the analytical proof that success can be attained,
The purpose of the ARE is tofurnish some of the crucial data that are lack-
ing so that with these data the solution of the problem can go forward, If
efficient use is to be made of the substantial effort being expended in the
ANP program, the expeditious completion of the ARE program is impera-
tive. It is the purpose of this report to describe the design and operation
of the facility and as accurately as possible to establish the time schedule
for accomplishing the Aircraft Reactor Experiment. The design data for
this reactor are given in Table 7. |

TABLE 1

AIRCRAFT REAC TOR EXPERIMENT DESIGN DATA

Heat flux (Btu/hr/sq ft)

Power (max/avg)
Power density, max (kw/liter of core)
Specific power (kw/kg of fissionable material)

~ Materials and Amounts -

Fuel

Uranium enrichment (% U235)
Critical mass (kg)

Total uranium inventory (kg)
Fuel elements

Fuel-element jacket
Moderator

General
Location Oak Ridge
Operator ORNL
Purpose Experimental
Neutron Energy Epithermal
Status Design
Heat, Max (kw) 3000

Heat transported out by
circulating fuel

~~2:1

5

< 400

NaF-KF-ZrF4-UF4

93 .4

~12.5

~65

66 parallel tubes (each 3 ft long, 1.235
in., OD, 60-mil wall) containing the
circulating liquid fuel

Inconel

BeO
46

(Table 7 contd.) o -

Reflector BeO

Shield Concrete .
Primary coolant The circulating fuel

Reflector coolant NaF-KF-ZrFy

Circulating Fuel-Coclant

Maximum fuel-coolant temperature (OF) 1500
Maximum sheath temperature (CF) 1510
Maximum Inconel temperature (OF) 1510
Maximum moderator temperature (OF) 1675
Consumption at max power (g/day) 3.0
Design lifetime (hr) 1000
Burnup in 1000 hr at max power (%) 0.25
Inlet temperature (°F) 1150
Outlet temperature (°F) ' 1500 -
Flow velocity (ft/sec) 4
Flow velocity (gpm) 83
Pumping power (hp) 10

Neutron Flux Density (Avg)

Thermal-max (n/cm2/sec) 3 x 1013 .

Thermal-av  (n/cm2/sec) 1.5 x 1013 "

Fast-max (n/cm2/sec) 7T x 1_013 '

Fast-av (n/cm?/sec) 3 x 1013 , ]

Intermediate -av (n/cm2/sec) 4 x 1013 L :
Dimensions

Core 33 in. dia, 35-1/4 in. high

Reflector thickness 7-1/2 in. on side, ends open

Shield thickness ' Approx, 7-1./2 ft concrete

Over-all (reactor and heat-exchanger pits) 42 ftwx85ft 1x28 fth
Contol

Shim control Increase UF4 concentration in fuel

Regulation ' One B4C absorber rod (2 in. OD

1-1/4 in. ID)
Safety Three B4C absorber rods (2 in. OD,

1-1/4 in. ID)
47

ENGINEERING DESIGN

General Description

 

A briefdescription of the over-all ARE system is given to facilitate an
understanding of the relationship between the various components of the
system, and the components are described indetail. A schematic arrange-
ment of the reactor system is shown in Figure 8.

Heat is generated within the circulating fuel as it is pumped through
the beryllium oxide-moderated core. The core is provided with a beryllium
oxide side reflector that is housed, with the core, within an Inconel pres-
sure shell. The reflector and pressure shell are cooled by a separate cool-
ing circuitemploying a mixture of fluorides containing no uranium. Circu-
lation of both the primaryand secondary circuit fluid mixtures is obtained
by sump-type centrifugal pumps. Theheatis abstracted from the circulat-
ing-fuel outside the core by means of four fuel-to-helium heat exchangers
through which the fuel is circulated. The helium is then cooled by passing
it through four helium-to-water heat exchangers, and the hot water is dis-
charged.

Helium flow rates in the fuel-to-helium heat exchanger are controlled
by variable-speed hydraulic systems that drive the helium blowers. Con-
trol of the helium flow rate in this manner permits smooth control.of re-
actor power at any reactor temperature for which the nuclear controls are
set, within the capacity of the heat removal system. At very low powers
the temperature of the helium passing through the fuel heat exchanger close-
ly approaches the fuel temperature. The system is designed so that when
the rate of power generation is equalto or lessthan the power loss through
insulation, electrically heated helium passes through the fuel-to-helium
heat exchanger.

Leakage monitoring is achieved bythe use of double-walled piping that
provides an annulus through which helium is pumped outside the fuel. The
helium is bledfrom the annulus at established stations and passed through
sensory equipment. Several detectors, including the GE type-H leak de-
tector, are under investigation, but the final decision regarding the type to
be employed has not been made.
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

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FIGURE8. SCHEMATIC ARRANGEMENT OF THE AIRCRAFT REACTOR

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In accordance with current concepts, preheating, as well as the addi-
tion of heat during low power operation, will be accomplished by means of
electrical heaters attached to the core pressure shelland to the outer pipe
of the double-walled fuel piping. In the piping, the helium flow inthe annu~>
lus acts as a heat conveyer, and transports heat tfo sections that are not,
directly in contact with heaters.

Fill and dump tanks for fuel are provided, and will be installed in a
shielded pit adjacent to the reactor and heat exchanger pits. The tanks
have not yet been designed in detail, but provision is being made for heat
addition to maintain the fuel mixture in a molten condition and for removing
from the fuel dump tank the afterheat evolved within dumped fuel after power
operation.

Control of the reactor is obtained by the following features:

1. Regulation-oneboronrod passesthroughthe corecen-
terline with approximately 0. 75% Ak/k.

2. Safety-threeboronrods passthroughthe inner portion
of the core with approximately 5. 0% Ak/k per rod.

3. Shim-shimming is achieved by varying the uranium
tetrafluoride concentration in the circulating fuel.

The external fluid circuit equipment is located in shielded pits adjacent
to thereacfor pit. Two reactor pits are furnished to permit installation of
a second reactor prior to complete decontamination of the pit used for the
first reactor. Theheat disposal equipment is placed in a room outside the
well-shielded reactor pit in order to allow servicing after draining and
flushing fuel. The heat exchanger room is shielded, of course, because of
the activity of the circulating fuel during and after power operation.

Core Design

The ARE reactor assembly consists of an Inconel pressure shell in
which beryllium oxide moderator and reflector blocks are stacked and
through which pass fuel tubes, reflector cooling tubes, and control assem-
blies. Elevation and plan sections of this reactor are shown in Figures 9
and 10, respectively. The innermost region of the lattice is the core, which
is a cylinder approximately 3 ft in dia and 3 ft long. The core is divided
into six 60 deg sectors, each of which includes one serpentine fuel tube coil
that passes through 11 stacks in series, as illustrated. The six serpentine
coils are connected in parallel by means of external manifolds.
DWG,%OZ Ri

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STRUCTURE CORE AND REFECTOR COOLANT

INLET {SHOWN ROTATED 30 DEGREES
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REINFORCING PLATES

FUEL OUTLET
MANIFOLD
ASSEMBLY

CONTROL ROD
ASSEMBLY THERMAL INSULATION

CORE AND REFECTOR
COOLANT QUTLET PIPE
( ROTATED 30 DEGREES
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NOTE :

ALL DIMENSIONS
IN INCHES 2 7.500 34.500

35.250 ACTIVE LATTICE

  
   

 

 

 

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09
51

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PLAN

FIGURE IO

 
52

A reflector with a nominal thickness of 7.5 in. is located between the
pressure shell and the cylindrical surface of the core. The reflector con-
sists of beryllium oxide blocks similar to the moderator blocks but with
0.5 in. holes for the passage of thereflector coolant. Thereflector coolant
is admitted at qne_eyd of the pressure shell, passes through the reflector,
bathes the pressure shell, fills the moderator interstices, and exits at the
other end of the pressure shell.

Primary Cooling Circuit

 

Hydrodynamic and thermodynamic considerations have influenced the
ARE core design to such an extent that an appreciation of these problems
appears tobe desirable priorto description of the primary cooling circuit.
In acirculating-fuel reactor in whichthe heat is generated, and transported
by the fuel, there may be a significant temperature gradient in thefuel and
tube wall. In order toprevent anexcessive fuel tube wall temperature, the
fuel flow in the tubes throughout the core must be turbulent. Analyses in-
dicate that laminar flow would result infuel wall temperatures of the order
of several hundred degrees F higher than the mixed mean outlet fuel tem-
perature. The transitionbetween laminar flow and turbulent flow occurs
in the Reynolds number range 2000 to 5000. Maintaining turbulent flow is
so important that it was thought best to design for a Reynolds number of
10,000 and thereby allow for some deviation in flow rate or fluid physical
properties.

It is inherent in the nature of the circulating-fuel reactors and the
fluids involved that the necessity of turbulent flow imposes more stringent
design requirements on the Aircraft Reactor Experiment than onthe actual
aircraft reactor. The power tobe abstracted from a circulating-fuelreac-
tor is a function of the fuel volume flow—a higher powered reactor results
if higher flow rates are employed. The viscosity of the fuel is such that the
flow rates required by the aircraft reactor result in fluid flow well into the
turbulent flow region. However, since flow rates are reduced to achieve
the desired power level for the ARE, the fluid flow—if straight-through
flow is employed—falls well into the non-turbulent flow region. In order
to maintain turbulent flow, a multipassflow arrangement, as well as a low
viscosity fuel, must be employed.

 
53

If the core isto include fuel tubes parallel to the core axis with a com-
bination of series-parallel connections, then:

R = RQNDL
KV
where R = Reynolds Number,
p = fluid density,
Q = fluid volumetric flow rate,
N = number of tubes in series,

D = tube diameter,

L = reactor core length,
pu = viscosity,
V = fuel volume within reactor core.

Analysis of the relationship between these design variables indicated
that attdinment of the desired Reynolds number would require a few series
passes, even with maximum feasible exploitation of the other variables.
This being the case, and since the design problems involved in the use of
many series passes appearedto be similar to those associated with theuse
of a lesser number of series passes, it was decided to employ 11 series
passes and avoid compromising the volumetric flow rate of the core fuel
volume.

Several means of achieving multipass series flow in the core were
studied. An arrangement consisting of six long tubes, each bent so as to
constitute 11 series passes, appearsto offer several important advantages.
The six bent tubes require a total of 12 end connections to manifolds out-
side the pressure shell. The fuel flow within, or adjacent to, the active
lattice is confined to the interior of constant-diameter circular tubes, and:
eliminates abrupt flow cross section transitions with indeterminate flow
patterns. The six parallel tubes are almost symmetrical relative to the
core centerline, and obviate the need for individual orificing.

The series fuel tube arrangement with vertical tubes involves problems
of gas removal and liquid drainage. Purging of bubbles from the coils has
been investigated experimentally and analytically. Both investigations in-
dicate that at the design flow velocities bubbles and gas traps should be
swept out of the core.
54

Secondary Cooling Circuit

The use of aliquidcircuit within the pressure shell, other than the fuel
circuit, is desired (1) to maintain the pressure shell essentially isothermal
at a temperature of approximately 11500F, (2) to cool the reflector with a
fluid other than circulating fuel, and (3) to fill moderator interstices and
other internal voids with a liquid maintained at a pressure higher than the
fuel pressure to prevent a fuel tube leak from adding reactive material to

the core.

The liquid to be employed must be compatible with the structural ma-
terial under dynamic conditions and should cause no undesirable reaction
with the fuel in the event of a leak in the separating wall. The fluoride
mixture NaF-KF-ZrF4 {mole % 36-18-46), i.e., the fuel carrier without
the UF, is well suited for this purpose.

The possibility of cooling the moderator by positive flow of this fluid
through the moderator interstices has been studied. It has been found,
however, that the mass flow through a gap of constant periphery varies as
the third power of the gap width (in the laminar flow region, with fixed
pressure drop). Temperature analyses with postulated gap width distribu-
tions indicated that gross temperature inequalities could exist throughout
the moderator because of tolerance accumulations affecting interstice
widths. Accordingly, no attempt has been made to have positive flow
through the moderator blocks, but the presence of the 'stagnant' salta-
round the moderator permits moderator heat to be conducted to the fuel

stream.,

Structural Assembly

The mechanical design of the Inconel pressure shell has been the sub-
ject of coordination between ORNL and the fabricator, Lukenweld Co., Inc.,
and it is expected that the shell as constructed will resemble very closely

that illustrated in Figure 9. The serpentine tubes maybe bent in flat planes,.

with welded connections employed from plane toplane, or ""U" tubés welded
to straight core tubes may be employed throughout. These alternatives
have been investigated by commercial, tube-bending concerns, and one con-
cern has contracted to furnish the completed coils.

Core Temperature Distribution

The circulating fuel in the ARE reactor core flows in paraliel through
six serpentine tubes, and each tube traverses the core 11 times. In pass-
ing through the core, the fuel is heated from 1150 to 1500°F,
55

The two sources of heat for the fuelcirculating throughthe reactor are
internalheat generation - resulting from fission and gamma-ray absorption -
and heat transferred to the fuel from the remainder of the core. The heat
transferred from the remainder of the core is produced both from gamma-
ray absorption in the moderator and the parasitic core material and from
neutron slowing-down inthe moderator. All thisheat is transferred to the
circulating fuel and results in a temperature gradient across the fuel, mod-
erator, and parasitic material.

The resultant temperature profile across the fuel and adjacent mode-
rator block and parasitic material is shown in Figure 11 for both the en-
trance and exit condition. There are four distinct media in each tempera-
ture profile: the fuel, the fuel tube wall, the stagnant salt between the fuel
tube and the moderator block, and the beryllium oxide moderator block.
The maximum tube wall temperature in the reactor is about 15000F.

The temperature differential in the beryllium oxide block is entirely
due to internal heat generation (heat transfer between adjacent beryllium
oxide blocks was neglected). The temperature differentials in the stagnant
salt and fuel tube wall are almost entirely caused by transferring the heat
generated externally, since the effect of internal heat generation is small.
The temperature differential across the fuel, which occurs mainly in the
boundary and buffer layers is principally caused by the temperature differ-
ential associated with transferring into the fuel the heat generated in the
moderator and parasitic material. The contribution to the temperature
differential resulting from the variation of fuel residence times with radius
is only about 5CF. The radial and axial distributions of fission, gamma,
and neutron heating used in this analysis were taken from data reported
in the section on "Reactor Physics. "

External Fluid Circuit

The ARE fluid circuit is intended tohandle toxic and corrosive fluorides
at 1500°F, and requires a considerable amount of design and developmental
effort. Where possible, commercially available components are employed
but pumps, heat exchangers, and certain other components are being con-
structed especially for use with the ARE. The locations of the major com-
ponents of the system, includingthe reactor pits, heat exchanger room, re-
actor, and two heat disposal loops are illustrated in Figure 12. The heat
exchanger room is shielded from the reactor pits to permit servicing after
fuel drainage and flushing. The shield is designed to minimize activation
of fluid circuit structure by reactor neutrons during power operation and
to attenuate reactor post-shutdown gammas to a level permitting access to
fluid circuit components after shutdown.
 

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FIGURE 1. TEMPERATURE PROFILE IN FUEL TUBE AND

MODERATOR
 

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PRIMARY COOLING

——
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LEGEND SYSTEM NOQ, 2
—— - s == e COQOLANT FROM SOURCE
— v sttt = ———— COQLANT TO SCURCE

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FIGURE 12. FLUID CIRCUIT EQUIPMENT ARRANGEMENT

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58

Primary Cooling System

 

Fuelflows from the six core tubes to a common header, and a common
trunk line conveys the fuel into the heat exchanger room. The fuel then
divides into two completely independent loops, each capable of dissipating
1500 kw. Within each loop the fuel divides into two parallel 750 kw fuel-to-
helium heat exchangers, passes through a centrifugal pump, and then to a
common return trunk. A line diagram of the primary cooling circuit is
shown in Figure 13.

In an individual loop, helium flows from a centrifugal blower through
a fuel-to-helium exchanger in which its temperature is increased from
250 to 750°F. After the helium is cooled to 2500F inahelium-to-water ex-
changer, itpassesthrough another heating and cooling cycle and is returned
fo the blower. This so-called 'double-sandwich' arrangement permits two
helium heat transports per cycle and thereby halves the helium flow rate
required through the blower and ducting for a specified power and temper-
ature rise,

Secondary Cooling System

 

The reflector andpressure shell cooling system, as in the case of the
fuel system, is divided into two heat disposal loops, each capable of han-
dling the load associated with 1500 kw reactor power. Heat is conveyed
from the primary coolant to helium and then to the water sink. A line
diagram of the secondary cooling circuit is shown in Figure 14,

Monitoring Circuit

 

Alllines and components containing fuel or reflector coolant are double-
jacketed and helium passes throughthe annulus. The helium pumping head
is maintained by drawinghelium from the system, cooling it, and admitting
it to rotary-type compressors at various points inthe system. Monitoring
for fluoride leaks into the helium is achieved by passing helium samples
through halogen detectors.

Pumps

The pumps to be used in the fuel circuit and in the moderator coolant
circuit are vertical-shaft, tangential-discharge, centrifugal pumps employ-
ing a gas sealand a maintained liquid level. A pump of this type, designed
specifically for the ARE, will provide flow of 41 gpm with a 40 ft head (Fig-
ure 15). Electromagnetic pumps cannot be used because of the poor elec-
2o TR e o
PIEL Y Py
:

59

 

 

To Dram
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FIGURE 13. PRIMARY GOOLING CGIRCUIT

 

 
60

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UNCLASSIFIED
DWG. ESK-15329aR1

oiL ouT

 

|. SHAFT, TYPE—316 STAINLESS STEEL. 8. SHAFT LINER, ANY 300 SERIES STAINLESS STEEL.
2. SHAFT LINER SPACER, ANY 300—SERIES STAINLESS STEEL.

9. SINGLE-ROW RADIAL BALL BEARINGS
3. "0" RING, MATERIAL TO BE RECOMMENDED BY SEAL MANU- 10. SLINGER, TYPE- 3i6 STAINLESS STEEL
FACTURER.

11, SPEGIAL GRAPHITE RING ,ALTERNATE WITH SEAL
4. IMPELLER, TYPE—3i6 CORROSION— RESISTANT STEEL

ARRANGEMENT.
GASTING.

5. PRESSURE SLEEVE, ANY 300-SERIES STAINLESS STEEL. I2. GASKET , GOPPER .

6. SEAL FACE RING, ONTARIO STEEL. 13. PUMP BODY ASSEMBLY.

7. BEARING HOUSING, TYPE— 316 CORROSION—RESISTANT 14.IMPELLER CASING ASSEMBLY.
STEEL.

15. BELLOWS SHAFT SEAL, FULTON SYLPHON BB8ISR.
6. BELLOWS SHAFT SEAL, FULTON SYLPHON 8816,7R.

FIGURE 15. ARE CENTRIFUGAL PUMP

19
62

trical conductivity of the fluoride fuels. It is possible that a centrifugal
pump employing a frozen fluoride seal=a technique in which fluorides are
frozen about the pump shaft effecting a seal—may be adapted; however, a
gas seal pump has been satisfactorily operated at temperatures above
12000F for several hundred hours.

The gas seal 1sformed by a floating graphite ring located between the
stationary nose on a bellows seal and the hardened rotating nose on the
shaft. The level of the pumped fluid inthe expansion chamber is maintsited
by gas pressure so that the fluid cannot come in contactwith the seal. The
primarypump seal enclosing the bearings carries no appreciable pressure
differential and, consequently, requires that an oil-circulating system a-
bove the seal be pressurized to balance that of the circulating system be-
low the seal. The circulating oil serves to cool the shaft bearing and lower
seal. In addition, internal cooling of the pump structure is obtained by the
circulation of helium.

Another advantage of this pump, in addition to the proved sealing de-
vice, isthatthe entire rotating assembly canbe assembled outside the pump
casing and in event of failure can be installed with relative ease. None of
the liquid lines need be broken to accomplish this replacement nor will any
of the welded joints in the pump permit fuel holdup. All mechanical pumps
that seal directly against system pressure are provided with a monitoring
space to permit the detection of leakage, should such occur. The pumps
will be constructed of type 316 stainless steel except that those parts in
contact with the fluid stream will be of Inconel.

Heat Exchangers

 

The fuel-to-helium heat exchangers are on order with the Griscom-
Russell Co., who designed them to ORNL specifications. The heat ex-
changers are of the cross-flow type with the fuel flowing through 1 in,OD,
0.109 in. wall Inconel tubing. The tubing is finned on the gas side with
helically wound, stainless steel strip. Each heat exchanger includes five
parallel tube banks; each bank consists of seven to nine horizontal tubes in
series with integral return bends. The fuel inlet and outlet headers are
constructed of 4 in.pipe to which the five tube banks are welded.

The helium -to-water heat exchangers also are of the finned-tube cross-
flow type, but steel tubes with copper fins are used.
63

Preheating Sys‘tefn

 

. The relatively high melting point of the circulating fuel and the second-
ary coolant (around 500°C for the NaF-KF-ZrF,-UF, fuel-with a composi-
tion in mole % of 34.7-17.4-44.4-3.5) requires that all equipment within
which these coolants would be circulated be heated to permit loading, un-
loading, and also the possibility of low power operation.

Reactor Assembly

 

The insulation heat-leakage at operating temperature has been calcu-
lated to be approximately 15 kw, and the heat capacity of theassemblyhas
been estimated at 4000 Btu/OF. After having studied various preheating
methods, including pumping hot fluids through the fuel passages or through
the moderator insterstices, it was decided that the simplest method was
the application of electrical resistance heaters to the outside of the pressure
shell, with heat transport to the inner regions by conduction. With a con-
stant heat input of 14kw in addition to the insulationlosses, it is estimated
that the assembly can be heated from room temperature to 12000F in ap-
proximately four days. |

Piping

The heat capacity of the piping is small relative to the piping-insula-
tion leakage. Consequently, the power required to preheat the piping is
substantially the same as the power required to maintain itat temperature.
If the fluid circuit pipes are empty, as during preheating, the pipe resist-
ance to axial heatflow is high, and the application of heat to specific points
on the pipe surface tends to cause large temperature valleys to exist be-
tween the points of heat application. By attaching electrical resistance
heaters to the outer pipe of the double-walled piping, however, the flowing
helium acts as a thermal conveyer and facilitates axial heat transport.
Accordingly, substantially isothermal helium contacts the inner pipe, and
enables it to be heated uniformly. 3

Heat Exchangers

 

The fuel-to-helium heat exchangers require preheating prior to filling
and require the addition of heat to compensate for thermal leakage during
filling and after filling at any time that the fuel pumps are inoperative.
These heat exchangers have a total of approximately 40 £t of exposed free-
flow area (bothupstream and downstream) that, ifallowed to radiate to ad-
64

jacent cold structure. would dissipate approximately 40 kw with the tubes at
13250F. Accordingly, it appears necessary to include radiation barriers
in the form of gates that can be iowered during warm-up, zero, or low
power operation. When these radiation barriers are in position, the fuel-
to-helium heat exchangers are enclosed and may be preheated by means of
hot helium supplied by the external piping annuli.

Reactor Controls

 

A significant characteristic of the entire ARE-ANP control philosophy
is that the power extracted from the reactor is determined solely by that
part of the over-all power plant external to the reactor. In the power re-
gime, reactor conirol rods do not influence the steady-state power produc-
tion. Conversely. the power extractionfrom the reactor does not influence
the steady-state reactivity of the reactor. A reactor that satisfies the cri-
terion given here is almost certain to have these control characteristics
regardless of whether or not circulating fuel is used.

The fact should not be overlooked that if the reactor behavior for a low
power level differs radically from the behavior of the aircraft reactor,
there is less incentive for building an ARE. Much of the data obtained from
such a low power experiment would be functionally dependent upon power
level. However, the two basic characteristics of the circulating-fuel re-
actor are: (1) the reactor operating temperature is independent of the ex-
ternal pumps and heat exchanger; and(2) the power extracted is independent
of reactor control rod motion.

Essentially then, the objective of the ARE control work is to provide
satisfactory reactor control and external system control for the entire ex-
periment. In addition to creating satisfactory control for the experiment,
the system should be such that control data can be obtained for use in de-
signing the aircraft reactor. Thesedata will be essentially time-response
data for perturbations in reactor temperature and external power load, as
wellas transient and steady-state data on the coupling between reactor and
load.

There are three separate parts to the control system for the ARE: (1)
shim control (2) regulating system, and (3) safety system. Since these
- three divisions perform distinctive functions and since each is almost, if
not entirely autonomous, each will be treated separately here. Reactivity
calculations of the various control mechanisms are reported under 'Re-
actor Physics.' Additional information involving the reactor controls is
discussed under 'Reactor Operation. "
65

Shim Control

The reactor will be made critical only after the fuel-carrying liquid
has beenbrought up to a temperature of approximately 12000F. The liquid
will then be maintained at as nearly constant temperature over the entire
primary circuit loop as the external heating system will allow. After the
system has been checked for leaks and found to be tight, a fuel enrichment
system will be used to bring the reactor to critical at the temperature of
12000F. This fuel-enrichment system comprises the shim control of the
reactor and is designed only to add uranium to the carrier; there is no
plan to try to go subcritical by fuel depletion.

At the surge tank, located between the heat exchanger and the reactor
inlet, a positive-volume fuel-injection system is provided. This fuel-in-
jection operation is accomplished by means of a mechanism shown in Fig-
ure 16. A small fuel line first carries the fuel from the storage tank to
the transfer tank and thence to the surge tank. The enriched fuel is held
in the transfer tank where it is weighed, and when the gas pressure over
the liquid in the tank is raised, fuel is forced through the connecting line
intothe system surgetank. The transfer tank ishung from a beam balance
sothatan accuratedetermination of the weight of fuel is directly obtainable.

This method of fuel addition permits the addition of known amounts of
fuel, and the magnitude of these guantities can be controlled over a con-
siderable range as a critical loading is approached. By introducing the
fuel into the circulating system just before the liquid enters the reactor,
safe loading of the system is assured. Should a slight excess of fuel be
added to part of the fuel stream, the reactor would give an indication of
supercriticality whenthe enriched liquid eniered the lattice and before the
uranium concentration of the entire fuel circuithad been increased. If this
occured, the fuel addition could be stopped and subsequent mixing of the
supercharged liquid with subcritical fuel would lower the reactivity.

After eachfuel injection, which will be made with the three safety rods
cocked and the regulating rod in inid-position, the regulating rod will be
withdrawn in accordance withregular procedure for critical loading of any
reactor. Should this shimmingbe slightly overdone, which would result in
the reactor going on a long positive period beyond the compensating worth
of the regulating rod when fully inserted, the temperature of the fuel-coolant
can be raised enough to restore the reactor tothe critical condition. After
 

  
  
   

 

 

 

 

   
  
    
  

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FIGURE 16 FUEL INJECTION SYSTEM

3 HEATER S

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e o et e P s ae e

67

the loading is accomplished at the starting temperature of approximately
1200°F mean fuel temperature, additional injections can be made when it
is desired to raise the mean fueltemperature. If it becomes necessary to
lower the mean fuel temperature before termination of the experiment, the
safety rods can be lowered slightly into the core. Although not planned in
the operation of the reactor, such action is possible if conditions should
demand.

Regulating System

The reason for providing a mechanical regulating system is that the
magnitude of the reactivity change effected bythe expansion of the fuel can-
not be accurately predetermined. To supplement the negative temperature
coefficient, in case it is of insufficient magnitude, a single absorber rod
having a total value of approximately 0. 75 % Ak/k is provided inthe center
of the reactor.

The absorber rod is located in a 3 in dia hole that runs through the
pressure shell and core in the center of the lattice. By means of a direct
drive, a 2 in.dia boron absorber rod is inserted into this hole. The rod is
moved by one of two possible drive motors. One of the motors is a 60-cycle
reversible motor of constant speedthat is manually actuated by a zero-po-
sition, momentary-contact, reversing switch located on the operating con-
sole. The motordrive is used to make manual adjustments during the crit-
ical experiment and during the rod calibration operation. It is also avail-
able for manual control of the reactor during the power regime.

The other drive, which can be independently used to actuate the reg-
ulating rod, is a d-c servo motor of the same design as that used on the
MTR and also on the MTR mockup - the LITR. 9 This drive can never be
operated manually, and it is actuated at all times by the control function
system in accordance with the operation of the reactor, as subsequently
described. The servodrive will run during the entire reactor experiment.
However, exceptfor the brief period of time required for the dynamic test
of the reactor, the servo will be disconnected from the rod and used only
to drive a monitor selsyn to aid in following the behavior of the reactor
while it is under manual control.

The servosystem is arranged so that the reactivity changes introduced
by the regulating rod behave like a negative temperature coefficient asso-
ciated with the fuel. When the regulating rod is connected to the servo
drive and temperature or flux perturbations are introduced into the reactor,
control of the reactor will be accomplished.

 

9. T. E. Cole, E. E. St. John, and S. H. Hanauer, 'The MTR Safety
System and Its Components,' ORNL-1139, April 25, 1952.
68

During normal operation of the reactor, when power changes are being
made at a slow rate, the reactor will be manually controlled. Improper
manipulation of the manual control, which would result in getting the re-
actor on too short a period, is safeguarded by the cocked safety system
plus the coefficient of reactivity associated with the expanding fuel. Pre-
cautions inthe form of limit switches, interlocks, alarms, etc., in accord-
ance with previous reactor installations, are provided to further minimize
the opportunity for improper manipulation of the manual control.

The regulating rod drives are located on the concrete slab directly
over the reactor and are accessible to personnel at all times. The rod
linkage goes through holes in the concrete slaband provides means for po-
sitioning the absorber portion of the rod in the center hole provided in the
reactor. The general arrangement of the regulating rod and drive mech-
anism is shown in Figures 17 and 18.

Since the regulating rod will run within the core while the reactor is
running at full power, the rod must be cooled. This is accomplished by
the forced circulation of helium in a closed loop down around the rod.

The nuclear instrumentation (fission chambers, ion chambers, etc.)
and the electronic components (preamplifiers, power amplifiers, etc. ) are
copies of similar instruments and gear in use on other reactors. There-
fore it is notnecessary to describe these items indetail. Since the fission
chambers will be located in a high-temperature region within the reactor
reflector, helium cooling will be used to keep the chambers within a safe
temperature range.

Safety System

Three safety rods for the ARE are located on 120deg points at a radius
of 7.5 in. from the center of the reactor. The rods insertinto 3 in.sleeves
that passthrough the reactor shell and core vertically from top to bottom.
The rods are of hot-pressed boron carbide slugs canned in stainless steel
sections slipped over a flexible tube. Sections of the rod are cored to per-
mit passage of helium down through the rod for cooling purposes. Helium
also passes down around the rod in the space between the rod can and the
core sleeve so that cooling of the outside surface and the inside of the rod
is accomplished by a parallel helium flow.

Located above the reactor, on top of the reactor pit, are the actuating
mechanisms for each rod. An actuator is provided for each rod but all
actuators are energized simultaneously and the maxium speed of withdrawal
of the rods is fixed by the speed of the constant speed a-c motor used to
actuate the driving mechanism.

e

 
 

 

69
Y-F7-212, PART 1
<outii.

# GoBER - SECTION SHOWN IN FOLLOWING  ILLUSTRATION

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FIGURE 17. REACTOR CONTROL ELEMENTS
70

M Y-F7-212, PART &I

ey

 

 

LOWER SECTION SHOWN IN PRECEDING ILLUSTRATION

FIGURE I18. REAGTOR CONTROL MECHANISM
71

-The:driving mechanism is comprised of a reversible 60-cycle motor
connected to a drive screw that is arranged so as to drive an electromag-
net in a vertical travel of 36 inches. The drive screw operates inside a
cylindrical enclosure, and the magnet has a keeper that fits as a piston in-
side the cylindrical enclosure. The piston has a small cable attached con-
centrically; the other end of the cable is attached to the absorber rod.
When the magnet is energized, the drive screw moves the piston vertically
within the cylinder and by the cable attachment either raises the safety rod
or allows it to fall under the action of gravity as the piston is lowered on
the drive screw. When the magnet is de-energized, the piston falls and
permits the safetyrod todrophome. Asthepiston reaches its end of trav-
el, gas is compressed in the cylinder to a degree controlled by vent holes
in the cylinder wall and acts as a pneumatic cushion for the rod. Details
of this actuator system and the safety rod are also shown in Figure 17 and
18.

The principle of suspending safety rods on electromagnets has been
used in the MTR, and the ARE designhas taken advantage of the experience
gained inthis reactor operation. A linediagram of the scram circuits with
the actuating signals and interlocks is given in Figure 19. Additional cir-
cuits may be added as the design is completed.

Since false scrams are particularly objectionable to the ARE because
of the limited life of the reactor, every effort is being made to reduce the
probability of their occurrence. The magnet circuits are expected to have
a releasetime no shorter than 0.1 sec, whichwill allow much more positive
operation of the system so far as spurious scrams are concerned. The
design of the reactor with the liquid fuel makes possible this relaxation of
the release time. In the event of a short pile period, the fuel and reactor
core can stand a greater power surge without ill effects than is possible
with solid fuel reactors of more complex core construction.

Each of the three safety rodshas approximately 5% Ak/k negative re-
activity when inserted inthe reactor. Thus a total of approximately 15% is
available to shut down the reactor. Three large rods are used because
statistics indicate that at least one of thethreerodsiscertaintodropwith-
out hesitation at a signal to scram.

Electrical Power Circuits

The extreme inconveniences associated with forced reactor shutdown,
fuel drainage, refilling, and restarting dictate the need of auxiliary systems
so that no one failure will necessitate a forced shutdown. Critical pumps,
blowers, etc.., are duplicated, and each set is served by an independent
72

MOTOR POWER

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Y-F7— 215

 

  

SOURCE IN

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Innumemrl |mzmserr| ML 3 INSERT
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SCRAM scRAM —
FOLLOWER FOLLOWER
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MANLIAL
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INSERT
INTERLOCK
T Mt LmiT

 

 

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FAST ScrAM
N> L5 Ne

1 sec. PERIOD

 

FAST SCRAM
N>LS N

1 SEC. PERICD

  

-~
-
INSBRT 3AFRTY RODS
-
-

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| MANUAL SCRAM

| [mwum. ScpAM ]

 

 

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MASNET POWER

% TO BE CHANGED TO LowW
LEVEL SCRAM FOR START P

X M NEUTRON LavEl.
Np = DESEN PoiNT
NEUTRON Lavel.

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FIGURE 19.

SAFETY

INTER

LOCK

AND SCRAM CIRCUIT
L]

73

b e

d-c circuit iheluding indépendent buses, switches, and a-c¢/d-c motor gen-
erator sets. Theuse of direct current for these circuits permits convenient
speed control where required and the use of battery sets floating on the line
to safeguard against outside power failures. Alternating current instru-
ments are fed from the d-c circuits via d-c/a-c motor generator sets to
enable these instruments to derive their power from the batteries in an
emergency. This electrical power circuit is shown in Figure 20.

~E,
gy
o

The heat capacity of the system is large relative to the heat leakage
rate sothat no large temperature loss will be incurred during a power in-
terruption of 2 to 3 hours. Consequently, it is not necessary to employ
battery power for system heat addition. System heaters therefore are
connected directly to the 440-v a-c circuit. Similarly, the building crane
and certain other items need not be operative during any short period in
which operation from battery power is required, and these componentsare
supplied only by the 440-v a-c system.

Instrumentation

 

Since a basic purpose of the Aircraft Reactor Experiment is the ac-
quisition of experimental data, the importance of complete and reliable

- instrumentation cannot be over-emphasized. Most ARE process instru-

mentation is intended to observe and record rotational speeds, flow rates,
temperatures, pressures, or liquid levels. The various stations and meas-
urements desired around the ARE fluids circuits are shown in Figure 21.
Since most commercially available equipment for performing these functions
is limited totemperatures considerably lower than the minimum operating
temperature of the ARE, an extensive devélopmental program has been
carried out to either produce new instruments or remove the temperature
limitations placed on existing instrumentation. Development has now pro-
gressed to the point where levels can be controlled inhigh-temperature
fluorides with great reliability for extended periods of more than 2000 hours.
Although equipment is still lacking for the calibration of flow-measuring
devices, such devices are now available and have operated for extended
periods. Thedegree of accuracy eventually attainable will depend upon the
procurement of accurate calibration equipment. Pressures are reliably
measured at many points in fluoride forced-circulation systems with both

commercial and locally developed instrumentation.

Flow Measurement

 

Electromagnetic flowmeters have proved to be very reliable when used
with systems circulating sodium but are not applicable with fluorides, which
have a very low electrical conductivity.
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 
 
 

 

 

 

 

 

   

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

    

 

 

 

 

 

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INSTRUMENTATION

CIRCUIT.

FLUID

FIGURE 2]

 

 
i

 
77

A venturi section, with associated pressure sensing instruments, ap-
pears to be the most attractive device for measuring flow in a high-tempera-
ture fluoride system because of low head loss. The extrapolation of water-
calibrated venturi measurements to fluorides is reasonably accurate.

Pressure Measurement

 

Todate, nopressure-sensitive instrument has been developed or found
commercially available that will operate at the temperatures to be encoun-
tered in the ARE fluid circuits. The highest temperature rating of any
commercial instrument known is 450°F. As a consequence, all pressure-
measuring devices have used trapped gas, and as a result the measured
pressure isthat at the gas-liquid interface instead of liquid stream. How-
ever, this error inpressure measurement can be minimized by calibrating
each set of instruments and associated gas traps with accurate weigh-tank
calibration equipment.

Bourdon-tube pressure gages are sufficiently reliable for rough pres-
sure measurements but are notdesirable ifa high degree of accuracy is to
be obtained. Bourdon-tube gages, which indicate the differential pressures
directly, are commercially available.

Null-balance type of pressure gages indicate differential pressure on
a single gage tothe nearest hundredth of a pound per square inch. Although
this instrument is temperature-limited, it was connectedto the above-de-
scribed gas trap tanks and operated successfully at room temperature.

Temperature Measurement

Temperatures to 20009F are measured by means of Chromel-Alumel
thermocouples and recorded on multipoint Brown temperature recorders.
It has been found that the accuracy of temperature measurement depends
to a great extent on the fabrication and installation of thermocouples. Two
methods of thermocouple attachment are used atpresent: (1)pulse welding
each thermocouple wire to the outside surface of pipes or containers, and
(2) forming a beaded junction of the two thermocouple wires. Ineachmethod
the thermocouple is given additional rigidity by tying it down with nichrome
wire. With either of these methods a temperature error is introduced if
the thermocouple is not insulated from ambient air or other adjacent mate-
rial. A borax coating has been found to form an airtight coating that pre-
vents further oxidation.

Tests conducted with thermocouples placed in deep wells centered in
the flowing stream and other couples directly opposite onthe external sur-
78

face of the pipe have revealed that there is no more than a 15 deg temper-
ature drop from the internal thermocouple to that placed on the outside.
Thermocouples made by beading the wires are accurate within £ 10F,

The Brown temperature recorders, with careful adjustment, will re-
cord temperatures accurately within £ 1°F over the full range from 0 to
2000°9F, They do, however, show atendencyto drift after extended periods
of operation and cause errors of the order of 5C0F., This necessitates fre-
quent calibration of the instruments; therefore a program for routine in-
spectionhas been setup that should reveal any instrumentation errors soon
after their occurrence.

Liquid Level Indicators and Controls

Four level-indicating devices have proved satisfactory for use in high-
temperature liquids. The conventional probe type of level control operates
a relay or solenoid when the level of the conducting fluid rises to short out
the probe. Such a level control has proved most satisfactory with molten
fluorides.

Another level indicator using the principle of a resonant cavity has been
designed and simulated experiments indicate that this instrument may give
satisfactory performance.

The variable inductance level indicator consists of small coil wound
on the outside of the tank in which the level is controlled, and a tapered
iron core is mounted on a float inside the tank. The change in level raises
and lowers the tapered core inside the coiland altersthe inductance. Tests
made with the iron core at room temperature demonstrate that it gives a
linear response that canbe directly correlated to a level inside the system.

The fourth level indicator consists of a small tube that extends to the
bottom of the tank in which the levelis to be controlled. Gas flowing through
this tube bubbles through the liquid. The pressure required to maintain a
constant flow of gas through the liquid is directly related to the height of
the liquid level above the end of the tube and is measured by a manometer
connected betweenthe gastube and the gas space above the free surface of
the liquid. Suchan instrument can be used for both indicating and control-
ling liquid levels.

 

 
79

REACTOR PHYSICS

The total uranium investment of the ARE comes out reasonably low,
in spite of the fact that thereactor hasnot beenoptimized to give the lowest
uranium investment. Engineering considerations and expediences designed
to shorten the construction time were in many cases regarded as more
important than optimization.

Little concernis felt about the possibility of self-sustained oscillations.
Because of the low power theywould be slow enough so that a servosystem
could damp them out. Furthermore almost all delayed neutrons are given
off in the reactor, and are available for the damping of oscillation.

Statics

The static calculations are based on the multigroup method, as de-
cribed in references 10 and 11. For reflected reactors this method is at
présent available only for spherical geomeiry. However, the geometry
of the ARE involves a cylindrical core with a side reflector and a partial
end reflector, the side reflector and end reflectors being of somewhat
different composition and thickness. In order to get kg for this geometry,
two reflected reactor calculations for spherical geometry were made.

The first of these calculations used a core of a radius and composition
equal to the radius and composition of the actual cylindrical core, and a
reflector composition and thickness equal to the composition and thick-
ness of the side reflector. The second calculation used a core of the com-
position of actual core, a core radius equal todhZ+ 1 RZ, (where 2h is the
height of the actual cylinder core and R is its radius), and a reflector
corresponding in composition and thickness to the end reflector.

A few bare reactor calculations for spherical geometry yield a curve,
giving ke¢r versus the buckling, or reactor radius. With this curve the
k s values from each of the above reflected reactor calculations were com-
pared. This gave for each reflected reactor an equivalent bare reactor
radius. By subtracting from each of these equivalent radii the corres-
ponding radius assumed for the core, the reflector savings is obtained.
The reflector saving for the side reflector is then added to the cylinder
radius, and the savings for the twoend reflectors are added to the cylinder
axis. An equivalent cylinder is thus obtained, for which a bare reactor

 

10. M. J. Nielson, "Bare Pile Adjoint Solution, ' Y-F10-18, Oct.27, 1950
11. D. K. Holmes, "The Multigroup Method as Used by the ANP Physics
Group, "' ANP-58, Feb. 15, 1951.
80

calculation can easily be performed. A number of these bare reactor cal-
culations for different amounts of uranium yields a curve from which the
amount of uranium required for any given k. ¢ is specified.

The above calculation is carried out assuming homogeneous distribu-
tion of the materials in the core and the same for the reflector. The effects
of "lumping' and self-shielding of the material are computed afterwards.
If they appear important, an iteration of the multigroup calculation is
usually necessary.

The multigroup calculations depend, of course, on the cross-section
data. Fortunately, the ARE has about two-thirds of the fissions occurring
at thermal energy and is, for this reason, not too sensitive to the uncer-
tainties of high energy cross-section data, including the ratio of the fission
cross-section to the absorption cross-section of U?%%, The dataused were
essentially preliminary data of the AEC Neutron Cross-Section Advisory
Group and are probably inagreement with those published in BNL-710 and
AECU-2040. Whether or not any minor discrepancies exist has not been
determined as the publications were just received.

The fissioning power distribution for the equivalent spherical reactor
is obtained from the multigroup calculation. For the cylinder, the radial
distribution in a plane through the center and perpendicular to the axis is
assumed to be the same as that for the equivalent sphere. The distribu-
tion along the cylinder axis is assumed to be a cosine distribution, reaching
zero at a point beyond the core boundary, by a distance given by reflector
savings plus extrapolation distance.

The fissioning power is, of course, proportional to the neutron flux
so that both can berepresented by the same curve. The radial distribution
curve is multiplied by a normalizing factor, so that the total power (or
neutron flux) per cc averaged over the whole core comes out to be one.
That is, the points on the curve are maultiplied by 2w and the radius r.
The result is integrated over r, and this integral is divided by the cross-
sectional area of the core to get the average over the cross-section; this
average is then multiplied by 2/m to get the average in the axial direction
under the assumed axial cosine distribution. The final result is one, due
to the choice of the normalizing factor. The axial power distribution can
be normalized in the same way.

The multigroup calculation alsc gives the neutron spectrum, from which
the power generation by moderation of neutrons can be computed. The
spatial distribution of fissions and absorptions - and hence the sources of
gammas - is derived from the spatial distribution of the neutron flux. The
Y heating may then be determined. The energy contained in the motion of
the fission fragments is given off entirely in the fuel tubes. Neutron mod-
eration creates heat essentially in the moderator. Gamma-ray heat is
generated in all parts of the reactor, in accordance with their energy ab-
sorption coefficients (see for instance ORNL-421).
81

From the neutron spectra, one also obtains the neutron leakage from
the reactor surface, which is of some importance to the calculation of the
activation of external structure and so on. The leakage from the ends is,
of course, obtained from the spherical calculation using the end reflector
composition and thickness, and the leakage from the sides is obtained using
the side reflector data.

A coefficient of reactivity is the change of k with the change in some
parameter of the reactor. For instance, the temperature coefficient of
reactivity is the change of reactivity with reactor temperature. Reactivity
coefficients are obtained by performing the multigroup calculations for at
least two values of the parameters in question, and taking the ratio of the
reactivity change over the change in the parameter.

It should be noted that this procedure will not give the change of re-
activity with fuel temperature correctly if a great fraction of the fissions
or absorptions occur in the high energy ''Doppler' region. However, the
ARE is largely thermal, and no concern was felt regarding the Doppler
effect.

In the past, the multigroup method has given good agreement with ex-
perimental data regarding critical mass. It is not absolutely certain that
this implies that the method will give power and spectral distribution cor-
rectly, in particular since the method is obviously based on a number of
approximations. For this reason, critical experiments, now in prepara-
tion, will be carried out prior to the operation of the ARE.

Reactor Volume Fractions

 

The design data of the reactor are given in Table 7, pp.- 45-46. For
the purpose of the reactor calculation, the total volume of the reactor was
broken down into five regions which are listed in Table 8, together with
their compositions. The "solvent" for the UFy was the fluoride mixture
NaF-KF-ZrF,(mole % 36-18-46 respectively). The fuel now pcstulated
for the reactor has a somewhat different composition, but the computation
can be performed in an analogous manner for the new fuel.

Critical Mass Total Uranium Investment

The kq; required by the ARE is 1.0415 as computed in Table 9. The
calculations described above then show that 27.6 lbs of U2 are required
in the reacting volume. Table 10 summarizes the computation of the ""hest
guess'' for the total uranium requirement, the result being 120 1bs.

The reflected reactor has about two-thirds thermal fissions and a leak-
age-to-absorption ratio of about 1:3.
TABLES

 

SUMMARY OF DESIGN DATA OF THE EXPERIMENTAL REACTOR

No. fuel tubes 66

Core length - 35.250 in,

Core diameter 33.000 in, _

Fuel tube size 1.235 in, O.D. x 0,060 in, wall
No. control rods 4
Item Volume (ft3) Volume (%)

 

Space Between Pressure Shell and Top of Core

 

 

Fuel . 0,134 6.57
Structure . 0,058 2.84
Inert salts 1.815 88.97
Control rod 0,033 1,62

Totals 2.040 100,00

Core

BeO 14.235 81.63
Fuel 1.315 7.54
Inert salts 1.225 7.03
Structure : 0,378 2.16
Control rod 0,286 1.64

Totals 17.439 100,00

Control Rod Structure In Core

 

Structure . b.030 0.17
BeO 0.24 ' 1.38
He 0016 _0.09

- Totals 0.286 1.64

Space Between Pressure Shell and Bottom of Core

 

Fuel 0,134 6.57
Inert salts 1.450 71.07
Structure 0.423 20.74
Control Rod 0.033 1.62
Totals 2.040 100,00
Reflector
BeO 17.947 92.2
Inconel 0,136 0.7
Inert salt 1.246 6.4
Void 0.136 0,7

Totals 19.465 100,0
83
TABLE 9

REACTIVITY EFFECTS PROVIDED FOR BY THE CRITICAL MASS

 

Effect Reactivity (k)
Critical 1.0000
Fission product override 0.0050
Excess for experiment 0.0240
Excess for delayed neutron loss 0.0050
Excess for control 0.0075
Total 1.0415
TABLE 10

URANIUM REQUIREMENTS OF THE EXPERIMENTAL REACTOR

Location Pounds
Uranium in reacting volume 28
Additional uraniuvm in tube bends at ends

of reactor 3
Uranium required in the reactor 31
Total uranium requirement in the system 140

A best guess of the ultimate uranium re-
quirement allowing -2 1/2% in k¢
for critical experiment correlation 120

Reactivity Coefficients

The values of the reactivity coefficients, summarized in Table 11 were
obtained by bare reactor calculation methods, for a fuel slightly different
from that now preferred. The values are, however, essentially correct
for the new fuel.

Power Distribution

The power distribution in the ARE fuel-coolant has been evaluated by
separating the total fission energy into three parts: (1) fission-fragment
energy absorbed in the fuel-coolant (2) fission-neutron energy absorbed in
the moderator and (3) gamma-ray energy from direct fission, fission prod-
ucts, and (n, v ) absorptions. The large energy density is that of fission
fragments in the fuel-coolant. The radial power distribution curve is shown

in Figure 22.

The experimental reactor core has an axial peak-to-average power-
density ratio of 1. 6. Three megawatts in 17. 4 ft3core vclume represents
an average power density of 6 kw/liter, 88% of which is in the fuel-coolant.
The normalizing factor for the radial power distribution curve is 1. 4.
84
TABLE 11
REACTIVITY COEFFICIENTS OF THE EXPERIMENTAL REACTOR

Change of Reactivity Range of
With Change Of Variable Szmbol Value

Thermal base (reactor /1283 to 1672°F %}i 5x 106
 femperature) 68 to 1283°F  Ak/k -0.011
Uranium mass at k=1.0 M 0.4
Am/ U
Coolant density 90 to 100% of Ak/K -0.05
quoted density \AQ/R
coolant
Moderator density 95 to 100% of Ak/K 0.5
quoted density Ap/,o mod
Density of structure 100 to 140% of /A g /K -0.17
(Inconel) quoted density AD/p Inc

Hence the multiplying constant to determine radial power-density distri-
bution (Figure 22) in the fuel coolant is thus 70 x 1.4 = 100 kw/liter. The
power density in the moderator has an average value of 12% of the reactor
power in 83% of the reactor volume.

Neutron Flux and Leakage Spectra

Neutron leakage from the surface of the reactor has been calculated.
Leakage from the side reflector surface in neutrons/sec. cm? is shown in
Figure 23 and leakage from the ends is presented in Figure 24. The rel-
ative importance of the almost-open ends on neutron flux out of the reactor
core may be noted.

At full power, 3 megawatts, the total flux in the center of the reactor
is 17 x 10*3 neutrons/sec. cm® The thermal flux is 3.0 x 10'® neutrons/
sec. cm?, and the fast fluxis 14 x 10*3 neutrons/sec. cm® The non-thermal
flux averaged over the whole core is 7 x 10'® neutrons/sec.cm? and the
likewise averaged thermal flux 1.5 x 10% neutrons/sec.ch. Neutron
flux spectra at three points in the reactor are given in Figure 25. The
important difference is that of amplitude. Note that high-energy neutron
flux is relatively somewhat higher toward the reactor center, but the dif-
ference is small. Figure 26 shows the corresponding plots of flux vs
radius for four encrgies. The importance of moderation by the reflector
is quite apparent.
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

RADIUS/Ar

.6
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uél 5 \\ IN THE FUEL- COOLANT IN
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. |NEUTRON PER cc OF CORE. I
m ,
a
g CORE --{ REFLECTOR
<
0 2 4 6 8 10 |2 14 16 |18 20 22 24 26 28 30

32

FIGURE22. RADIAL POWER DISTRIBUTION IN THE EXPERIMENTAL REACTOR

.

L&
S

» ol

58
LEAKAGE PER UNIT LETHARGY PER

LEAKAGE PER UNIT LETHARGY PER FISSION

 

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A907 L
THERMAL L
004 ESCAPE "-n_l__
AT |
= =
S.003 uz 18.6
o
n
& 002
001 4
0
8 16 14 12 10 8 6 4 2
LETHARGY (u)
FIGURE 23. LEAKAGE SPECTRUM FROM REFLECTOR OF
EXPERIMENTAL REACTOR
016 e
0l4 ]
012
010
| —
008
.00022 r_r—’
THERMAL
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AT n
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0

FIGURE 24.

18

LEAKAGE SPECTRUM FROM ENDS OF

THE

16 14 12 10 8 6
LETHARGY (u)

EXPERIMENTAL REACTOR

4
 

87

ANP—-PHY=-259

 

THERMAL VALUES: AT CORE CENTER

AT —%— RADIUS

AT CORE RADIUS

AT CORE CENTER

FISSION per cc per sec IN GCORE)

AT —'2— CORE RADIUS
AT CORE BOUNDARY

NEUTRON FLUX, nper cm? per sec (FOR AVERAGE OF ONE

O 2 4 6 8 10 12 14 16 18 20
LETHARGY (u)

FIGURE 25. NEUTRON FLUX SPECTRUM AT VARIOUS
| LOCATIONS IN THE EXPERIMENTAL REACTOR
NEUTRON FLUX PER UNIT LETHARGY nper cm? per sec per u

(FOR AVERAGE OF ONE FISSION per cc per sec IN CORE)

3|
@)

H
O,

H
o

Ol
O

W
o

N
an

n
O

! Mev

194

500 ev NEUTRONS

o

)

O
o
N
H

FIGURE 26.

THERMAL NEUTRONS REACTOR 8l

NEUTRONS

‘CORE  INTERFACE

0.4 ev NEUTRONS

 

6 8 0 12 4 16 I8 20 22 24 26 28 30 32
RADIUS/Ar

RADIAL NEUTRON FLUX DISTRIBUTION IN THE EXPERIMENTAL REACTOR

34

0S2-AHd-dNV
 

89
Kinetics

The above temperature coefficients of reactivity refer to slow changes
in temperature. For sudden power surges, only the fuel temperature fol-
lows the power, and onl rAi/f‘- and are effective. The

P Y Am m): A p/pooiant Y
give, together, a negative temperature coefficient of about 10-4/0C. There
appears to be no conceivable mechanism by which a large change of keff
could be introduced very much faster than it can be compensated for by the
control rods. However, if the control rods lagged behind the reactivity
change to the extent that Ak, remained uncompensated, a temperature in-
crease of Ak.s /(107%/°C) would result. For Ak.; = 1%, this amounts to
100°C.

Fast power oscillations would have a period given by 2w}/ \aSP 12
where
g is the reciprocal of the mean neutron lifetime, about 10*4 sec™},
a the temperature coefficient of reactivity about 10 *oc
S the reciprocal of the heat capacity of the total fuel, 6.4 x 10 ¢/
watt sec (assuming a heat capacity per unit volume of about 1
cal/ccOC)

P the reactor power, 3 megawatt

o
This yields a period of 1-1/2 sec, long enough for a servo system to
keep up with the oscillation.

Furthermore, at 4 ft/sec it takes the fuel about 8 sec to pass through
the eleven tubes, each 3 ft. long. Hence, about 3/4 of the delayed neutrons
are given off inside the reactor and the loss of delayed neutrons due to fuel
circulation does not appreciably decrease the damping which the delayed
neutrons afford. Furthermore the circulation of the fuel itself creates a
certain amount of damping, as yet not evaluated quantitatively.

The damping effect due to the circulation of the fuel is illustrated by
Figure 27 takenfrom Y-F10-109, '"Note on the Non-Linear Kinetics of Cir-
culating-Fuel Reactor,' by S. Tamor. The curve illustrated refers to a
somewhat simplified model, having the following properties:

(1) All particles of the fuel spend the same time @ in the reactor.
(2) The flux and power distributions are constant over the reactor;
so that the temperature rise of a fuel particle during a time in-
terval dt is proportional to the total reactor power P(t), the pro-
portionality constant being independent of spatial coordinates; and
so that a temperature change of an element of fuel influences the
reactivity to an extent independent of the position of the element.

 

12. ORNL-63, p. 25, eq. 9.
POWER — Po

24

2.2

2.0

4.0

3.8

3.6

P
———— TEMPERATURE (— 31%9F,

dt
34 ———— POWER (Po)

3.2

3.0

2.8

2.6

 

@
_r
=< &

AVERAGE TEMPERATURE

o J 2 3 4 5 6 T 8 9 10 LI 12 13 14 15 16 I7 18 19 20 21 22 23
TRANSIT TIME —&

FIGURE 27 POWER AND TEMPERATURE OSCILLATIONS OF A CIRCULATING FUEL REACTOR

 
e t—— . —taem——

 

91

(3) A deviation T of the average fuel temperature from the equilibrium
value T = Ocauses instantaneously a reactivity change proportional
and opposite in sign to T.

(4) The reactor inlet temperature is constant.

The numberical values used were: temperature coefficient of reactivity
a = 10™/°C, fast generation time U= 107*sec, reciprocal total heat ca-
pacity of fuel 2.67 x 10°%C/watt sec, average power P, = 3 x 10® watts,
fuel transit time through the reactor 1/8 sec. These values fall in the
region applicable to ANP size reactors, but the general kinetic behavior
is the same for an ARE. The initial conditions were: power equal to P,
at all timesup to 1/10 of a transit time before t = O. Fora period of 1/10
transit time just before t = O, the power is forcibly held at 10 i,. At t>0O,
the system is left to itself. The unit of the abscissa is the transit time 0.
The ordinate on the right side is P in units of Py, and the ordinate on the
left is the average temperature T in units of T/a8. It can be seen that al-
ready the second overswing is much smaller than the first disturbance.
This problem of damping fuel circulation is discussed in somewhat more
general terms in Y-F10-98.

Control

Static calculations of the various mechanisms for the regulation, safety,
and shim control of the circulating-fuel ARE are givenin terms of changes
in reactivity. The one boron carbide regulator rod, fully inserted, effects
a net change in reactivity of 0.0075. The three safety rods each effect a
net change in reactivity of 0. 053. Details of the design of these control
mechanisms are given under "Engineering Design' and their operation is
further discussed under "Reactor Operation. "

Shim Control Requirements

Estimated reactivity changes, Ak/k.y, from room temperature to
1000°F and to an assumed controlled reactor temperature of 1283°F are
given in Table 12. (Though this table was computed for a somewhat dif-
ferent fuel, it is essentially correct for the presently preferred fuel.)

The effect of xenon on reactivity at full power is to reduce the kqfof
the clean reactor by Ak = 0. 0031. Maximum transient xenon reduces K
further by Ak = 0.0082.  The temperature coefficient as a result of the
xenon at full power is(gi‘-é:k)( =+6.4x 10 'per OF inthe vicinity of 1400°0F.

€
TABLE 12

SHIM CONTROL REQUIREMENTS

Z6

A k/k g for Change

 

 

Effect Assumption

 

 

Expansion of liquid fuel  Volume expansion coefficients
per OF: fuel, 1.67 x 10~% (melt-
ing ignored). Inconel, 0.20 x

1074, Difference, 1.47 x 1074,
Radial expansion determined
by inconel, axial by BeO.

Linear expansion coefficients
per °F: Inconel, 6.7 x 1075,
BeO, 4.9 x 1075, Weighted |
average expansion coefficients
linear, 6.1x 1078 volume.

18.3 x 1075 per ©°F. )

Dimensional expansion
with constant material
density

Change of BeO density

—

Change of density of
inconel in core

Inconel volume expansion co-
efficient per °F, 20 x 1076,

Same as metal cooled ARE
(See Y-F10-77),

Change of density of
reflector BeO

Change of cross sections
with reactor temperature

Except Xe cross section.

Change of Xe cross
section with reactor
temperature

0.26 of decay products in the
core, remainder in external
circuits.

TOTAL

Equation for (Ak/K.¢¢)/AT

 

-1.47 x 10-4| Ak/Kk
(flm/m)U

-6 Ak/k
6.1x 10 OR/B)

corec

- -6 dk/k
18.3 x 107 ¢ 80,4

-6 Ak[k
-20 x 10
(Ap/p)lnc

Ak/k

A k/k

 

A k/k

from 100  from 1000 of i°F at
to 1000°F 10 1283°  oper. temp.
- -0.046 -0.015 -5.2x 107°
( p/p)coolant
+0.002 +0.001 +0.3 x 1079
-0.008 -0.003 ~0.9x 1070
+0.003 +0.001 +0.3x 10°°
-0.005 -0.002 0.6 x 10°°
-0.027 -0.005 ~0,2 x 1072
+0.1x 10~°
0.081 -0.023 6.2 x 107°
93

Regulating Rod

In the poison control system there is one axialboron carbide regulator
rod lying along the longitudinal axis of the cylindrical reactor. The per-
manent reactivity effect of the regulator structural material is included in
the core volume fractions. Detailed calculations have been made for the
sodium-cooled ARE reactor of this reactor control system design. The only
important reactor characteristic that affects the control rod effectiveness is
the neutron spectrum, and this spectrum is quite similar for the sodium-
cooled reactor and the present ARE. Hence, the control rod effect on re-
activity is assumed to be the same. The net change in reactivity when the
regulator rod is fully inserted is 0. 0075.

Safetx Rods

Three safety rods are equally spaced on a 15-in. circle around the
reactor axis. The spacing is sufficiently large to reduce the shadowing
effect of each rod on the others to a relatively small value, so this effect
is not considered. The rodsare 2-in. -dia cylinders of boron carbide with
wall thickness of 0.335 in., and a beryllium oxide rod is attached to the
end of each rod. The normal position of the safety rods is out so that
beryllium oxide will be added to core material with a resulting decrease
in the uranium requirement. The net change in k. g for each rod is thus
the sum of two reactivity contributions: (1) from the removal of beryllium
oxide moderating material, and (2) from the insertion of a 2-in. -dia boron
carbide neutron-absorbing rod. Each boron carbiderod is worth -5. 3% in
Ak/k, and each beryllium oxide rod in the same position is worth +0. 16%.
The Inconel around the boron carbide rods is not an effective poison when
the rods are inserted and results in a decrease in poison rod effect of
0.21%. The net effect of insertion of the three safety rods is thus 15. 75%
In kegs . This value is larger than that quoted for the metal-cooled reactor
becausethere isno poison NaK todisplace. The 15. 75% inkeff corresponds
to about 6 1b of uranium in the core.
94

REACTOR MATERIALS
High-Temperature Liquid Fuels

High-temperature reactors in which the fuel is liquid promise certain
advantages in power output per unit weight, control characteristics, and
ease of fuel recovery over comparable solid fuel models. However, such
designs impose stringent specifications on the liquid chosen as the fuel.
Among the general and specific characteristics of the liquid which must be
considered are:

1. Theliquid must contain sufficient uranium for criticality.
. No elements of high cross section can be present.

3. The melting point must be safely below lowest design
temperature.

4. The vapor pressure must be low at the highest design
temperature.

5. Theviscosity must be low throughout the design temper-
ature range.

6. The coefficient of thermal expansion must be large.

7. Theheat transfer properties (specific heat, thermal con-
ductivity, and heat transfer coefficient) mustbe favorable.

8. The material must be radiation stable atdesign temper-
ature and otherwise suitable structural metals must be
able to serve as container materials.

9. The fuel must be amenable to simple chemical reproc-
essing.

10. Additional advantages may be obtained if the liquid con-

tains sufficient light elements to make it self-moderating.

The recent adoptionof a design in whichthe liquid fuel is circulated to
the heat exchanger effects some considerable engineering simplifications,
particularly in removing the heat transfer system from the core. Such
designs, however, impose additional demands on the fuel system in terms
of efficient start-up of the reactor. It will obviously be advantageous to
melt the fuel to a mobile fluid at the lowest possible temperature to mini-
mize the required preheating of the reactor. In addition, the liquid system
must be chosen so that a dilute (subcritical) fuel solution may be intro-
duced into the core, and small increments of a concentrated solution of
uranium inthe same solvent may be added to bring the system to criticality.
This requires that solutions of widely differing uranium content melting at
temperatures well below the operating range be available and that no high-
melting-point compounds be formed at intermediate concentrations.
95

Liquid fuels possessing the capacity for self-moderation would permit
the simplest reactor design. Although a mixture of sodium hydroxide and
lithium hydroxide will dissolve sufficient uranium, hydroxide mixtures
cannot be contained at reactor temperatures, and adequate quantities of
the heavy lithium isotope are not available. For liquid fuels in which no
moderator is present, solutions of uranium in liquid metals and inorganic
salts are the two outstanding possibilities. Although it has not been pos-
sible to prepare a solution of suitable uranium concentration inliquid met-
als, a number of fluoride salt combinations, all including UFs, have been
developed as potential reactor fuels. Of these, the system NaF-KF-ZrF;-
UF4 appears to best fulfill the required characteristics of the fuel for the
ARE, as itemized above.

Homogeneous Fuels

 

The logical, and indeed almost the only, choice of materials for self-
moderating fuels are solutions of uranium compounds inthe alkali hydrox-
ides. Preparation of such solutions in the proper concentration ranges is
not 2 simple matter. Uranium trioxide seems to be soluble in a mixture
of sodium and lithium hydroxides to an extent that would permit construc-
tion of a homogeneous high-temgerature reactor if the low cross-section
lithium isotope were available.!? Lack of this isotope and a suitable con-
tainer for the resulting liquid does not permit construction of a feasible
reactor at present.

Nonmoderating Fuels

 

From the standpoint of heat transfer properties and probably from that
of minimum radiation damage, solutions of uranium metal inliquid metals
would be a superior fuel. It does not, however, seem feasible to prepare
a solutionwith suitable uranium concentration and especially solutions with
relatively wide ranges of concentrations at the proper operating tempera-
tures in such systems.

Liquid fuels in which uranium compounds are dissolved in fused salts
appear more promising. (Although such fuels seem practical for the ARE,
considerably more experimental effort will be required to demonstrate
their utility in a full-scale aircraft reactor) The restrictions on vapor

 

13. L. G. Overholser, D. E. Nicholson, J. D. Redman, 'Suspensions or
Solutions of Uranium in Molten Hydroxides,'" Aircraft Nuclear Propul-
sion Project Quarterly Progress Report for Period Ending June 10,
1951, ANP-65, p. 96.
96

pressure, thermal stability, and low neutron cross section, as well as
preliminary experiments on radiation stability, suggest fluorides as the
best chemical compound for this purpose. Of the uranium compounds, UFy
seems to offer the most promise although the little-known UF3 might offer
certain advantages. The high melting points of the pure compounds in-
volved seem to be the greatest disadvantage.

A number of inorganic fluorides, notably NaF, KF, BeFy, ZrFy4, LiF,

PbFg, and BiFj, are useful, especially in combination, in lowering the
melting point of UF,. As the data in Table 13 indicates melting points
sufficiently low for reactor operation may be obtained by dissolving UFy
in any of several different systems. Most of these systems, however,
show low melting points only at high uranium concentrations.

TABLE 13
MELTING-POINT DATA FOR POTENTIAL FUEL SYSTEMS

U Concentration Range

 

 

 

Melting below 550°C Lowest Melting Mixture
System {mole % UFy) M. P, (£10°C) Composition (mole %)
NaF-BeFg-UF4 0-15 330 50 - 47 -3
LiF -NaF-KF-UF4 0-6 450 97 (LiF-NaF-KF)* - 3 UFy
NaF-RbF-UF4 24-41 500 45 - 23 - 32
NaF -KF-RbF ~-UF 4 18-41 500 56.3 - 3.7 - 15.0 - 25,0
NaF-KF-UF4 26-30 530 46.5 - 26.0 - 27.5
NaF-LiF-UF4 15-39 450 16.5 - 52.5 - 31.0
LiF-UFy4 ' 24-32 480 73.5 - 26.5
LiF-NaF-BeFg-UF4 0-15 235 49 - 30 - 20 -1
NaF-PbFg -UF4 0-22 430 57 - 28 - 15
NaF-KF-ZrF4-UFy 0-5 ' 425 5-51-42-2

* LiF - 42 mole %
NaF - 11.5 mole %
KF - 46.5 mole %
97

~For the circulating-fuel reactor the desired uranium concentration is
about 4 mole % (15 to 20 1b of U2¥ per ft3). Thus, many of the potential
systems are not suitable with high U239 enrichment.l4 Use of more nearly
normal uranium would result in increased critical masses and would prob-
ably be feasible at reasonable penalty only for the NaF-RbF-UF4, Li'F-
NaF-UF,, or Li'F-RbF-UF, systems,

* Phase equilibria in the system NaF-BeFo-UF, (Figure 28) show this
system to have a wide range of uranium concentration even at melting
points as low as 350°C. However, because of its relatively high viscosity
the system may not be used in the ARE although it could be advantageously
employed in an aircraft reactor system.

The most promising fuel for the ARE seems to be obtained by addition
of UF4 to suitable compositions in the ternary system NaF-KF-ZrF4 shown
in Figure 29. Mixtures obtained by addition of 5 mole % of UF4 to a melt
containing 46 mole % ZrF;, 36 mole % NaF, and 18 mole % KF melt inthe
range 465 to 4950C. It appears practical, therefore, to start the reactor
by adding small, uranium-rich portions to a dilute, subcritical fuel. The
viscosity of this fuel has been shown to be low enough to permit turbulent
flow in the ARE core system.

In addition, the ternary system NaF-RbF-ZrF, seems to be of con-
siderable value as a fuel solvent. The cross section of rubidium is mark-
edly lower than that for potassium, and the melting points available with
RbF appear to be considerably lower than those for the KF prototypes.
Although RbF is notnow commercially available in quantity ata reasonable
price, it would definitely seem to be worthy of future consideration.

If necessary, UF3 could probably be substituted for UFys with little
penalty so far as melting point is concerned. It should be noted, however,
that if the heavy isotope of lithium were available in pure form, several
additional low-melting-point systems of possible value would warrant im-
mediate consideration.

Chemaical Properties

 

Uranium tetrafluoride is the most reactive of the materials contem-
plated for use in the reactor fuel. This well-known material reacts with
oxygen at elevated temperatures to yield UO9F9 and UFg as the major re-
action products. It reacts readily with water vapor at temperatures above

 

14. For all fuel solutions an isotope concentration of 93. 3% U“° has been
assumed.
98

/

® 680°C
@90 /)\
/ o @ 710°C

1035°C UF,

9%0
/

900

 

 

 

e

INDETERMINATE
BeF,

=

FIGURE 28.

380°C NaBeF;  590°C Na,BeF,

THE SYSTEM NaF-BeF,~UF,

 

Gaaaw
DWG. 14621

995°C
NaF
ZrF4

 

 

 

 

 

 

 

FIGURE 29, THE SYSTEM NaF-KF-ZrF,

(TENTATIVE)

99

 
100

5000C to give HF and UO,. The alkali fluorides are capable of slight hy-
drolysis at elevated temperatures into HF, and the alkali hydroxides and
ZrF, are intermediate between the alkali fluorides and UFy in ease of hy-
drolysis. None of the latter compounds react appreciably with oxygen.
UF; would almost certainly hydrolyze readily and probably would be oxi-
dized simultaneously; the major products probably would be UOy and UFy
if the reaction took place in insufficient oxygen.

It isessential that such reactions be avoided since the oxides formed
would be insoluble in the melt. In addition fo the possibility that the ox-
ides would contribute to the corrosion difficulties, they would certainly
segregate and lead to control troubles during reactor operation. Preven-
tion of oxidation of these materials should present little difficulty since
they would be prepared in a closed system and thoroughly outgassed at
low temperatures before the melting operation. The alkali fluorides and
beryllium fluoride, however, are extremely hydroscopic materials and
cannot be dried in reasonable periods of time by evacuation at low temper-
atures. Elevation of the temperature after addition of UFy would result
in hydrolysis of the material.

It should be possible to dry the materials separately and thenperform
all the operations incompletely dry atmospheres. It seems more reason-
able, however, to mix the materials in a reasonably dry condition and
accomplish the final drying and reconversion to fluoride of any oxide pro-
duced by hydrofluorination at 500 to 600°0C. Under these temperature con-
ditions UQg would be completely converted to UF4by treatment with dry
HF. The acid fluorides of the alkali metals are completely decomposed
at 4000C sothat complete removal of the excess HF would presentno prob-
lem. The purified and oxygen-free fuel would be maintained under a com-
pletely inert atmosphere in its original container and charged as a liguid
directly into the reactor.

Physical Properties

 

Since a final choice of the fuel composition cannot be made until after
the critical experiment, it is not possible to cite specific values for the
physical properties. Several compositions in the system NaF-KF-ZrF,-
UF, are under concerted study at present.

It seems apparent that the vapor pressure of the NaF-KF-Zr¥F,-UF,
system will be low (probably less than 5 mm) at 1500°F. The vapor pres-
sure of ZrF, is estimated in the open literature at 10 atmospheres at
9009K and the vapor pressure of UF4 is known tobe about 6 mm at 12739K.
Sodium and potassium fluorides are even less volatile. However, neither
ZrF, nor UF, exists as such in these fuels. The zirconium is apparently
 

101

combined in complexes of the type NagZrFs and NaZrF;, whereas the ura-
niui” fluoride seems to exist in similar compounds. These materlals,
along with the alkali fluorides. are considerably less volatile. |

The coefficient of thermal expansion has been studied for a number of
fused salt mixtures containing uranium tetrafluoride. The value of3x10"%/
OC should be sufficient for control purposes. The density for a fuel con-
taining 18 to 20 1b of uranium tetrafluoride per cubic foot is about 3.0 g/cc
at1500°F. The density of this fuel NaF-KF-ZrF,-UF, (mole % 34.7-17. 4-
44.4-3.5) is shown in Figure 31.

One of the more important physical properties, sofar as the ARE de-
sign is concerned, is the viscosity of the fuel throughout the temperature
range of operation. Examination of the literature has indicated that the
viscosities of pure fused salts, including fluorides, should be 1 to 2 centi-
poises at temperatures 300 to 4000C above their melting points. Prelimi-
nary experiments with the system NaF-KF-UF, at high uranium concen-
trations indicated that the viscosities were relatively high(2 1/2 to 4 cen-
tipoises at 800°C). There was definite indicationthat evenhigher viscosity
values would be obtained if hydrolysis occurredand if apprec1ab1e amounts
of UO, were suspended in the measured liquid. |

Experiments with various mixtures in the NaF-KF-ZrF,-UF, system
containing-2. 0 mole 9% UF4 have shownthe viscosity to beless than 10 cen-
tipoises at 6250C; these data are shown in Figure 30. Although further
study will be necessary to establish accurate values for the actual fuel
mixture (containing 3.5 mole % UF,) it is evident that the viscosity values
will be within acceptable limits. |

Chemical Processing

 

An attractive feature of the liquid fuel reactors is the ease with which
the fuel may be removed from the core, manipulated by remote-control
devices, and chemically processed to decontaminate and recover the ura-
nium for subsequent reuse. For the liquid fluoride fuels, at least two
simple methods of chemical processing should be feasible. :

The fluoride fuels may be dissolved in aqueous solutions containing
oxidants and agents capable of strong complexing of fluoride ion. The lig-
uid fuel could be drained to a remote, shielded container, cooled, dissolved
inan aqueous solution of aluminum nitrate andnitric acid, and treated with
additional aluminum nitrate to form a solutionfrom which the uranium may
be recovered by extraction with organic solvents. The organic solvent
layer could subsequently be selectively washed to remove various fission
products and finally scrubbed to recover the uranium in aqueous solution.
102

 

 

 

 

 

 

 

 

 

 

 

 

 

 

12
— FUEL GCOMPOSITION
210 NaF — Smole%
0 \ NaF ~ Slmole %
g 8 N ZrF4— 42mole %
';:‘ \ UFs4 - 2mole%
S 6
L
=
s 4
o
?
S 2
0
400 500 600 700 800 900 1000

TEMPERATURE (°C)
FIGURE 30 VISCOSITY OF NaF-KF-ZrFs—-UF4 AS A FUNCTION OF

 

 

 

 

 

TEMPERATURE
wl_—_—
3.4
3.2 A
o
3.0 \GOOQ

 

 

 

FUEL GCOMPOSITION
NaF-34.7mole %
KFq-17.4 mole %
ZrFy- 44.4 mole %

2 2| UF4-3.5mole %

N
o

 

DENSITY (g/cc)
n
®

n
S

 

 

 

P=3.72-0.824x10"3 ¢t

500 600 700 800 900 1000
TEMPERATURE (°C)

FIGURE 3| DENSITY OF NaF-KF-ZrFq—UFs AS A FUNCTION OF
TEMPERATURE

 

 

 

 

2.0

 

 
103

This uranium would, presumably, be precipitated with ammonia or hydro-
gen peroxide, calcined to UsGsor UQOs, reduced with hydrogen to UOg, and
hydrofluorinated to UF,. If all these operations were to be carried out by
means of remote-control devices, as should be possible, a particularly
high degree of decontamination would not be required.

Another possible recovery cycle of apparent attraction involves suc-
cessive hydrofluorination and fluorination of the solidified fuel. In this
method the fuel would be drained from the reactor into storage tanks be-
hind proper shielding, allowed to solidify, and treated with gaseous HF at
elevated temperatures to remove all fission products whose lower valence
fluorides are volatile. At the conclusion of the hydrofluorination reaction,
gaseous fluorine would be admitted at considerably lower temperatures to
volatilize the contained uranium as UFg The sodium, potassium, and zir-
conium fluorides contaminated with the fission products that failed toform
volatile fluorides would be left. The UFg could be reduced directly to UF,
by use of anorganic reducing agentat a moderate temperature inthe pres-
ence of sufficient sodium, potassium, and zirconium fluorides to form a
liquid suitable for reuse as fuel. This set of reactions could probably be
conducted by remote control; it remains to be seen whether adequate de-
contamination of the uranium would be provided.

Structural Metals

The structural metals for the final aircraft reactor must be carefully
selected to satisfy a large number of very stringent requirements. Since
heat transfer will be a major problem, the heat exchanger design must
make use of large numbers of extremely thin-walled tubes. The metal
chosen must therefore be easy to fabricate in reasonably intricate leak-
tight assemblies and must be virtually completely resistant to the fluorides
sofar as corrosion and mass transfer are concerned. In addition, high-
temperature strength, creep strength, etc., of the metal must be satis-
factory in the thin wall sections, and these properties must be unaffected
by the fluorides and their corrosive action.

Not all these problems are of great importance in operation of the
ARE. The heat exchanger system to withdraw power at the modest design
rate need not be elaborate; therefore thin-walled tubing in the heat ex-
changer region is not necessary. Within the reactor the critical mass is
not particularly sensitive tothickness of the metal tube walls; the ARE may
be designed withlarge factors of safety from all points of view and, if nec-
essary, considerable corrosion by the fuel may be tolerated.
104

Molybdenum, which would be superior from a high-temperature
strength viewpoint, cannot yet be fabricated easily. Among the common
high-temperature structural metals none appears clearly superior in
strength, and several of them (Inconel, 400 and 300 series stainless steels)
possess sufficient high-temperature strength and creep resistance to be
suitable, especially if thick-walled tubing is used. Choice of the metal for
the ARE can therefore be made largely on the basis of resistance to corro-
sion by molten fluorides.

Many structural metals, including nickel, Inconel, and the 300 and 400
series stainless steels, have been shown to be adequate containers for the
fluoride mixtures under static, isothermal conditions. Nickeland Inconel,
however, appear to be most resistant to these materials under dynamic
conditions. In 500 hr of exposure, with the heated portion of a convection
loop at 1500 and 250°F temperature gradient, some 3 to 7 mils of pitting
has been observed on an Inconel test loop containing carefully prepared
NaF-KF-ZrF,-UF, mixture. The ARE Inconel fuel tubes with 60 milwalls
should certainly be adequate for the scheduled 1000 hr of reactor operation;
meanwhile, it is believed that this corrosion can be materially reduced.

Static Corrosion

 

Corrosion of a variety of structural metals under static isothermal
conditions was shown to be very slight when the melt consisted of UF4 in
alkali and alkaline—earth fluorides with and without zirconium fluoride.
Mixtures containing lead fluoride were unsatisfactory since structural
metals reduced this compound to metallic lead. :

The metals which proved to have adequate resistance to static corro-
sion by mixtures of alkali, alkaline earth, zirconium, and uranium fluo-
rides are Inconel, nickel, molybdenum, and the stainless steels of the 300
series. The corrosionobserved by metallographic examinationof the cap-
sule wall generally consisted of intergranular penetration of the metal and
attack at the grain boundaries to the extent of 2 mils or less. In these
static tests Inconel seemed to be slightly more resistant to attack than did
the 300 series stainless steels; the 400 series stainless steels appeared
somewhat less satisfactory. On the basis of a few tests, molybdenum ap-
peared somewhat better than any of the metals listed. It appears possible
that molybdenum, when its fabrication is better understood, will be of value
in later and higher temperature models.

The data in Table 14 indicate that at 1500CF, and for intervals of 100
hr, it is possible to contain the fluorides in Inconel or type 316 stainless
steel with little corrosive action.
TABLE 14

105

STATIC CORROSION OF STRUCTURAL METALS BY POTENTIAL FUELS

Fluoride Fuel

NaF-KF-LiF -UF4*
NaF-KF-LiF -UFy4
NaF-KF-LiF-UFy,
NaF-KF-LiF -UF4
NaF-KF-LiF-UF4
NaF-KF-LiF-UF4
NaF-KF-LiF-UFy4
NaF-KF-LiF-UF4
NaF-KF-LiF-UF4
NaF-KF-LiF-UF4
NaF-KF-LiF-UF,
NaF-KF-LiF-UF4
NaF-KF-LiF-UFy4
NaF-KF-LiF -UF4
NaF-KF-LiF-UF4
NaF-KF-LiF-UF,
NaF-KF-ZrF4-UF4™
NaF-KF-ZrF4-UF4
NaF -KF-ZrF4-UFg
NaF-KF-ZrF4-UF4
NaF-KF-ZrF4-UFy4
NaF-KF-ZrF4-UF4
NaF-KF-Z1F4-UFy4
NaF-KF-ZrF4-UFy4

(100 hr Tests at 1500°F)

Containe:

Type-316 stainless steel
Type-316 stainless steel
Type-~316 stainless steel
Type-316 stainless steel
Type-316 stainless steel
Type-316 stainless steel
Type-316 stainless steel
Inconel
Inconel
Inconel
Inconel
Inconel
Inconel
Inconel
Inconel
Inconel
Inconel
Inconel
Type-316 stainless steel
Type-316 stainless steel
Inconel
Type-316 stainless steel
Inconel

Type-316 stainless steel

*  Composition in mole %: 10,9-43.5-44.5-1.1.
#  Composition in mole %: 5.0-51.0-42.0-2.0.

Weight Change

(mg/dm?/day)

-21
-33

+10
+14
+14
+ 1
- 6
-2
-16
-9

Penetration

No attack

No attack

Slight pitting

1 mil

0.5 mil

0.5 mil

1 mil

No attack

No attack

No attack

1 1o 2 mils

1 to 2 mils

1 mil

1 mil

No attack

1 mil

Light pitting
Very light pitting
No apparent attack
No apparent attack
Very light pitting
Very light pitting
Very light pitting
Very light pitting
106

In similar static tests other specialalloys such as Timken alloys Nos.
3 and 6 and Stellite, as well as Hastelloy B, Nimonic, and some molyb-
denum-nickel alloys, have shown good corrosion resistance. The former
alloys will probably be of value in bearings and seals, and the latter may
be shown to be superior structural materials.

Dynamic Corrosion

 

In tests in which the fluorides are circulated through closed loops by
forced thermal convection, however, a definite and more severe corrosion
pattern has been revealed. The loops for these tests have been prepared
by heliarc welding of 1/2 in. —IPS schedule 40 —pipe of the desired ma-
terial. The hot leg of the loop is maintained at 15000F by electrical re-
sistance heaters. Temperatures inthe cold leg of the loops are maintained
at 150 to 1800F below those of the hot leg. The loops are sealed with an
atmosphere of dry, oxygen-free helium above the liquid. Under these con-
ditions of operation the circulation rate is 6 to 8 ft/min.

In initial tests using mixtures of UF; with alkali fluorides only loops
of Inconel and nickel have operated for 500 hr at temperature without dif-
ficulty. Loops of series 400 stainless steel were plugged by the deposition
of high~-melting-point material in the cold leg after less than 30 hours.
Loops of 300 series stainless steel showed similar but less rapid plugging,
and circulation was stopped by this phenomenon after 100 to 150 hours.
Examination of the Inconel loops after scheduled termination of the tests
indicated some mass transfer of metal from the hot to the cold portion of
loops, but the quantity was insufficient to cause failure of the loop.

Solutions of UF, in NaF-BeF, have been shown to be somewhat less
corrosive than similar solutions of this material in NaF-KF-LiF mixtures.
In general, however, the NaF-KF-LiF mixtures without UF4 have been
less corrosive than either.

Addition of small quantities of NaK to the NaF-KF-LiF mixtures (with-
out UF,) considerably improved the corrosionresistance of structural met-
als. Tests of 500 hr duration in loops of stainless steel have indicated
virtually negligible corrosion; photomicrographs of hot-leg sections of
loops exposed to NaF-KF-LiF mixture with and without added NaK are
shown in Figure 32.

Recent tests in which solutions of UF, in NaF-KF-ZrF, mixtureshave
been exposed for 500 hr in Inconel yielded considerably more favorable
results. The hot legs of the loops showed scattered pitting to a depth of 3
to 4 mils with infrequent intergranular attack to a depth of 7 mils (see
Figure 33). A thin (perhaps 1 mil) nonmetallic layer was observed in the
 

  

Foar
i\,

& UNCLASSIFIED
PHOTO NO T-934

 

   
   
 

    

       

 

: x.’; M‘ “: _7 ,)/’; ; o 5 . . i i Lo A R

CORROSION OF HOT LEG OF INCONEL LOOP AFTER {000 HOURS B

NaF-KF-LiF(11.5-42.0-465 MOLE %) 250X
UNCLASSIFIED

PHOTO NO.T-966

   

'cofinosmN OF ”HOT LEG OF INCONEL LOOP AFTER 500 HOURS BY
NaoF—KF—LIiF (11.5-42.0-46.5 MOLE %) WITH 35 VOL.% NaK AEE(I)T)I(VE
FIGURE 32. EFFECT ON NoK ADDITIVE ON THE CORROSION
OF INCONEL THERMAL CONNECTION LOOPS BY NaoF —
LiF-KF AT 1I500° F.

107
108

 

PHOTO NO, T—I1335

.....

 

 

 

 

 

 

 

 

 

 

 

CoLD~1275 °F— LEG OF LOOP 250 X

 

PHOTO NO. T-1334

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

HOT - IS00°F -LEG OF LOOP 250 X

FIGURE 33, CORROSION OF INCONEL THERMAL CONVECTION

LOOP BY NoF-KF-ZrF, —UFs (48—-501-41.3-38 MOLE%)
AFTER 500 HOURS.

 
109

cold legs of the loops; virtually no metallic deposits, such as are produced
by mass transfer, were visible. It is believed that the improved results
can be attributed to the recently initiated techniques for purification of the
liquid fuels before test. It seems certain that the optimum conditions for
such purificationhave not been established and that continued improvement
in corrosionresistance of Inconel to the melts istobe expected. Chemical
analysis of the fuel before and after operation in the loop is given in Table
15.

TABLE 15

CHEMICAL ANALYSIS OF THE NaF-KF=-Z1F4-UF4 FUEL FOR CORROSION PRODUCTS

 

(circulated in an Inconel loop for 500 hr)

 

 

 

Temperature Fe Ni Cr
Origin of Fuel Material (°F) U(%) F(%) (ppm) (ppm) (ppm)
Original batch analysis 75 5.90 39.4 180 4170 50
Hot vertical leg of loop 1550 3.88 40.1 445 30 480
Hot horizontal leg of loop® 1250 to 1500 5.69 39.6 240 20 480
Cold vertical leg of loop* 1250 5.68 39.5 250 40 460
Cold horizontal leg of loop 1370 5.61 39.8 190 30 510

* The results presented are the average of two samples from these sections of the loop.

The very considerable improvement in corrosion by uranium-free
alkali fluorides that resulted frora addition of NaK would seem to offer
another possible technique for controlling the corrosion problem. To the
extent that this improvement was the result of the cleansing action of NaK
on oxidized metal surfaces, and, therefore, to the extent that prior treat-
ment with NaK would effect the benefit, this improvement can be readily
realized. If the mechanism requires the continued presence of elemental
alkali metal there is the possibility of reduction of UFy; and ZrF,4 to com-
pounds of lower valence states and to the metals. Whether the free alkali
metal can be tolerated (and to what extent) in the fuel and what benefit can
be obtained by cleaning with NaK and its removal prior to fuel addition is
under study at present.

The increased corrosion resulting from inclusion of uranium tetra-
fluoride in the fused salt mixtures, if it is real, may be caused by the
oxidizing power of UF4 or more probably some material {(perhaps UOgF:)
added as an impurity with the uranium. If the oxidizing action of UFy is
the cause, then UF3 should be a superior material to use. In the other
event, increasing the purity of the fuel mixture should improve the situa-
tion.
110

In any case, it appears that for the ARE the corrosion that must at
present be considered normal in Inconel for periods up to 1000 hr would
not be prohibitive. Present designs for the reactor call for 60-mil tube
walls in the hot zone. Corrosion at the rate observedat present would not
seriously weaken this structure; it is apparent, therefore that the ARE is
feasible even with the corrosion rates presently observed.

Control Rod Fabrication

 

The control rod system for the Experimental Reactor consists of three
safety rods and one control (regulating) rod. The material for these rods
will be produced by powder metallurgy techniques employing hot pressing.
The safety rods are constructedof a canned Fe-B,C mixture. Theannular
cans are fabricated of type-316 stainless steel and have an inner diameter
of 1 1/4 in. and an outer diameter of 2 inches. The annulus between the
two walls is filled with the Fe-B,C mixture. The cans are 1.95 in. long,
and about 18 cans will be required per control rod assembly. The B4C-Fe
slugs contain approximately 20% by volume iron, which acts as a binder.
The B4C slugs are 1.86 O.D. by 1.26 I.D. by 1.895 in. long.

At this time 20 of the B,C slugs have been successfully hot pressed.
The slugs were fabricated by hot pressing at 15200C under a pressure of
2500 psi using graphite dies. The first few slugs made with this process
cracked longitudinally owing to the difference of thermal coefficient of ex-
pansion of the B,C-Fe mixture and the graphite mandrel. This difficulty
has been overcome by pressing the mandrel out of the slug while at the
hot-pressing temperaiure.

The regulating rod will be made by using slugs of Al,O, impregnated
with B4C contained in one of the 300 series stainless steels. Rods of two
strengths, in which the only difference will be the amount of B4C added to
AloO3, will be made for evaluation in the hot critical experiment. In one
case 0.022 to 0.024 g/cc of B,C will be added to AlyO5 and the other rod
will contain 0. 005 to 0.006 g/cc of B4C with Al,0;. The Al,O; slugs will
have the same dimensions as the B4C slugs except that they will be longer.
The dilute slugs will be 2.14 in. long and the strong slugs will be 2 in.
long. The slugs for the regulating rod being made are hotpressed at 16750C
in graphite dies under pressure of 2500 psi.

Metallographic examination of a slug of Al,0; and B,C has shown that
the two materials are quite compatible at the operating temperature of the
control rods and that there is no undue reaction between this mixture and
the 300 series stainless steel container materials. Some of the B4C-Fe
slugs have been canned in 300 series stainless steeland brazed through the
Nicrobrazing cycle, which consists of heating at 1130°C for 10 min; they
111
were then held at 815°C for 100 hr with slight reaction. Therefore it can

be concluded that there will be no difficulty in the fabrication and operation
of the regulating and safety rods for the ARE.

Stress-Corrosion Tests

 

Any time a structural metal is used in a stressed condition and also
subjected toa corrosive media there is the possibility of a stress-corrosion
phenomenon. Physical tests are being conducted to obtain data that will
indicate whether the structural metal of this reactor, e.g., Inconel, is
subject to this phenomenon when in contact with the fluorides. Two types
of physical tests have been conducted: (1) tube burst tests to furnish
hoop stress data for metals in contact with the fluoride medium and (2)
stress-rupture tests in which the specimen is in a fluoride environment.

In the stress-rupture tests, dead weight loading is transmitted to a
small sample by means of abellows. Elongation is measured by the move-
ment of the end of the load lever arm. The sample is held in a bath of
fluoride maintained at 8160C. In a tube burst test one section of the tube
wall is reduced to a known wall thickness and serves as the gage length.
The hoop stress is produced by placing an inert gas atmosphere inside the
tube while the outside of the tube is bathed in the fluorides at 8169C. This
test only yields time-to-rupture and total elongation data.

Table 16 is a summary of the stress-rupture data being obtained on
Inconel specimens in the fluoride salt, NaF-KF-LiF-UF; (mole % 10.9-
43.5-44.5-1.1), at 816°C. The Inconel is used in the as-received condition.
These data are compared with that obtained from tests using fine-grain
as-received Inconel in an argonatmosphere. As canbe seen, the fluorides
did not impair the load carrying abilities of the materials appreciably.

TABLE 16

INCONEL STRESS -RUPTURE DATA

 

 

Stress Time to Rupture (hr)

{psi) NaF -KF-LiF -UF4 Argon
12,500 6.5 6
10,000 20 - 19
7,500 25.7* 80
5,000 575" 270
4,000 e61** 430

*  Average of three tests.

Test apparatus failed before rupture.

** Specimen ruptured in weld area of grip rather than in test
section.

8
112

In addition to the testslisted in the table, one Inconel sample has been
under a stressof 4000 psi for 350 hr and another has withstood 5000 psi for
525 hr without undue elongation. Furthermore, Inconel tube burst tests
have successfully operated from 500 to 1000 hr at stresses of 2000 to 3000
psi without severe corrosion being induced by the stress.

Radiation Damage

 

Inconel capsules containing several different fluoride mixtures have
been irradiated in both the LITR and the X-10 graphite pile. Radiation
damage studies with the NaF-KF-ZrF,-UF; fuel are not yet available be-
cause of the time required for the analysis of hot specimens. Neverthe-
less, the irradiation of other fluoride mixtures atpower levels inthe range
that will exist in the ARE (approximately 100 watts/cc) have shown no ap-
parent increase in decompositionor corrosion owing to the radiationfield.
However, in the case of one fuel, irradiation at power levels approaching
that in the aircraft reactor have yielded results that may be interpreted as
considerably increased corrosion under strong irradiation, but no increase
was evident in similar tests with a second type of fuel.

Significant data are now available on irradiations of both the NaF-KF -
UF, and NaF-BeF,-UF; mixtures at various power levels. These data are
summarized in Table 17 for several irradiations. Metallographic exami-
nation of the metal specimen and analysis of the fluoride melts for metal
corrosion products are the methods used for detecting radiation damage.
The data from the irradiated specimen are compared with similar data
from a contirol specimen not irradiated but otherwise subjecied to asnearly
as possible identical treatment. Theirradiation intensity is givenas watts
dissipated per cubic centimeter of specimen, since this is more meaning-
ful than neutron flux when irradiating fissionable material.

As indicated in Table 17, exposures of samples of the NaF-KF-UE,
fuel in Inconel at the maximum flux of the X-10 graphite pile caused a
power dissipation of approximately 65 watts/cc of material and have been
quite satisfactory. The irradiated samples showedno detectable gas evolu-
tion because of the irradiation, nor-—after exposures up to 400 hr—was
there any positive evidence of damage. The corrosion of the capsule, as
measured by metallographic techniques, was the same order of magnitude
of that in similar tests conducted outside the radiationfield (Figure 34), as
was the chemical analysis for structural elements in the solidified liquid.

Exposure of similar samples in the LITR at 800°C resulted in power
dissipations of 800 watts/cc of material but yielded results that may be
interpreted as considerably increased corrosion under the strong irra-
diation. Although it appears probable that at sufficiently high radiation
Fuel Composition (mole %)

 

PILE IRRADIATION TESTS ON FUSED FLUORIDE FUELS IN INCONEL

Temperature
{oF)

 

NaF-KF-UF,, 46.5-26~27.5

NaF-KF-UFy, 4b.5-26-27.5

NaF-KF-UF,, 46.5-26-27.5

NaF-KF-UF,, 46.5-26-27.5

NaF-KF-UF,, 46.5-26-27.5

NaF-KF-UF,, 46.5-26-27.5

NaF-KF-UF,, 46.5-26-27.5

NaF-KF-UF,, 46.5-26-27.5

NaF-KF-UF,, 46.5-26-27.5

NaF-KF-UFy, 46.5-26-27, 5%

NaF-KF-UF,, 46.5-26-27.5%

NaF-KF-UF,, 46.5-26-27.5

NaF-KF-UF,, 46.5-26-27.5

NaF-KF-UF,, 46.5-26-27.5

NaF-KF-UF,, 46.5-26-27.5

NaF-BeFz-UFy,

NaF-BeF,-UF,,

NaF-BeFz-UFy,
NaF-BeFy-UF,,
NaF-BeF,-UF,,
NaF-BeF2-UF,,
NaF-BeF;-UF,,

NEF—BEFQ‘UFp

NaF-BeF;-UF,,

NaF-BeF,-UF,,

NaF-BeF;-UF,.

25-60-15

25-60-15%

25-60-15%
25-60-15
25-60-15
47-51-2%
47-51-2%

47-51-2%

47-51-2%

47-51-2

47-51-2

1500

1500

1500

1500

1500

1500

1500

1500

75

1500

Time

{or)

458

458

458

472

299

299

299

344

 

1500

1500

824 (solid)

1500

1500

1500
1500
1500
1500
1500

1500

1500

1500

75

* Pressure tested, no evidence of buildup.

164

161

136

126

131

139

137
137

26
136

143

115

145

TABLE 17

Inconel Components in Fuel
Irradiation

After Test (ppm}

113

 

 

‘watts{cc) Ni Cr Fe Capsule Condition {Photomicrograph)
65 194 194 1,030
65 262 20 1,630
65 20 188 1,130
0 (bench) 160 125 2,630 No apparent attack.
65 93 23z 1,530
65 148 532 2,340 Intergranuiar attack, maximum pene-
tration 1 mil.
65 183 146 622
Q (bench) 93 186 548

.- 123 207 31 Original salt analyses.

800 26,800 1,870 16, 300 Intergranular, plus "diffusion” type of
attack to a depth of 2 to 3 mil.
Discarded because of faulty weld.

g (bench) ) D30 754 1,270

800 45,100 950 6,410 Intergranular attack, not dense, to a

depth of 0.5 to I mil.
0 (bench)y 1,100 645 2,650 No apparent attack.

800 1,380 160 1,220 Very small amount of intergranular at-
tack, approximately 0.5 mil with gen-
erally clean interface.

554

0 (bench) 1,540 7,380 2,300 Subsurface voids or pitting effects 1 to
1 1/2 mil deep.

554 1,490 1,060 1,200 Generally free from attack.

554 Metallurgy in process.

554 450 1,810 T80 Metallurgy in process.

84 5,610 3,900 1,430
0 (bench) 1,120 3,300 811
84 58,500 2,450 3,100 Some coarse intergranular attack ap-
proximately L mildeep. Thecorrosion
interface is generally free from attack.
84 1,880 1,880 2,690
0 (bench) &0 4, 450 445 Subsurface voids or pitting effecis to
1/2 mil deep (effect less dense than in
sample seven lines above).
—_ 50 330 540 This was the original salt analysis be-

fore loading in capsule.
114

 

Y-35460

 

CORROSION OF UNIRRADIATED CAPSULE
AFTER 115 HOURS. 300 X

 

Y-5939

 

CORROSION OF IRRADIATED CAPSULE
AFTER 299 HOURS IN A NEUTRON FLUX
WITH AN EQUIVALENT POWER DENSITY OF
65 WATTS PER GG OF FUEL. 300X

FIGURE34. EFFECT OF NEUTRON [RRADIATION ON THE

STATIC CORROSION OF INCONEL BY NaF-KF-
UF, (46.5—26.0-27.5 MOLE%) AT 1500°F
115

intensity the pure fused salt fuels under static conditions may suffer con-
siderable radiation damage, the above experiment is not conclusive evi-
dence of this since temperature gradients in the fuel or irradiation of im-
purities in the fuel could be the mechanism for the observed increase in
damage. Refined preparation and handling techniques will reduce the im-
purities. It might be possible to reduce the potentially available fluorine
by substitution of UF3 for the UF4 or by dissolving a small amount of an
active metal in the fluoride melt.

In another series of fuel-capsule irradiations the NaF-BeF,-UF, sys-
tem in an Inconel capsule has been exposed inthe LITR to power dissipations
of both 84 and 554 watts/cc of fuel. Inonly one of five exposures(Table 17)
was there any indication of damage beyond that experienced in the control
specimen. The nickel content of the salt washigh, but upon metallographic
examinationfew signs of radiation inducedcorrosion were present. These
particular results are most encouraging since they represent power levels
considerably in excess of that to be realized in the ARE.

Unless studies in progress on the NaF-KF-ZrF,;-UF; system show it
tobe more suspectible than either the NaF-KF-UF4 or NaF-BeF,-UF; sys-
tems, it is not apparent that radiation damage will introduce any consider-
able uncertainty at the fluxes tobe maintained in the ARE. On the basis of
the above static tests, the ARE at this point appears feasible from the
standpoint of radiation damage.
116

CRITICAL EXPERIMENTS

A preliminary experimental study of the nuclear design of the experi-
mental reactor is to be made during the summer of 1952 at ORNL by build-
ing some of the major components into a system that will be made criti¢al
with enriched uranium. The experiments will be done at a power level of
the order of 1 watt, maximum, at roomtemperature and will yield informa-
tion on uranium requirements, control problems, and power and neutron-
flux distributions. It is expected that assembly of the mockup will start
about June 1 and that the system will be made critical about July 15.

Experimental Materials

 

The beryllium oxide hexagonal blocks prepared for the ARE will be
used as the reflector and moderator inthese experiments andwill be trans-
ferred to the ARE later. The fuel is to be a dry-packed powder having nu-
clear properties closely approximating those of the ARE fuel. At present
the latter is designed to be a molten mixture of the fluorides ZrF4, NaF,
KF, and UF,. In the preliminary assembly the fuel is expected to be a
mixture of ZrQOy;, NaF, KF, enriched UF, and graphite. The ZrO; is being
substituted for the ZrF4 because the latter is unavailable at this time; the
graphite is being added to give total moderation equivalent to that of the
fluorine. The composition will be that required to give the design uranium
density of the ARE fuel. It is probable that the relative quantities of the
constituents willapproach ARE design, with the over-all density somewhat
lower. The simulated reflector coolant has the same composition as the
simulated fuel except for the omission of the UFy.

The materials will be packedas a dry powder in stainless steel tubes,
1 1/4 in. in dia and 40 in. long, that can be sealed to minimize moisture
pick-up. The stack of beryllium oxide blocks will be surrounded by an
Inconel shell.

Experimental Assembly

 

The beryllium oxide blocks will be stacked, with axes vertical, in a
right cylinder, 36 in. high and 47 in. in dia;, alsc with its axis vertical.
The fueland coolant mixtures in the stainless steel tubes, will be inserted
vertically through the axial holes in the beryllium oxide blocks. Some of
117

the tubes will be movable by remote operation and will serve as control
rods. A section of tubing containing a nuclear poisonwill be attached above
other fuel elements, and the fuel elements will be magnetically supported.
Upon release by the magnet, the fuel tubes will fall out of the core and will
be replaced by the poison. This safety mechanism will be operated both
manually and by a high-flux-level detection instrument.

Provisions will be made for inserting neutron-detecting foils, such as
indium and gold, through the reflector and moderator. The usual neutron-
sensitive and gamma-sensitive detectors will be placed adjacent to the
assembly for operational control and safety and will feed recording and
indicating instruments in the control room.

Experimental Program

 

The first result, obtained from the initial loading, will be the mass of
uranium required for the assembly to become critical, thereby fixing the
uranium density. After calibration of the control rods, reactivity changes
incurred by the introduction into the reactor of other materials canbe eval-
uated. For example, at least one fuel element designed for the ARE itself
will be used. In addition, the nuclear properties of different types of con-
trol devices can be initially tested. It is intended that extensive studies
will be made of the distribution of neutrons throughout the assembly by the
usual activation method.
118

REACTOR OPERATION

The operation of the reactor may be broken down into three separate
phases, each of varying importance. The start-up phase involves pretest-
ing of the fluid circuit, loading to criticality, and regulating rod calibra-
tion. The subsequent power operation of the reactor following the addition
of enriched fuel material is the most important phase asfar asthe aircraft
reactor is concerned. It is there that the kinetic response of the aircraft
reactor may be determined and all other nuclear experiments are per-
formed. Following the power cperation is the reactor shutdown phase.

The probable elapsed time for each of the steps and sub-steps in the
operational procedure is difficult to predict. Of the 500-odd total hours
available for operation of the system after the original loading with en-
riched fuel, probably not more than 100 hours will be available for power
operation. A careful budgeting of time and rationing of operating hours
is being made and will be revised as zero hour for operation approaches.
The actual number of hours of power operation and the magnitude of the
power cutput from the system cannot be known until the experiment has
become history. However, all pertinent experimentation must be per-
formed during this time.

During the power regime it is planned to run a shielding experiment
completely independent of the primary ARE objective. This will involve
the determination of energies of the delayed fission neutrons. Arrange-
ments are being made to make these measurements in the fluid circuit
external to the reactor pit. The circulating fuel system provides a unique
facility for making such a determination.

Determination of one other basic phenomenon, the evaluation of xenon
or its precursor, will be attempted. This will markedly affect design and
operational considerations for an aircraft propulsive unit particularly as
it relates to behavior after shutdown when xenon poison becomes of major
importance.

The schedule outlined below represents an ideal and almost certainly
it will have discontinuities inthe form of start-ups following scrams, shut-
downs for plumbing and mechanical repairs, etc. Every effort will be
made to minimize these backings and fillings but it is unrealistic to assume
that none will be necessary.
119

Pretesting

After the reactor plumbing hasbeen installed and the system is ready
for checking, both the primary (fuel) and secondary (reflector-pressure
shell) circuits will be given a pneumatic test. These tests will be made be-
fore the annular chaser pipeshave been joined over the welds in the liquid
lines and beforethe heaters and insulationhave been applied to the annular
chaser lines.

Air or other gas will be held at a pressure of approximately 75 psi at
room temperature and all welds and joints of any kind will be gone over
with a bubble solution for the purpose of locating leaks. This technique
should reveal all gross leak faults in both liquid circuits.

After both circuits are tested pneumatically, helium at 75 psi will be
substituted for the gas in the primary circuit and a vacuum will be pulled
on the secondary circuit. The secondary circuit will be monitored for
helium to determine whether or not there is a leak in the primary circuit
within the core.

After both circuits have been made tight as far as the above tests are
able to reveal, they will then be filled with NaK, which will be circulated
at some point near design flow rate. In addition to giving a further leak
test, this will allow checking of pumps, tanks, and level controls under
ambient temperature conditions.

After the system has been checked out with room temperature NakK,
the heaters will be installed, annular joints closed and insulation installed
around all the system. Since this job is quite extensive, it will require
many mandays of effort. When the heating system is installed, both liquid
circuits will be operated at increasingly higher temperature levels until
the system operates satisfactorily at an isothermal temperature of some
1200°F with NaK. In addition to checking the system hydraulically at high
temperature, this circulation of hot NaK cleans the system of oxide. The
Na,O formed in this cleaning is trapped out of the system as itisformed.

Following the hot NaK test and after any leaks have been repaired the
NaK will be drained from bothliquid circuits. While maintaining tempera-
tures at 1200°F both circuits will be evacuated by pumping with vapor traps
ahead of the pumps. When all residual NaK has been boiled out of the sys-
tem, it will reach a steady state pressure, from which point rates of rise
of pressure can be used to detect the fluid tightness of the system. It is
hoped that this final test will leave the circuits tight and that no further
leaks will occur. After the fluorides are introduced, repair of the circuit
will be more difficult, even in some exireme instance impossible. '
120

Loading and Start-Up

 

In initial loading the fuel carrier fluoride melt is released from the
fill tanks into the evacuated fuel system. Checking for voids in the core
and piping will be done by observing any liquid level change in the surge
tank when the gas pressure over the liguid is raised. After the system
has circulateduntil it is proved tobe tight and all temperatures are stable,
the system is ready for critical loading. (The secondary circuit filling is
an identical procedure. )

Critical loading is effected by a fuel injection system which is filled
with highly concentrated UF; in the NaF, KF, ZrF; carrier. The system
is so arranged that samples of known weight can be introduced into the
surge tank at a controlled rate. The first load will be injected at such a
rate that the 'quantum'' of fuel, which is about six pounds, will require
three circuit transit times for complete injection. Since the fuel transit
time is about forty seconds, the 'quantum' injection will require about two
minutes. This will insure a uniform distribution of fuel in the circulating
system.

As the calculated critical loading is approached, the amount of fuel in
each "quantum”can be reduced from the maximum of six pounds to any
reasonable low value so as to minimize the chance of loading to a super
critical degree. The standard checksfor criticality inloading any reactor
will be employed in the ARE and thereactor will be brought tocritical with
the use of all possible safety precautions. The objective will be to go crit-
ical at a temperature of 1200°F with the shim rods as nearly completely
withdrawn as the controlled loading will permit.

As a safety precaution, attempts will be made to measure the temper-
ature coefficient of reactivity and to obtain a rough calibration of the regu-
larity rod as soon as possible. The first attempt to make these measure-
ments will be done whenthe reactor has reached a value of k = ~.98 or at
a source multiplication of about 50. The first experiment will be an attempt
to determine the sign of the temperature coefficient of reactivity. To do
this the fuel temperature will be lowered from the equilibrium value of
12009F to approximately 11750F when the source multiplication is approx-
imately 50. If the multiplication increases with this lowering of the tem-
perature, it will show that the temperature coefficient is negative, and fur-
ther loading can proceed. If, however, the multiplication should decrease,
it will show that the temperature coefficient is positive, and the experiment
should terminate at this point since a reactor which has a positive coeffi-
cient is considered to be too dangerous to operate. However, this latter
eventuality is not anticipated.
121

Following this experiment, an attempt will be made to obtain a rough
calibration of the regulating rod. Since the moderator is beryllium oxide
there will be anindeterminate number of gamma-n neutrons, and when the
fuel is circulating, some of the delayed neutrons are lost in the reacting
system. These two effects make the calibration of the regulating rod by
pile period measurement practically meaningless. However, a roughcali-
bration canbe obtained by noting the change in source multiplicationas the
regulating rod positionis changed. It is believed that afirst measurement
canbe made whenthe reactivity is approximately 0.98 with higher accuracy
results obtainable as k is increased toward 1.00. This will give a fair
calibration of the rod interms of Ak/k before actual criticality is reached.
After criticality is reached the rod will be calibrated by noting the change
in rod position for criticality with small additions of fuel.

After the rod calibration has been made to the best degree warranted
the reactor will be raised in temperature by the external heating facilities
to atemperature of some 13000F since this is the mean temperature level for
all future opetation. Atthisnew temperature fuel additiomwill be made to
again bring the reactor critical—but with the shim rods now inserted to a
distance of some four inches in the core. This will provide of the order
of 2% in k available' in shim rods to take care of poison as the reactor
operates sothat nofurther fuel addition should be necessary after the criti-
cal loading at 1300°F has been accomplished.

The next step inreactor operationcomprises raising the reactor power
from near zero ito some nominal value. This will be done by withdrawing
the regulating rod sufficiently to start the reactor on a positive period of
some twenty to thirty seconds. When this period has increased to near
an infinite period, as it will by virtue of the temperature coefficient, the
reactor will be at a power level of some ten kilowatts. This pointcan also
be checked roughly by the reading of the Log N meter. The reactor is now
ready for the power regime.

Power Operation

 

Operation at power will be attained as quickly as possible after the
reactor has been brought up to the nominal value of some ten kilowatts.
Up to this point there has been no power extraction from the system, but
when the reactor has stabilized at a nominal (ten) kilowatt level, the he-
lium blowers will be started. Heat extraction will be at a controlled rate,
so that the reactor inlet temperature will not be brought too low. Every
precaution will be used to insure no danger of freezing the fuel-coolant.
122

Since the object of the experiment is to obtain as much operation time
as possible at full power, an effort will be made to reach rated power as
quickly as possible. The mean fuel temperature of 1300°F will be main-
tained quite closely, but the fuel inlet temperature will drop and the reactor
outlet temperature will rise until at full power there will be a temperature
rise across the reactor of some 3500F.,

Should there be a false scram when operating at power, the helium will
automatically be cut off to insure against freezing the fuel. During the
power regime the system will be left very nearly alone. A minimum of
changes in power demand will be made and as nearly as possible the re-
actor will be operated in a steady state condition. Undoubtedly, interrup-
tions of one kind or another will occur, but there will be no intentional
"juggling' of any part of the system beyond the minimum required to estab-
lish a pattern of the transient response. The servo-regulating system will
not be employed unless it is found to be necessary because of an insuffi-
ciently large negative temperature coefficient.

Operation of the Servo-Control

 

During allof the runoutlined so far the reactor has been under manual
control. The servo system has been operating but it has been disengaged
from the regulating rodand hence has not been controlling the reactor. At
some time during the power operation regime, the servo will be connected
to the regulating rod so that the reactor will be on automatic control.
Under these conditions, since the servo is made to operate exactly as the
fuel temperature coefficient, the system response to power demands will
simulate the kinetic behavior of a reactor operating at high power output.

A suitable servo error signal is the following:

E =g, +a© +236, +a,
where a,, a5, a3 a,are proportionality constants,
flux
reactor inlet temperaiure
reactor outlet temperature

=
S,

#

In the steady state this error signal is independent of the external load.
For load changes both ©; and 6, change, and the regulating rod will main-
tain its original position. Furthermore, the power demand is made through
©; from the external load.

To raise the mean fuel temperature at constant load, a, is the driving
function and can be changed manually and slowly. Increasing a, causes
123

withdrawal of the regulating rod, an increase in ® followed by an increase
in @i, and later a decrease in @ until the error signal is again zero, after
the entire system, reactor and heat exchanger, have reached equilibrium.
Once equilibrium has been reached the regulating rod will no longer be at
its mid-positionbecause the reactor has a temperature coefficient independ-
ent of the servoerror function. Changing the shim rod position after equi-
librium has been reached merely moves the regulating rod to its mid-posi-
tion. This error signal characteristic is to be used for another purpose.
It will be noted that changing the shim rods will give a steady state zero
error signal. There isno way toraise orlower the meanfuel temperature
by moving the shim rod. Furthermore, any slow depletion or poisoning
would cause the regulating rod to move slowly out from its mid-position
at constatn power. An effect like this should be looked for since it may
suggest fuel precipitation in some part of the circuit.

It is proposed to use the recently developed pile simulator for deter-
mining the Ak/k for the regulating rod. It is likely that this simulator is
only reasonably accurate for a circulatingfuel ARE because it has not been
considered advisable to simulate transport lags in the system. However,
if the simulator is operated with no provision for the temperature coeffi-
cient and adequate control is maintained solely with the external servo
loop, then the actual control of the reactor with a negative temperature
coefficient should also be adequate. It will be possible to determine the
maximum Ak/k step for which the system is stable and permits no more
than 10% overshoot of meantemperature. Essentially this is done by sim-
ulating the MTR servo system so that the servo simulator controls the
pile simulator. By this means the Ak/k permissable in the regulating rod
can be determined and if it is desired a suitable factor may be employed.
With a safety factor in the permissible regulating rod Ak/k and inthe mini-
mum period for actuating the scram, the control system would be quanti-
tatively described, but not in terms of some arbitrary step function re-
sponse. The only arbitrary quantities would be the selection of the MTR
servo system and the length and weight of the regulating rod. It might be
said that the reactor will be controlled by the servo system to the limits
of the system's capabilities and that for more severe control demands the
scram system will take over. The safety factors insure that these two
systems overlap.

Shut Down

After the power run has been made, the reactor will be scrammed.
When temperatures have reached equilibrium and after the iodine has had
time to decay into xenon, an attempt will be made to go critical again. If
xenon or iodine is evolvedas a gas from the liquid fuel, the reactor should
124

be made to go critical at essentially the same conditions (as far as rods
are concerned) as obtained when it was shut down since the xenon poison
will not be present. If, however the reactor will not go critical with shim
rod adjustment, no fuelwill be added but additional time will be allowed to
elapse and another attempt to go critical will be made. If this latter at-
tempt is successful it may be assumed that the xenonhashad time to decay
and that the poison on the first attempt was xenon which had not evolved
from the fuel.

The fuel circuit will then be drained as much as possible but it is ex-
pected that some of the fuel mixture will remain in the bends of the fuel
tubes. Non-uranium bearing fluorides will then be introduced into the sys-
tem to flush out the fuel into the hot fuel dump tanks. This flushing will be
continued until the fuel is sufficiently extracted from the system. There-
maining fluoride mixture will then be drained (probably leaving some in
the core) and the system will be allowed to cool so that physical inspection
can be started as soon as possible.
125

ARE BUILDING

The ARE building is a mill type of structure designed tohouse the ARE
and the necessary facilities for its operation (Figure 35). The building has
a full basement 80 x 105 ft, a crane bay 42 x 105 ft, and a one-story serv-
ice wing 38 x 105 feet. The reactor and the necessary heat disposal sys-
tems are located in shielded pits in the part of the basement serviced by
the crane bay. Half of the main floor area is open to the reactor and heat
exchanger pits in the basement below; the other half houses the control
room, office space, shops, and change rooms.

Building Schedule

 

The Nicholson Company, contractor for the Test Facility Building,
was notified to proceed on July 6, 1951, and this contract work is now ap-
proximately 95% complete. The architect-engineer for this phase of the
work was the Austin Company. The AEC is now negotiating an addendum
to the initial contract for the remaining portion of the capitalized addition
and for some of the noncapital addition. The additional work includes the
concrete test pits, concrete heat exchanger pit, basement floor, crane
bay floor, additional partitions, process water supply lines, process water
disposal lines, settling pond, parking area, and paving. The engineering
for this phase of the work was performed by the ORNL Engineering Depart-
ment, and it is estimated that the Nicholson Company will complete the
work by June 1, 1952. At completion, drawings will be available so that
the Laboratory may begin installation of equipment, shielding, and addi-
tional process services.

Arrangement

The building consists of a full basement 80 x 105 feet. In half of the
basement are located the shielded reactor and heat exchanger pits; the
other half of the basement is servicearea. The control room, office space,
and some shops are located on the first floor over the service area. The
first floor does not extend over that half of the basement containing the pits,
and this area is serviced by the crane bay. The crane is a floor-operated
10-ton bridge crane having a maximum lift of 25 ft above the main floor
level. Plan and elevation drawings of the building are shown in Figures
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129

The basement includes the test and heat exchanger pits and space for
emergency power equipment and direct-current operational power equip-
ment. Crane access is provided to the heat exchanger and test pits by re-
movable roof slabs. The slabs cover the entire roof area of the test and
the heat exchanger pits except for supporting beams. Lifting hooks in each
slab enables it to be removed and stored in the slab storage pit, or if con-
taminated beyond a usable condition, it may be loaded onto a truck and dis-
posed of. The main floor plan shows the control room, general shop, in-
strument shop, conference and office space, and toilet and locker room.
The western portion or crane bay provides a truck loading area, a slab
storage pit, and the removable roof slabs of the heat exchanger and test
pits.

The building foundation, basement walls, and all floors are of rein-
forced concrete. The frame is structural steel with insulated flexboard
walls, and there is built-up roofing on precast gypsum slabs. The parti-
tions in the service wing are movable and made of insulated metal, and
the ones in the basement are of masonryblock. The test pits are provided
with process water supply and disposal lines to permit flooding for disas-
sembly work. The walls and floor are designed to be watertight. It is
planned to further waterproof this structure by application of Amercoat on
the interior. Flanged sleeves that may be covered after removal of piping
and before flooding are provided through the walls.

The shielding of the testand heat exchanger pits is provided by mono-
lithic concrete walls and solid masonry concrete blocks stacked dry on the
sides and removable concrete slabs on the top. The masonryblocks stacked
dry were used to permit greater flexibility for future experimentis. The
monolithic concrete walls were therefore held to a minimum thickness
consistent with accepted design. Removable blocks within the test pit are
shown in Figure 37. These blocks are designed to prevent contamination
of the monolithic concrete walls and will be disposed of as required.

Utilities

The building is to be heated with plant steam, and the pipe lines are
being constructed. Fin-type radiators and steam unit headers are used
throughout the building. Forced ventilation is furnished to the interior
rooms of the service wing. The crane bay is provided with gravity-type
ventilators mounted on the roof.

The electrical supply to the building consists of an extension of the
plant area 13.2-kv line. Both 440 and 110 to 208-v distribution are made
within the building. Direct current for all prime operating power will be
130

provided by ac.—dc. rotating equipment. A portionof direct-currentpower

will be converted to alternating-current power for instrumentation. Bat-
~ teries will be provided in a floating arrangement to insure continuous op-
eration.

A 100, 000-gal earth settling pond is provided approximately 1500 ft
south of the building and connected to the test pits by a buried steel pipe
line. The pond will permit disposal and decay of contaminated water used
as shielding in disassembly work. The pond may be emptied by means of
a syphon into a branch of White Oak Creek.

C ey
131

CONSTRUCTION SCHEDULE

Scheduling of the construction of the ARE is obviously afunction of the
manpower available. So far as can be determined, the activities of the
Laboratory that will be competing for available manpower have been taken
into consideration in the preparation of the timetable for the ARE. Every
effort has been made to be completely realistic in setting up the schedule.

At present all major layout plans are frozen and no basic changes are
contemplated. Core detail shop drawings are 99% complete. Heat ex-
changers, dump tanks, helium blowers, and valves are being fabricated
either by outside vendors or by local shop facilities. Control components
(actuators, housings, and absorber rods) are in process of manufacture.

Heaters for the fluid circuits are on order.

Instruments are about 90%

ordered, the remaining 10% are awaiting final detailing as the operational
procedure modifications dictate.

A schedule of the Experiment is tabulated below:

Comgletion Date:

1. Building accepted from contractor June 15, 1952
2. Detailed equipment fabrication drawings July 1, 1952
3. Approximately 60% of components available July 1, 1952
4. Pump for use of fuel circuit tested at design flow and temperature August 15, 1952
5. Cold critical experiment completéd September 15, 1952
6. Installation of electrical system, pipe fitting, and all reactor circuit

components completed November 15, 1952
7. Reactor plumbing, ducting, and interconnecting completed January 15, 1953
8. Control room wired and checked January 15, 1953
9. Reactor system hydraulically tested and tight March 1, 1953
10.Reactor dry-run test with inert fuel at temperature, completed

and reactor ready for final loading April 15, 1953
11.Critical experiment at temperature followed by run at power May 15, 1953
132

In order to carryout the manydetails of the ARE, the effortis divided
into four main categories, each of which is the definite responsibility of

one engineer. These divisions are:

Building construction
Equipment installation
Equipment design, procurement, and assembly

Operation

All operations are under way and although eventualities cannot be foreseen,
the ARE should meet the schedule of operation outlines above.
 

 

133

CONCLUSIONS

It is felt that sufficient basic work hasbeendone to establish the possi-
bility of building a circulating-fuel reactor for operation ata maximum fuel
temperature of approximately 1500°F for a significant number of hours.
Much more work on corrosion and fuel characteristics will be done and
significant improvement, such as in the case of corrosionresistance, is
anticipated during the construction period of the ARE.

The present reactor does not represent an optimum design. Rather,
it represents a compromise dictated by several concepts evolving from
procurement difficulties. No important principle has been sacrificed in
effecting the final design, but in the interest of getting much-needed data,
it was decided to get the ARE into operation as quickly as possible.

The first effect to be checked on the experimental reactor is the re-
activity change as a function of temperature. General experience in
the operation of a high-temperature circulating-fuel reactor will be gained.
Checks on nuclear calculations that must be used on aircraft reactors will
result from the operation of this experiment. Some additional corrosion
data will be obtained but should not be significantly different from that
which can be taken from less involved experiments. Informationinvolving
interaction of the various separate problems, which could not be achieved
except by the integrated operation of the entire system, will be obtained.
The ARE, in effect, established the first outpost in a new frontier of re-
actor design. When design considerations are as critical as those de-
manded for aircraft adaptation, they provide a sufficient, if rather nebu-
lous, criterion by which to evaluate the worth of the experiment.

A second ARE almost certainly should be built, and possibly a third.
Each reactor should be planned to supplement the previous one and to
bridge the gap as completely as possible between its predecessor and the
ultimate aircraft reactor.

Materials research should continue at the greatest possible level of
effort. Specifically, the fabrication of molybdenum should receive con-
siderable attention. Liquid hydrogeneous moderators, such as hydroxides,
134

have been studied experimentally. The results have shown promise but
are far from conclusive. Further work on such moderators should be
carried on at a high level of endeavor. A second ARE should incorporate
improved structure, moderator, and fuel materials so far as these de-

velopments warrant.
 

 

PART III

APPENDIXES
 

 
e ————

 

-

ReEort No.

ANP-65

ORNL-1154

ORNL-1170

ORNL-1227

ORNL-1294

Lex P-1

ANP-~52

NEPA-1830

ANP-53

ORNL-1287

Y-F15-10

BIBLIOGRAPHY

Title of Report

Aircraft Nuclear Propulsion Project Quarterly
Progress Report for the Period Ending June 10,
1951

Aircraft Nuclear Propulsion Project Quarterly
Progress Report for the Period Ending Sept,
10, 1951

Aircraft Nuclear Propuision Project Quarterly
Progress Report for the Period Ending Dec.
10, 1951

Aircraft Nuclear Propulsion Project Quarterly
Progress Report for the Period Ending March
10, 1952

Aircraft Nuclear Propulsion Project Quarterly
Progress Report for the Period Ending June
10, 1952

Nuclear Powered Flight - A Report to the
Atomic Energy Cominission by the
Lexington Project

Report of the Technical Advisory Board to
the Technical Committee of the Aircraft
Nuclear Propulsion Project

Final Status Report of the Fairchild NEPA
Project

Report of the Shielding Board for the Air-
craft Nuclear Propulsion Program

Preliminary Investigation of a Circulating-
Fuel Reactor System for Supersonic Aircraft
Propulsion

Several Circulating-Fuel Reactor - Heat
Exchanger - Shield Designs for Supersonic
Aircraft

Author(s)

R. C, Briant
C. B. Ellis
W, 3. Cottrell

R, C. Briant
C. B. Ellis
W. B. Cottrell

R. C, Briant
W.B. Cottrell

R. C. Briant
A.J. Miller
W. B, Cottrell

R. C. Briant

A.J]. Miller
W. B, Cotrell

R. W. Schroeder

A. P. Fraas
G. F. Wislicenus

137

Date Issued

Sept, 13, 1951

Dec. 14, 1951

Feb. 18, 1952

May 3, 1952

to be issued

Sept, 30, 1948

Aug, 4, 1950

no date, circa
late 1951
October 16, 1950

to be issued

to be issued
138

Y-F10-103

ANP-064,
Part 1

ANP-64,
Part II

_"_“-' B

Statics of the ARE Reactor, Summary Report C. B. Mills

 

The Technical Problems of Aircraft Reactors, C. B, Ellis
Part I

The Technical Problems of Aircraft Reactors, C, B. Ellis
Part 11

 

to be issued

June 1, 1951

Jan, 2, 1952