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Reactors-Research and Power
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CRITICAL EXPERIMENTS ON DIRECT
CYCLE AIRCRAFT REACTOR

Dixon Callihan
R. C. Keen

 
  
 

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OAK RIDGE NATIONAL LABORATORY
OPERATED BY

CARBIDE AND CARBON CHEMICALS COMPANNY
A DIVISION OF UNION CARBIDE AND CARBON CORPORATION
1] m

POST OFFICE BOX P
OAK RIDGE. TENNESSEE

   
Index No. ORNL-1615

This document contains 95 pages
This is copy 20 of 154 Series A

 

Subject Category: Reactors -
Research and Power.

CRITICAL EXPERIMENTS ON DIRECT CYCLE ATRCRAFT REACTOR

Work by: Dixon Callihan
E. V. Haake (Consolidated-Vultee)
R. C. Keen (Louisiana State University)
W. G. Kennedy (Pratt & Whitney Aircraft)
J. J. Lynn
Dunlap Scott
D, V. P. Williams

Preparation by: Dixon Callihan
R. C. Keen '

0CT 22 1953
DATE ISSUED

PHYSICS DIVISION

"A. H. Snell

Director
Contract No. W-7405, Eng 26 —_—
OAK RIDGE NATIONAL LABORATORY |f MARTINMARIETTA ENERGY SYSTEMS Lig

 

CARETE A RN CRRIONLS compATY IINHWIINI!HINI!H((H((NHlll(ffliIHMHINIHNI

A Division of Union Carbide and Carbon Corporatio
Post Office Box P
Oak Ridge, Tennessee

3 4456 0349579

  
 

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ORNL 1615
Reactors-Research and Power

 

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Powerplant La
Pratt and Whit

USAF Headquarters
U. 5. Naval Radiologii@lDefense Laboratory

University of Calif ediation Laboratory, Berkeley
University of Cal Wiation Laboratory, Livermore

 
 

ABSTRACT

The critical experiment program on the Direct Cycle Aircraft Propulsion
Reactor was planned Jolntly by General Electric Company and the ANP group
of Oak Ridge National Laboratory. The experiments were performed by the
Laboratory group at the Oak Ridge Critical Experiment Facility.

A cylindrical assembly having a core composed of alternate layers of
hydrogeneous material and of an open lattice of stainless steel and metallic
uraniuvm epproximately 36" long and 51" in diameter was constructed. The
assembly was reflected by a 6" thick annulus of beryllium around its circum-
ference and 6" of AGOT graphite on the ends which covered both the core and
the beryllium jacket. The reactor required a critical mass of 41 kilograms
of uranium enriched to 93.3% U-235, and used Plexiglas as & mock-up moderating
material. Two mock-ups of the Direct Cycle Reactor are reported here. In one
of these modifications the moderator, the fuel disks and the stainless steel
supporting layers for the fuel disks were horizontal. The second mock-up con-
sisted of a series of concentric rectangular shells formed by the rotation of
an appropriate quantity of the core materials from a horizontal to a vertical
position.

Core tjpe fuel and moderator removal control rods were calibrated by
both the stable period and the "rod drop" methods.

Power and neutron flux distribution studies were made on microscoplc
and macroscoplc bases for both these assembly mock-ups. Aluminum catcher
foils in contract with the U-235 metal were used in the power distribution
measurements and the flux dlstrlbutlons were determined by indium foil
activation.

Measurements were made of the changes in reactivity resulting from the
removal of a section of the beryllium reflector and the substitution of other
materials for it. Losses in reactivity were also measured in terms of the
separation of the two halves of the assembly at midplane and compared with
calculated values, treating the assembly as a bare thermal reactor. Reason-
able agreement was obtained between experimental and computed values.

Evaluations were made of certain experimental poison rods and. of sub-
stituting limited amounts of molybdenum for stainless steel and of graphite
for Plexiglas.

The end reflector was removed from approximately one eighth of the
aggembly and a stack of Plexiglas erected and other alterations made to
further mock-up the Direct Cycle Reactor. The effects of these changes
on the reactivity and neutron flux were determined.

 
|

TABLE OF CONTENTS

Page
AE’IRACT » . e . * ¢ . e [ L] . > * | - * . & . - ’ o - . * . e 5
IlIST OF FIGURES [ * e o ¢ & * o * - o ) ® * e . [ ) * . * - L] * ' 7

IIST OF TABIES . . ® . * . * * - - o o . * - * ® * a . * * ® . lo
T. INTRODUCTION o « o « o o o o o « o o o o o o o o o o s wao. 11
II. IESCRIPTION OF EQUIPMENT AND ASSEMBLIES . . . . + . . . . .. 11

A. Permanent Equipment .. . . ¢« & ¢ ¢ 4 ¢ o o o o o o o » o o 11
B. Shelf: Type Mock-Up:of Direct Cycle Reactor . . .« . « « . . 1h
C. Conversion from Shelf Type Assembly to Rectangular

SHOLL TFPO = o v o o o o o o o o o o o o o o o e v o o4 18
D. Conversion from.Rectangular Shell to Water-Cone

Reflocteod MOCK-UD +« « « o o o o« ¢ o o o« o o o o « o » o 18

III. EXPERIMENTAL STUDIES IN SHELF TYPE ASSEMBLY . . . « « . . .« & 22

A, Control Rod Calibrations « « « o o o« o o o« o o o o o o « o 22
B. Temperature Effect8 . .« « « ¢ o ¢ ¢ o o 0 o ¢ o o ¢ o o o« 27
C. Reflector Studles e e s o s o 4 e o s o s o s e e e e+ 29
D. Power and Neutron Flux Determinations . . « « « ¢« « » » « 36

IV. EXPERIMENTAL STUDIES Ofi RECTANGULAR SHELL TYPE ASSEMBLY . . . 56

A, Control Rod Calibratlions . « o o o o o « o ¢ o s o ¢« o« « « 56
B, Temperature Effects . . « ¢ ¢ ¢ ¢ ¢ ¢ ¢ 0 ¢ o o o @
C. Plexiglas Thicknees Reactivity Coefficlent . . . . «. « . & 59
D. Effect of Separation of Core at Midplane . . . . . .

E. Rofloctor StUdle8 o « + o o o o o o o o o o o o o o o o o 63
F. Danger Coefficient Typs EvAluatlons . « « o+ « o o o « o+ o 66
G. Power and Neutron Flux Determination . . « ¢« « ¢ ¢« ¢ « & T2

V. EXPERIMENTAL STUDIES ON WATER-CONE REFLECTED ASSEMELY . . . . 86
A. Travefsés ianater-Cone Reflected Assembly . « « « . . . . 86
VI.comciusxoms......_,,.,............-.....9o
VII. APPENDICES + & « « + v v o o o o oo v v o n oo v ueoon Ol
A, Conversion from Shelf to Rectangular Shell Mock-Up . . . . 9l

B. Analyses of Reactor Materials . . . . « ¢ ¢« ¢ ¢« o o o« « « = 93
C. Summary of Materials in Reactor Assemblies . . . . . . . . 9

mnliiy

6.

 
11.

12,

13.

1k,

15.

16.
17.
18.
19.
20.
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22,

23.

-

LIST OF FIGURES

 

Photograph of Two Halves of Aluminum Matrix . o « + « . .

Photograph of Fuel Element o o o o o o « o . .

Loading Chart of Shelf Type Assembly . . . .

Photograph of Interface of Shelf Type Assembly . . . . . .

Loading Diagram for Rectangular Shell Type Assembly . . .

Photografh of Midplane of Rectangular Shell Type Assembly

Photograph of Small Area of Midplane of Recténgular Shell

Type Assembly . * . . . . . o o . Q * * * .

CalibratimnAof'Control.Rod A v v v e 6 o e s s e 0 e e

Sengitivity .of Control RodiA e e e 4 e 4 e s

Reactivity Change ve Temperature for Shelf Type Assembly

Refleétor Value vs Void Width for Composite Reflector . .

Sensitivity of Reflector Value vs Vold Width for Composite

Reflector * . e L] * * . * * . » . . . . * .

Reflector Value vs Composition for Composite Reflector of
Stainless Steel and Plexiglas located at Bottom of Reactor

Reflector Value vs Composition Reflector of Stainless Steel
and Plexiglass located at Side of Reactor . . . « o « « « &

Photograph of Materials used in Power and Neutron Flux

Determinations . ¢« ¢ ¢ ¢ ¢ ¢ o o ¢ & o 4 e

Radial Power Distribution in Shelf Type Assembly . . » . . .

Axial Power Distribution in Shelf ije Agsembly . . . .

Power Distribution 10 Cell Mel2 & v v v v o o o o o o o o &

Radial Neutron Flux Distributlon In Shelf Type Assembly . .

Axial Neutron Flux Distribution in Shelf Type Assembly . .

Vertical Radial Neutfon Flux Distribution in Shelf Type Assembly

Outline Drawing of a Unit Cell . . . o ¢« « + &

Vertical Neutron Flux Distribution in Unit Cell Through Center

Of Fue l Di Sk * -* - o e ° . o . * e * - o . o

_7-

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20

21
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L8

L9
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3L,

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h’Oo

L1,

el

Vertical Neutron Flux Distribution in Unit Cell off Center .
Of Fuel Disk * . @ ® » 0 o ® e . o ® * 0 ® a * o . L] *® ® - * *

Vertical Neutron Flux Distribution in Unit Cell near Edge of
Fuel DiSk . * . o e * e o e - o ° ® . . » » ® . ’ * o - *

Vertical Neutron Flux Distribution Through Axis of Unit Cell
Between Fuel Disks . & . & 4 v ¢ ¢ 2 « o o o o o o o o o »

Vertical Neutron Flux Distribution off Axis of Unit Cell
Betwoen Fuel Diske o o v o ¢ o ¢ o o o o o o o o o o o o

Vertical Neutron Flux Distribution Near Edge of Unit Cell
Between Fuel DiBKB o o o o « o o 6 o o a o « o o o o o o

Neutron Flux Distribution Through a Unit Cell M-12 Having
2-Mi1l Disks Parallel to. PlexXiglas . « o o v o o o o o o o o

Neutron Fiux Distribution Through a Unit Cell M-12. Having
2-Mil Disks Perpendicular to Plexiglas . . . . . . . . . .

Bare and Cadmium Covered Indium Traverses Across a 10-Mil
Fuel DiSk 2 L] o ? L] e N * o ° . * L] ® * . e L] L] o * . 2 - *

Change in Reactivity ve Plexiglas Thickness . . + « « « + . .
Change in Reactivity vs Separation Distance at Midplane . . .

Change 1in Reactivity ve Plexiglas Thickness for Stainless
Steel-~-Plexiglas-Boral Composite Reflector . . . . . e e

Change in Reactivity vs Plexiglas Thickness for Stainless Steel-

Plexiglas-Cadmium Composite Reflector . « o s o o o o ¢ o @

Bare and Cadmium Covered Indium Traverses through Composite
Reflector with Boral Strips Present . o ¢ ¢« ¢ ¢ ¢« o ¢ o « &

Cadmium Fraction vé Distance from Bottom of Test Section for

Composite Reflector with Boral Strips Present . . . . . . .

Bare and Cadmium Covered Indium Traverses Through Composite
Reflector with the Boral Strips Removed . « « ¢« ¢ ¢ o « o &

Cadmium Fraction vg Distance from Bottom of Test Section for
Composite Reflector with Boral Strips Removed « « &« ¢« + .

Power Distribution Through Cells M-12, M-8, M.6 and M-4 in
Rectangular Shell Type Agsembly . o o « o+ o ¢ o o o o o o &

Radial Power Distribution at Midplane in Rectangular Shell

Type Assembly * o o o o * . . . o e o o ° * . L] . o . & * e *

-8-

Page

50
51
52
o3
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22
o7

58
60
62

6l
&5

67

69
70
73

7>
42,
43.

L,

L5,

L6,

47,

49.

50.

5_10

52,

 

el

Shelf-Shield.ing Of Fuel DiSk o’ * o - * 4 L} * * . . * . e * . . *

Vertical Radilal Neufiron flux Distribution between Plexiglas
in Rectangular Shell Type Assembly v « o « « « ¢ o & « » o &

Horizontal Radial Flux Distribution Through Plexiglas in
Rectangular Shell Type Assembly . o o ¢ ¢ o ¢ o o ¢ o ¢ o &

Axial Flux Distributibn‘Through Cell M-12 in Rectangilar Shell
TypeAssembly....A..._..._......-........

Horizontal Neutron Flux Distribution at Plexiglas Intersection
in Rectangular Shell Type Assembly . . . « « + v ¢ + o o &

Vertical Neutron Flux Distribution Across Centrial Rectangular
She ll Type Cell’ M-12 -+ . ” . * * . ® ° . o * » . * * * - *

Horizontai Neutron Flux Distribution Around General Electric
Company Experimental Polson Rod + + & & o ¢ « ¢ & o o o & &

Vertical Neutron Flux Distribution Around General Electric
Company Experimental Polson Rod . « &+ « « ¢ o & v o & & © &

Axial Neutron Flux Distribution Through Cell R-12 In Water-Cone
Reflected. MOCk-UP . ¢ . * . o .. * . * . . LI . * ‘u e o . .

Axial Neutron Flux Distribution Through Cell U-12 1n Water-Cone
Reflected MOCk-UP * * . * * * * o ) * [ ] o e 4 * * - . * . . .

Axial Neutron Flux Distribution Through Cell H-12 in Unaltered
Segment of Water-Cone Reflected MoCk-Up . « « o« & ¢ & & « &

 

Page
_77
78
79
81
82
83
8
8
87
88

89
1I,

IIT.

XIV.

Xv,

 

Page
LIST OF TABLES
Comparison of Rod Calibratlons « « « « « v ¢ o ¢ o o o« o o o 26
Reactivity Change Introduced by Substituting Beryllium
for Air * & * 9 e * * . * * ® ., * * 4 * * * .. o * * * - ’ 29
Reactivity Change Effected by Substlitution of Stainless
Steel and Plexiglas for Beryllium . . « o o « o ¢« o & o & 30
Reactivity Losses for Varlable Void Width in Reflector . . 31
Reactivity Losses Effected by Substitution of Stainless Steel
and Plexiglas for Beryllium . ¢ . ® & * . * * L ] ¢ * [} . * ) 31
Reactivity' Losses Effected by Varying Composition of the
Bottom Reflector ® . . ) * [} . .. e v o . L} s e * - o e . L ] o ' 3'+
Reactivity Losses Lffected by Varying Compositlon of the
Side Reflector . * * . e . .. & e & e * . & . * e L] * & . * . 36
Thermal Neutron Distribution in a Unit Cell . . « « o . . + . 56
Calibration Ihta"for Control Rod AT 4n Rectangular Shell
Tneflssemblyoooo.olqo‘ooo--oonooooo.ooo 59
Summary of Control Rod Cai_l.ibz_'a'tion DRt o ¢ ¢ o ¢ o ¢ 0 0 o 59
Effect of Core G’ap on ReactiVity ¢ ¢ o & o 0 e ¢ e . o e 61
Polson Rod Reactivity Changes « + « « o ¢ ¢ s ¢ ¢ o ¢ ¢ « o T1
Compar‘ison of Molybdenum and Stainless Steel . ¢« « ¢ ¢« ¢+ o & 11
Relative AXial POWBI‘ s o @ .",.)...l... s 9o & @& o % o & & » & @ 72

Cadmium Fractions of Fuel Disks . « « o « + o s o o ¢ o o o 76

.

«l0a
I. INTRODUCTION

This report is a summary of the Information contained in five progress
reportsl, of limited circulation, published at intervals during the course
of these experiments. Since these progress reports were of limited cir-
culation and avallable only to those directly connected with the project, it
seemed advisable to make the information obtained from the critical experiments
available on a wider basis in the form of an overall report.

A description of the program of the critical experiments to be performed
by the Oak Ridge National Laboratory contributing to the design of the Direct
Cycle Nuclear Reactor by the General Electric Company is given in their Doc-
ument DC-51-11-6, dated November 3, 1951, entitled "9213 Critical Experiment
Program", and supplements, dated December 14, 1951; June 12, 1952 and others
subsequently lssued as work on the program progressed.

The first mock-up of the reactor as actually constructed consisted of a
right circular cylinder, with axis horizontal, and a core 36" long and 51"
nominal diameter containing 41.2 kilograms of uranium or 38.5 kilograms of U-235,
The H to U-235 atomic ratio was 225. The core composition by volume fractions
was as follows: Plexiglas (to simulate water) 0.319; stainless steel 0.043;
uranium 0.00185; aluminum 0.063 and voids 0.573. It was assembled in a matrix
of square aluminum tubing which will be described later. This mock-up, in
which the Plexiglas was in horizontal layers separating groups of six layers
of an open structure of type 302 stainless steel sheets holding the uranium
metal disks, was designated asg the "shelf type assembly”. This mock-up is
also referred to as CA-5. |

The second mock-up consisted of a series of concentric rectangular shells
formed by rotating an appropriate number of Plexiglas strips and their accompany-
Ing structure of stainless steel sheets and fuel disks from a horizontal to a
vertical position. This rotation was supplemented by minor changes in some core
and reflector materials but with no change in the amount of uranium present.

This modification, which will also be described in detail later, was known as the
"rectangular shell type assembly" and was used to obtain additional data pertin.
ent to concentric annuli in the design of the Direct Cycle Reactor. Thls mock-
up is referred to as CA-6. Additional minor changes were made to the rectangular
shell type assembly during the course of the experiments and this third mock-up
1s designated as CA-T.

" II. DESCRIPTION OF EQUIPMENT AND ASSEMBLIES

A. Permanent Equipment:

 

The critical experiment assembly has been adequately described in

other reportse, and only a brief description of the components which

are essential to understanding of current experimental. techniques

will be included here. Such features as shielding, interlock system,
safety systems and scrams. are omitted since they have very little bearing
on these experimental results.

 

1 Callihan, D. '"Preliminary Direct Cycle Reactor Assembly Part I, Y-B23-1,
February 26, 1952; Part II, Y-B23-2, May 21, 1952; Part III, Y-B23-5, June 18,
1952; Part IV, Y-B23-7, June 30, 1952, and Part V, ORNL 52-12-225, December 15,

1952.
2 Bly, F.T., et al, NEPA Critical Experiment Facility, NEPA-1769, April 15, 1951.

-Xl-
Assembly Structure

 

' The assembly apparatus consists basically of a matrix of square

2S aluminum tubes stacked horizontally which, when assembled,
form a &' cube and into which the reactor materials may be placed.
The &' cube 1s divided into. two identical halves, except that one
i1s stationary and the other can be moved by remote control a dis-
tance of 5' from the stationary half. Xach half consists of 576
tubes, stacked in a 2L x 24 cell array. These tubes are 36" in
length, 3" x 3" outside, and 0.O4T" wall thickness. Part of the
reactor materials are placed in each half and the assembly 1s made
critical by control rod adjustment after the two halves are to-
gether. Figure'l 1s a photogiaph of the two halves of the assembly.
(The materials visible in the upper portion of the moveable half
have no relation to the Direct Cycle Reactor.) The moveable half
is driven by a power screw coupled by gears to a 1 HP. D.C. shunt
wound motor. The speed of the motor and the reducing gears 1s so
ad justed that the moveable half approaches the fixed half wilth a
speed of 33" per minute between 5' and 8" separation, 6" per
minute between 8" and 2", and 1" per minute between 2" and the
stopping position of the moveable half. Positions of the moveable
half relative to the fixed half are indicated to 0.01" over the
entire 5' travel by a Selysn Veeder-Root mechanism coupled to the
power screw. Separations less than one half Iinch are Indlcated

to 0.001" by Starrett Micrometer dial indicators coupled through
Autosyn 400 cycle Selsyns to points at the bottom of the matrix.
The precision of the incremental separatlions of the two halves
was the order of¥ 0.001". Due to a lack of parallelism of these
facing halves, however, the lnaccuracy of the overall separation
was estimated to be about 0.02",

Instrumentation

 

The radiation level of the reactor 1s monitored by eight in-
dependent instruments: a scintillation counter, a fission
chamber, and six BF3 filled neutron detectors, two of which are
proportional counters and four are ionization chambers. Leakage
radiation is measured by placing, the detectors at varying dis-
tances from the reactor core. @f special interest in the present
experimental procedures are the logarithmic meter and the plle
period meter, both of which derive their incoming signals from a
BF, ionization chamber, and are used in the calibration of the
cofitrol rods.  The logarithmic meter records power levels over a
range of approximately 6-1/2 decades, and the period meter records
instantaneous pile periods from infinity to plus 3 seconds and from
infinity to minus 30 seconds. Stable pile periods may be observed
directly from the slopes of the curves traced by the logarithmic

recorder.

-12- .
13

 

Fig. 1. Photograph of Two Halves of Aluminum Matrix.

 

 
Bfl

The Shelf Type Mock-Up of the Direct Cycle Reactor:

1.

Core Construction

 

The enriched uranium metal 1s in the form of circular disks
approximately 0.01" thick with diameters of 2.860" and 1.430"
and masses of 18,0 F 0.1 and 4.5 * 0,1 grams respectively. Since
it was not possible to roll uranium metal foll of sufficlent
uniformity for this emall mass tolerance, meny of the disks

had small holes punched in them for mass adjustment. Each

disk had a 0.196" hole drilled in the center. The fuel ele-
ments were assembled by placing the large disks on 3.6" centers
above an open array of three sheeth of stainless steel and cover-
ing them with an inverted three plece open array. Because of
insufficient large disks, 1t was necessary, In some cases, to
substitute small ones. Quarter size fuel elements were also
assembled and used near the circumference to approximate a cir-
cular core from square elements. Photographs of the fuel elements
are shown In Fig. 2.

The moderating material, Plexiglas, which has been shown to
simulate water in thermal neutron systems3, was in horizontal
layers, one inch thick. The separate layers support the open
structure of stainless steel and fuel disks as shown in Fig. &4,

Some detalls of the core construction are shown in the loading
chart, Fig. 3. The numbers and letters in the cells show the
positions of the safety and control rods respectively. Safety
rods ' 1, 2, '3, and 4 and control rods A and B are in one half of
the assembly and an equal number are in the other. The assembly
was critical with some of the control rods rémoved. Correction
for the accompanying voids, based on an empirical calibration of
the rods -in units of mass of uranium indicates the "clean™ critical
mags to be Ll.1 kilograms of uranium or 38.4 kilograms of U-235.
The structure of the control and safety rods 1s, of course, the
pame as that of the fuel elements.*

Reflector Construction

The core“of the reactor is enclosed in an end reflector of 6"

' of AGOT graphite and a nominal 6" thick side reflector of beryllium.

The beryllium thickness varies slightly because the annular reflector
is fabricated from square pleces. The graphite end reflector extends
to the outside surface of the beryllium annulus. The control and
safety rods contain the graphite reflector. A photograph of the
assembled core at the midplane is shown in Fig. L. :

 

P.

Callihan, D., et al, "Critical Mass Studles, Part VI", Y-801, August 8, 1951
12, .

A tabulation of the materials built into the assemblies. is given in -
Appendix C pg. 95.

-14.
 

 

Gl

Fig. 2. Photograph of Fuel Element.

 
16

owG. 21337 N

A B CDEFGHI JKLMNOPQRSTUVWIX

 

Be REFLECTOR, JACKET 18" LONG, EACH HALF BACKED BY 6"
GRAPHITE (END).

 

FUEL ELEMENT, EACH 18" LONG, EACH HALF BACKED BY 6"
GRAPHITE. (TOP)—6 LAYERS STAINLESS WITH 5-10 mil x
3" U DISCS, HORIZONTAL; (BOTTOM)— I' PLEXIGLAS.
TAINLESS STEEL, AIR, FUEL

LEXIGLAS

FIGURE 3. LOADING CHART
 

17

Y-12 Photo 11077

 

Fig. 4. Photograph of Shelf Type Assembly.

 
Conversion from Shelf Type Assembly to the Rectangular Shell (RS) Type.

The converslon from the shelf type criticel assembly, shown in Fig, 3
to the rectangular shell (RS) type shown in Fig. 5, consisted essen-
tilally of the rotation of the planes of the fuel dlsks, the stainless
steel sheets and the moderator strips from a horlzontal to a vertical
position In approximately one half of the core cells. The effect on
reactivity of the rotation of these core materlals has been calculated
and is contained in Appendix A, Minor changes in the amounts of core
materials.were made by substituting stainless steel sheets for Plexi-
glas in the central cell, M-12, in which control rod A was located.

It is recalled that the control:.and safety rods, all of which contaln
fuel elements, extend from the end of the reactor to the midplane of
the core. The only change in the reflector was the addition of a 2"
layer of graphite reflector to one end, making it 8" thick, the other
remaining 6". The lengths of control rods A and B were thereby in-
creased 2". '

No changes were made in the locations of the control rods and the

.safety rods nor in the amount of uranium in the core of the reactor.
‘Figures 6 and 7 are two photographs of the RS type assembly taken

at the :midplane perpendicular to the axis of the reactor. Figure 6
shows the overall loading and Fig. 7 shows, somewhat more in detail,

the arrangements of stainless steel and Plexiglas within the cells.
The vacant cells shown in the photographs indicate the locatlons of

either control or safety rods. The irregular boundary between those
cells which have the planes of the fuel elements vertical and those

in which they are horizontal is also visible.

If the rotation of the fuel, stainless steel and moderator strips

is in such a direction as to place the fuel adjacent to the beryllium
reflector the assembly is more reactive than when the rotation is in
the opposite direction since the latter puts a thick layer of hydro-
geneous material between the core and the reflector. It was estimated
that the RS type loading, as shown in Fig. 5, was less reactive than
the former shelf type by about one percent in A k/Xeff.

Converslon . from the Rectangular Shell. to the Water-Cone Reflected
Mock-up.

The water-cone reflected mock-up was constructed to better simulate
water cooling channels of the reactor, ascertaining their effect on
the flux distribution and reactivity. This information, céupled

with the reflector studles, determined the severity of the penalty

' apsociated with the contemplated. substitution of the iron-water

reflector for beryllium.

Reflector changes made in the (RS) type mock-up of the Direct Cycle
Reactor to simulate the water-cone reflector proposed for the end of
the reactor were the following: .The six inch graphite end reflector
was removed from one quadrant of dne half of the assembly consisting
of the following cells: N-12; 0-11, 12; P-10 through P-1%, inclusive;
Q-9 through Q-15, inclusive; R-8 through R-16, inclusive; S-7 through
S-17, inclusive; T-8 through T-16, inclusive; U 10 through U-1k4, in-
clusive. The Plexiglas in this group of 49 ¢ells was extended 10 /4",
These alterations in the reflector made the reactor subcritical even

~18.-
O D g O O H NN -

9

A BCDEFGHI JKLMNOPQRSTUVW X

 

Be JACKET REFLECTOR, 36" LONG

FUEL ELEMENT, 36" LONG
STAINLESS STEEL, URANIUM, AIR SPACE, l%" THICK

PLASTIC, I" THICK

FIGURE 5. LOADING DIAGRAM
Y-12 Photo 11452

 

Fig. 6. Photograph of Rectangular Shell Type Assembly.

20
 

Assembly .
lar Shell Type

Il Area at Midplane of Rectangula

r

i 'ograph of Sma

Fig. 7. Photogra

12

 

 
with all the control rods in. Thie subcriticality was compensated
for by the addition of 36" lengths of beryllium to the following
cells: B-8 through B-16, inclusive; C-6, 7, 17, 18: D-5, 19; B-4,
20; F-3, 21; G-2, 22; H-1, 2, 22; I-1, 22; J-1, 23; K-1, 23; L-1,

23. The 6" graphite end reflector was removed from the portion of
the beryllium Jacket which was adjacent to the quadrant contalning
the 10 1/L4" Plexiglas extension and replaced by a 10 1/4" beryllium
extension making the reflector the same length as the altered. core.
The final change conslsted of erecting a stack of Plexiglas, approx-
imately 12" thick, adjacent to one end of the reactor and starting
at the table top and terminating at top of the following cells: Q-7,
R-6, S-5, T-5, U-5, V-5 and W-6. The vertical edge of this Plexiglas
stack, which simulated the water-cone, was 3/4" from the end of the
aluminum matrix. The width of the simulated water-cone began at the
first safety and control rod supporting column and extended to within
about 1/2" of the edge of aluminum matrix.

The change in reactivity produced by the removal of the graphlte and
the extension of the Plexiglas in the 49 cells listed in the loading
change was a decrease of 85.2 cents. The addition of the 36" extension
to the beryllium reflector produced a gain in reactivity of 106.3 cents.
Gains in reactivity were also observed due to the 10 1/4" extension of
the beryllium reflector and to the erection of the simulated water-cone
in the amounts of 17 cents and 8.4 cents, respectively.

I1T. EXPERIMENTAL STUDIES ON SHELF TYPE ASSEMBLY

A.

Control Rod Calibrations:

The displacement of a calibrated control rod serves as a measure of

the changes in reactivity of a critical system. In the calibration

the changes In reactivity introduced by the linear displacement of a
control rod may be determined by either of two methods: 1) the measure-
ment of the stable period resulting from the reactivity cliange, and

2) the "rod-drop" method'.

1. The Period of a Super Critical System

A measurement is madé of the period of the super critical system

resulting from the insertion of a control rod. When the changs
in reactivity is small, it 1s related to the super critical
period by the following expression:

Q:E“____ - _L£24%§i_.= —__4éé___. (1)
kKeff. )L /")4‘7_

{
Where &k = change in multiplication or excess reactivity

keff.

-22-
|

T = stable plle periocd in seconds :

_n = number of groups of delajed neutrons

')3_= decay .constant for neutron precursors of group 1
= fraction of fission neutrons in delayed group 1

or in units of cents:

7 / \
| 7. | =100 ST 77 (2)
Korf E ]54 A\ TN Z,ga
| ' /
/ . .
where 100¢ is the reactivity equivalent to the effective
fraction of delayed neutron. The symbols in equation (2)
have the same significance as in eguation (1). Reactivity
results are reported in cents values to avold the uncertainty

in the effective value of the delayed neutron fraction nec-
essary ‘in converting .to dk/kerr-"

 

/2= 100

"Rod-drop" Method -

 

Tn this msthod a measurement is made of the translent occuring
when the safety rod is quickly removed from the assembly. The
change in reactivity in cents (/) may be computed from the
equation

N. _ N -
/o: 100 .._9_.__._1__ ' (3)
Nl ’
where No = power level at critical |
Nl_s extrapolated power level of the slow transient

at the instant the rod is removed

The change in reactivity produced by the linear displacement

of each control rod was determined by these two methods. The
supercritical period was measured for various increments of
displacement until the rod had been displaced its full length

and the corresponding changes in reactivity were computed from
equation (2). Each of the safety rods,which was ¢colinear with'a
control rod,was evaluated from the transient occuring in the flux
as it was dropped. Evaluations of the safety rods were made from
equation (3) and compared, respectively, with the integrated value
of the symmetrically located control rod. Table I glves a summary
of the total rod evaluations, in cents, for the four control rods
and the values obtained by the rod drop method for the correspond-
ing safety rods. Typical data obtalned in the incremental call-
bration by the stable period method are shown in Figs. 8 and 9.

The change in reactivity occuring when the rod is withdrawn from
the reactor, plotted as a functlon of the rod position from the
midplane, 1s shown in Fig. 8. In Fig. 9 1s plotted the change in
reactivity per unit displacement, the rod gensitivity, as a function
of the same rod position. The horizontal lines represent the aver-
age sensltivity over the incremental displacement. The peaks

. _23_‘
£

ROD VALVE

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Pl
/
s
7
//
7
P
v
P
L’
/| //
//
/)
0 12 {6 20 249
¢ ROD POSITION FROM MID PLANE 1IN INCHES

FIGURE 8.

CONTROL ROD A

6E€E1C "OMa

be
SENSITIVITY @/IN

o
o

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

P o

10 1S
ROD POSITION FROM MIDPLANE —IN INCHES

FIGURE 9. CONTROL ROD A

20

25

— ovEld "OMd

G¢
In Fig. 9 are attributed to the structure of the rod, but

this result was not investigated. It should be noted however,
that these peaks occur at the approximate positions where the

channel at end of the reactor is opened first by the reflector
removal and then by the removal of the complete element.

TABIE T

COMPARISON OF ROD CALIBRATIONS

 

 

 

Control Rods Safety Rodse

(by integrated periods) (by rod drop)
A 18.2¢ 5 164
B 15.9 6 16
C 15.9 1 18
D 18.3 3 18

It is believed that the precision of the rod drop method 1is
no greater than “$2 cents so that the agreement in some of the
above cases 1s perhaps coincindental.

In the course of the rod calibration experiments by the in-
tegrated period method it was noted that the positions of the
control-rods, when critical at an increased power level 1n a
reactor which has been operated for several hours reproduced,
within the limits of experimental error, the previous critical
positions at a lower power, whereas the reverse was not true.
That is, when the power was decreased by a factor of ten 1t was
found necessary, in order to re-establish a critical system, to
remove the control rod a few tenths of an inch beyond the po-
sition where the system had been critical at the higher power
level. (

This variation in rod position for criticality 1s attributed

to an effective source of neutrons arising from the Be (¥,n)
reaction which is relatively stronger at lower power levels
following operation at higher power. In order to minimize

errors arising from this source the following procedure was

used in the calibration of, say, .control rod A. With the neutron
flux constant at the lower level sufficlent excess reactivity was
introduced by inserting some rod, not A, to produce a stable period
of between 250 and 400 seconds. The power was allowed to increase,
on this period, approximately ten fold and the assembly was made
ceritical by withdrawing rod A. This displacement of A corresponded
to the observed period. The power was then reduced and leveled by
the appropriate manipulation of other rods and the procedure re-
peated, thereby evaluating successive Increments of rod A. It is
believed that, except for the first measurement, the rod displace-
ment which produced the positive period is not exactly related to
the period because, at the lower level, the assembly is not critical

6.
due to the (¥, n) neutrons. In the Direct Cycle Reactor there
was no beryllium in the core and this effect may not have been
very large.

An investigation was made of the linearity of response of the
amplifier ion chamber circuit used In the perlod measurement
by comparing it, at several lower levels, with proportional
counters. A small deviation from linearity was found at power
levels above those used in the control rod calibrations so the
results are not significantly in error™.

B. Temperature Effects

 

At various times during the operation of reactor assemblies varl-
atlons have been observed in the day-to-day settings of control

rods required for criticality under constant loading conditions.

In ‘the work reported here these varlations exceeded the sensitlvity
required to detect the reactivity differences produced by some
structural changes under study. Investigatlons have indicated the
probable cause to be ambient temperature irregularities. The sta-
bility of the room temperature has been improved so that the exlt

alr temperature is constant to one half Fahrenheit degree at 80CF.

An empirical relationshlp between reactor temperature and reactivity
has been established. The temperature, measured at two points in

the matrix with an iron-constantan thermocouple, has been varied
from 63°F to 80OF by altering the room temperature and allowing
sufficient time for the temperature throughout the reactor to equal-
ize. The attending reactivity differences, from<the reactivity at
72.9°F, were measured in terms of previously calibrated control rods
and are plotted as a function of temperature in Fig. 10. It 1s noted
from this graph that the reactivity increases with Increaslng temper-
ature, i.e. fuel control rods must be removed from the reactor to
maintain criticality if the temperature increases. It is realized
that, if referénce is made to a reactor: temperature coefficlent, the.
sign of the coefficient is opposite to that expected for hydrogeneous
systems. The effect is possibly due to thermo-mechanical changes in
the assembly which are complex because of structure, loading and oper-
ation. It might also be noted that an experiment in which the reac-
tivity as a function of the Plexiglas thickness was measured, showed a
maximum at a thickness slightly less than one inch. The above tem-
perature effect may be accounted for, at least in part, by conslder-
ing the concomltant change in density as effectively altering the
thickness of the Plexiglas. No systematlc investlgation of this
phenomenon has been made. If, during experimental operations, tem-
perature variations became gsevere the empirical relation of Fig. 10
allowed the interpolation of all reactivity data to some arbltrary
value. It has also been shown that, to at least the first order, the
magnitude of the temperature effeot, 1.25 ¢/Fo, 1is 1ndependent ‘of the
experimental ‘structiural changes described below.

 

4 Zimmermen, E.L., A Graphlte Moderated Critical Assembly -- CA-L, 1-881, "
Dec. T, 1952, Appendix E, page T1l.

-27-
+10

CHANGE IN REACTIVITY ¢
O

 

 

 

 

 

 

 

/ “
o
o
L
i3°F 72.19°F EI?°F
9 1.O LI .2 .3 1.4

 

 

 

 

 

 

 

FIGURE 10. CHANGE

MILLIVOLTS (IRON-CONSTANTAN)

IN REACTIVITY VS. TEMPERATURE
REFERRED TO 72.9°F

Iveld "OMA

8¢
C. Reflector Studies

 

Some comparisons have been made of the reactivity effectiveness of
the beryllium reflector and of one of a Plexliglas-stainless steel
composite,

1, Reflector Comparisons With Constant Void Thickness

A very approximate value of the beryllium reflector was obtained
by removing beryllium sections, 22" long from cells M-21, V-11
and W-11 of Fig. 3. It is to be pointed out that these measure-
ments were made before the shelf-type mock-up was In final form
and when the length of the beryllium jacket was 44" with no.
graphite end reflector behind the beryllium. The initial purpose
of the experiment was to compare the reactivity value of a re-
flector element with that of a fuel element in order to ascertain
the most satisfactory locatlion of a safety rod. For this reason
only a 22" length, l1.e. half of the reflector length, of beryllium
was removed and the resulting reactivity change ascertained from
a calibrated control rod. In each case the data compare the
beryllium filled tube with an empty one. The error in the re-
sults, as given in Table II, is estimated to be. ¥ 1¢.

TABLE II

REACTIVITY CHANGE INTRODUCED BY SUBSTITUTING BERYLLIUM FOR AIR

 

 

Test Cell Reactivity Increase in Cents
M.21 bk
V-11 10.9

V-11 and W-11 27.5

The reactivity change for cell M-21 was checked by the rod drop
mothod and found to be the same within limits of experimental
error. This change 1s small because of the presence of the one
inch layer of Plexlglas in the core cells adjJacent to this region
of the reflector. The changes obtained for cells V-1l and W-1ll
were expected to be larger since the removal of these two re-
flector cells opens a partial. voild extending deep into the reactor.

In a second series of experiments, with final reactor geometry,
a section of the beryllium reflector was replaced by a composite
of Plexiglas and stainless steel, each one-half the beryllium
thickness. The beryllium in a section 9" wide, 36" long and 6&"
thick™ was replaced by a 9" x 36" x 3" section of Plexiglas ad-
Jacent to the core and backed by a section of stainless steel,

 

* It mist be noted, that since the outside dimensions of the square aluminum
tubing 1s 3" x 3", the cross-section of the units of both the reflector and
the fuel is 2.7/8" x 2-7/8", the remainder being the aluminum tubing wall
and air space. For convenlence in discussion the 3" dimension, and multiples
thereof, will be used to designate the reflector alterations. In the quan-
titive presentation of data, actual thicknesses of material will be stated,
with the understanding that the volds are present.

- -29-
. .
Type 310", of the same dimensions. This change was made separately
at three positions at the side of the reactor, at the top and at
the bottom.

The results, giving the changes in reactivity effected by the sub-
stitutions and the locations of the reflector alterations using the
designations of Fig. 3, are tabulated in Table III. The system was
more reactlive with the beryllium reflector than with the combination
of stainless steel and Plexiglas, Lo o e : S

s AR ’
- . ) St
_,'. ,‘J. v .. . Lo e

TABLE IIT

REACTIVITY CHANGE EFFECTED BY SUBSTITUTING PLASTIC AND STAINLESS
STEEL FOR BERYLLIUM

 

 

Position Original | Altered Reflector: Reactivity

 

 

Reflector: Plastic Stalnless Loss

Bg In Cells in Cells Steel in Celle 1in Cents
Side v, w-11, 12, 13 V-11, 12, 13 w-11, 12, 13 - 56.5
Top L, M, N-2, 3 L, M, N-3 L, M, N-2 41.3
Bottom L, M, N-21, 22 L, M, N-21 L, M,.N-22 33.0

2, Variable Reflector Void Width and Thickness of Stainless Steel and
Plexiglas Constant.

 

A larger slab in the reflector, 15" wide, 36" long and 6"
thick, occupying cells K-21 through 0-21 and K-22 through 0-22,
inclusive was gelected and measurements made in the following
order. The beryllium was removed in 3" x 36" x 6" thick units,
the first being at the center, M-21 and M-22, next N-21, 22, .then L-21
and L-22, etc., until all cells in the group were vacant, the system
being made critical between each removal., It was, of course, ne-
cessary to add some additional beryllium reflector near the top and .
gides of the reactor to override the reactivity loss due to this re-
flector removal at the bottom. The successlve changes in reactivity,
due to these reflector alterations, were noted from the critical po-
sitions of the control rods. The void was then filled with a com-
posite reflector of 1-7/8" of Plexiglas and 3-7/8" of stainless steel,
the former being adjacent to the core. This composite reflector was
loaded step-wlse in the reverse order to the removal of the beryllium,
and the successlve increments in reactivity measured.

Table IV contains the losses in reactivity for various void widths and
also the losses per unit width of the void as the beryllium 1s re-
moved from the reflector.

¥ An analysis of the steel 1s glven in Appendix B,

-30-
TABLE IV

REACTIVITY LOSSES FOR VARIABLE VOID WIDIE IN REFLECTOR

 

 

Vold Width Reactivity Loss Reactivity Loss
in Inches _ in Cents in Cents per Inch
0.0 0.0 — |
3.0 19.5 6.5
600 36-0 5‘5
3.0 ' 52.0 5.3
12.0 6k4.5 4,2
15.0 T7.0 4,2

Table V containg the losses due to the substitution of 2" of
Plexiglas plus 4" of stainless siteel for the 6" of beryllium in
these bottom reflector cells, referred to zero loss for a complete
beryllium reflector. .

TABIE V

REACTIVITY IOSSES EFFECTED BY SUBSTITUTION OF STAINLESS STEEL AND
PLEXTGLAS FOR BERYLLIUM

 

 

 

Width of Section Reactivity Loss Reactivlity Loss
Substituted in Inches in Cents in Cents per Inch
0.0 0.0 -

3.0 T.5 T 2.5
6.0 13.9 2.1
9.0 20.6 2.0
12,0 26.5 2.0
- 15.0 32.5 T 2.0

Figure 11 is a graph of the losses, referred to a complete
beryllium reflector, plotted-as a function of the vold width when
the reflector was removed. Thls graph also shows the losses in-
curred by substituting the 2" of Plexiglas and 4" of stainless
steel for the 6" of beryllium as a function of the same vold width.
Figure 12 1s a graph of the loss per unit width plotted as a function
of the same distances, that is, the slopes of curves in Fig. 11. As
the leakage path becomes wider, the Incremental value of th®d reflector
approaches a constant value.

3. Variation of Bottom.RefléEtor Composition

In another serles of studles in these cells, K-21 through 0.21
and K-22 through 0-22 the relative thicknesses of Plexiglas and
stainless asteel in the composlte reflector were variled for two over-
‘all thicknesses. For one series of measurements the thlcknesses of
the individual layers varied from zero to the nominal six inches and
in the second merles from zero to four inches. The accompanying
changes in resctivity were noted for sach incremental thickness.

-31-
IN CENTS

DECREASE

IN REACTIVITY

90

80

70

N
o

owc. 21342 N

32

 

 

| I | | | I

 

50 —
6" BERYLLIUM
40 —
)
[ )
30 —
)
2" PLEXILGLAS AND
20 A - 4" STAINLESS STEEL —
®
10 —
9
& | I | I | |
3 6 9 12 15 18

VOID WIDTH OF REFLECTOR TEST SECTION IN INCHES

" FIGURE 11. DECREASE IN REACTIVITY VS. VOID WIDTH

 
REACTIVITY LOSS IN CENTS PER UNIT VOID WIDTH

DWG. 21343 —

33

 

 

I I I I I I

  
 

6" BERYLLIUM

 

 

2" PLEXIGLAS AND
4" STAINLESS STEEL

] I | | I I

 

3 6 9 12 15 18
VOID WIDTH OF REFLECTOR IN INCHES

FIGURE 12. REACTIVITY LOSS PER UNIT VOID WIDTH

 
Table VI contains data showing the losses in reactivity
as & function of the material thickness in the composite re-
flectors for both the nominal four-inch and six-inch thicknesses.
The changes refer to the reactivity with the beryllium reflector.
The results are plotted in Fig. 13. It is to be noted in Fig. 3,
that this test section of the reflector is adjacent to a 1" thick
layer of Plexiglas which is nominally a part of the core. The
measurements were extended by replacing this Plexiglas with stain-
less steel, that 1s, extending the stainless steel 1" into the core,
giving a structure simulating that at the top of the reactor. Fig-
ure. 13 1s plotted with this 1" layer of Plexiglas included in

‘the reflector.

TABLE VI

LOSSES IN REACTIVITY EFFECTED BY VARYING COMPOSITION OF BOTTOM REFLECTOR

 

 

 

Four-Inch Composite Six-Inch Composite
Material Thickness Material Thickness
in Inches Reactivity in Inches Reactivity
Stainless - Losses Stalnless Losses
Plexiglas Steel in Cents Plexiglas Steel in Cents
0 L.7/8 49.2 o 6-3/k W7
1/2 4-3/8 39.7 1/4 6-1/2 37.2
3/L 4-1/8 39.8 1/2 6-1/h 35.3
1 3-7/8 40.5 3/h 6 35.6
1-7/16  3.7/16 42,1, 1 '5-3/k4 36.5
1-7/8 3 . 43.6 . 1-3/16  5-9/16 o 37.7
2.7/8 2 4.7 - 1-7/16  5-5/16 39,2
3-7/8 L 46.1 1-7/8 4-7/8 41.8
4-7/8 0 45,7 2.7/8 3-7/8 44,9
. | 3-7/8 2-7/8 45.9
L-3/L 2 L6.3
5-3/& 1 4W6.7 -
3/& 0 46.8

Variation of 81de Reflector Composition

Since the planes of the fuel and the Plexiglas moderator in
the shelf-type mock-up were horizontal it 1s apparent that the
reflector savings per unit of reflector perimeter was a function
of its location. For this reason the above experiment was re-
peated in a test section 9" wide, 36" long and 6" thick at the
side ‘'of the core using cells V-11 through V-13 and W-11 and
through W-13. The width of the test sectlon was reduced from 15"
to 9" in order to remain critical without increasing the loading
or adding excessive extra beryllium to the reflector.

Table VII contains data showing the losses in reactivity as
a function of the thickness of the reflector materials., These
deta are again referred to a complete beryllium reflector as rep-
resenting zero reactivity loss. Filgure 14 is a graph of the losses
in reactivity plotted as a function of Plexiglas thickness.

~3h-
DECREASE IN REACTIVITY IN CENTS

 

| DWG. 21344 e
° | I | [ I I | |
REACTIVITY WITH COMPLETE
N BERYLLIUM REFLECTOR
30

35

40

45

50

55

 

@® 6" THICK COMPOSITE REFLECTOR

 

 

O 4" THICK COMPOSITE REFLECTOR

 

 

O
—o——@
——EDGE OF CORE
| ] | | | I | |
I 2 3 4 5 6 7 8
THICKNESS OF PLEXIGLAS IN INCHES
FIGURE 13. CHANGE IN REACTIVITY VS. COMPOSITION OF REFLECTOR FOR 15" SECTION

AT BOTTOM IN CELLS K THROUGH 0-21,22

GE
TABLE VII

LOSSES IN REACTIVITY EFFECTED BY VARYING THE COMPOSITION OF THE
SIDE REFLECTOR

Reflector Material Thickness in Inches Reactivity Loss

 

Plexiglas . - Stalnless Steel in Cents
0 5-3/4 61.4
3/16 5-9/16 57 +5
7/16 5-5/16 54,2
7/8 L-7/8 53.3
1-7/16 4L-5/16 5k L
1-7/8 3-7/8 55,5
2.7/8 2.7/8 56.4
4-3/4 1 57.2
5-3/L 0 57,1

The ourve for the 9" section is extrapolated to a width of
15", bagsed on a comparison of reactivity change produced by
replacing beryllium reflector at the bottom, rows 21 and 22,
with a composite one of stainless steel and Plexiglas.

5. Effect of Voild Between Core and Reflector

A preliminary experiment was run in which sbout one-half of
‘one. 6" graphite-end réflector was withdrawn one Inch leaving a gap
between the end reflector and both the core and the beryllium slde
réflector.. . The cohcomitant’ reduction 1in reactivity was 25.1 cents.
When this:.gap was betweén the: end .refilector and the core only, that
18 with .the graphiteiin: céhtact withthe beryllium' jacket,. the re-
activity decrease was only 17.4 cents.

D. DPower Distribution and Neutron Flux Determination in Shelf Type Assembly:
1. Description of Foil Techniques and Counting Equipment

If an aluminum foil is in contact with uranium fuel while =
reactor i1s in operation. the activity of the fission fragments
collected on the aluminum is a measure of the fission rate at
the surface of the fuel disk. Fission rates measured by this
method Bave ‘been shown to be the same as those obtained from counmt-
ing the uranium disks activated in thils reactor. Essentlally all
power distributions reported here were obtained by this "catcher
foil" method., The foils used were 28 aluminum disks, 2" in dlameter
and 0.005" thick with a 0.5" hole in the center. These foils, in
all cases, were exposed 20 minutes, counted in a Radiation Counter
Laboratories gas flow proportional counter. Their activitles were
normalized to 60 minutes after shut-down using an empirically de-
termined decay curve?. The counting time for each catcher foil
varied from 2 to L4 minutes which gave a minimum of 10° counts. The
average variation in successive counts from the same foil 1s slightly
less than 2%.

5 Ibid. Appendlx B, p. 66.. . L
-36.-.
DECREASE IN REACTIVITY IN CENTS

DWG. 21345 — 37

 

 

 

 

   

 

 

 

1 \ [ I [ I |
REACTIVITY WITH COMPLETE BERYLLIUM REFLECTOR
50— |
CURVE FOR 9" SECTION
55— —
—0)— O
60i——— —
65— —
e+———— EDGE OF CORE
70— —
75 CURVE FOR 9" SECTION
— EXTRAPOLATED TO (5" WIDTH
80— —_
85— —
90 f— —_
o5 | | 1 | 1 |
0 | 2 3 4 5 6

THICKNESS OF PLEXIGLAS IN INCHES

FIGURE I14. CHANGE IN REACTIVITY VS. COMPOSITION OF REFLECTOR
FOR 9" SECTION AT SIDE IN CELLS V-1l THROUGH I3 AND W-II

THROUGH 13
Indium foile were irradiated in the critical assembly for
a study of the neutron flux distributions. These foils were
5/16" in diameter, 0.01" thick and weighed 0,0288% 0.0001 grams.
They were composed of 90% aluminum and 10% indium by weight*,
with an equivalent indium thickness of 0.0004". These folls
were covered with cadmium 0.02" thick when 1t was desirable to
shield from the low energy neutrons. All indium foll activa-
tions were determined with an Atomic Instrument Co. Model 1010
binary scaler using an Amperex Model 120-C mica window**Geiger
tube. Each foil was counted for two minutes on both sides in
each of four counters and the eight values were averaged after
they had been corrected for dead-time, background, decay and
counter sensitivity-. The average varlation between successive
counts of a given foil is about 3%. ‘

Since 1t was not possible to expose all the indium foils
necessary for one traverse simultaneously hecause of the result-
ing perturbations arising from interactions of the foils, 1t was
necesgsary to normalize the exposures .for varlations in power level,
An aluminum catcher foll was located in the same position in the
reactor during each exposure and the appropriate ratio of their
activities was used for these normalizations.

Pigure. 1518 a ' photograph.of some of theimaterials used in
obtaining power and flux distributions. An aluminum catcher foll
1s located on one of the uranium fuel diske and some indium foils
are seen between the disks. The countersunk beryllium blocks were
used when the neutron flux was measured In the reflector. Some
cadmium boxes, used to cover the iIndium folls, are also shown.

2. Power Distribution in Shelf-Type Mock-Up

 

Three radial power distribution curves taken in the midplane
of the assembly, along a horizontal radius, a 45° radius and
along a vertical radius, are shown in Fig. 16. These power trav-
erses originate in Cell M-12, Fig. 3, and extend to Cells U-12, R-T
and M-U4 respectively. The vertical power traverse, which terminates
in Cell M-L, shows a comparatively large decrease in power as it
approaches the edge of the core. A zero order Bessel function of
first kind curve is included in'this figure for comparison.

The power traverses shown in Fig. 17 were measured parallel
to the axis of the assembly, one being along the axis through
Cell M-12, and the others along the periphery of the core, orig-
inating in Cells U-12, R-7 and M.L4 at the termini of the three
radial power traverses. It should be noted that some of the data
were obtained from the aluminum catéher foils while other are
measurements of the fission product activity induced In the uranlium

* This alloy was prepared by the Oak Ridge National Laboratory Metallurgy
Division using the method of R. F. Lutz of General Electric Co. described
in a memorandum IC-51-3-29, March 1k, 1951.

#* This window was 2" in diameter and had a thickness of 5.6 ng/cm?,

6 Ibid., Part III p. 25. . .
=30~
Fig. 15. Photograph of Materials Used in Power and Neutron Flux Determinations.

 

 
ACTIVITY

RELATIVE

HORIZONTAL

45° (QUADRANT I)

VERTICALL
upP

 

6 9 12 IS5 18 21 24 27 30
INCHES FROM AXIS OF REACTOR

FIGURE 16. GROSS POWER DISTRIBUTATION RADIALLY IN MID-PLANE OF REACTOR

33

HELC "OMA

or
DWG. 21347

o
©

o
®

o
~

GRAPHITE
REFLECTOR

o

RELATIVE ACTIVITY (NORMALIZED)
o

o
>

CELL M-12 URANIUM DISC DATA
CELL M-I2 AL FOIL DATA
CELL u-12 ° " "
CELL M-4 " " "
CELL R-7

 

 

3 6 9 12 15 18 21 24
INCHES FROM REACTOR MIDPLANE

FIGURE |7. POWER DISTRIBUTIONS EXTENDING FROM REACTOR MIDPLANE TO GRAPHITE
END REFLECTOR

  
    

r
disks. The results from the two methods are not distinguishable,
validating the catcher foil technique, at least for this reactor.

The best fit zero order Bessel functions of the flrst kind
which represent these power distribution curves extrapolate to
a core dlameter of 78" and a length of 49". Using these values
of extrapolation dlstances, reasonably good agreement was ob-
tained between the measured and calculated values of reactivity
losses due to separation of the two halves of assembly at the
midplane, which are tabulated in Section IV, Part D of this report.

Power Distribution in Unit Cell

 

Because of the inhomogeneity of both moderator and fuel iIn
this assembly, the distrlbutlons of power and neutron flux inside
the unit cell are of interest. Figure. 18:shows’ the vertical power
distribution in Cell M-12 when 0.002" thick fuel dieks are in :
Cells L-12, M-12 and N-12 and parallel to the planes of the Plexi-
glas layers. This distribution was obtained by placing an aluminum
catcher foll on each face of the thin fuel diske. The curve drawn
through these relative activities assumes all the absorption be-
twoen the fuel disks occurs in the stainless steel. The average
activity over the cell is approximately 89% of the maximum value.
The self shielding of the fuel disks and the fuel cadmium fractions
are discussed in detall in Part IV, Section G of this report.

 

Neutron Flux Distribution in Shelf-Type, Mock-Up

Macroscoplc, or gross, neutron flux distributions were made
by placing, successively, bare and cadmium covered indium folls
at an equivalent position wilthin each of the cells In the reglon
of interest. The position chosen was in the fuel plane between
the uranium disks as shown in Fig. 15, The results of the hori-
zontal traverses, taken radially at the mid-plane, and the axial
traverses are shown in Figs. 19 and 20, respectively. In Fig. 19
the foll activity 1s plotted as & function of the dlistance from
the axls of the reactor, parallel to the mid-plane, and in Fig. 20
the activity is plotted as a function of the distance from the mid-
plane along the axis of the reactor in Cell M-12. The vertical
radial traverse, Fig. 21, was also taken in the mid-plane and ex-
tends from cell M 12 vertically dowvnward through the beryllium
reflector. It is to be noted that in this lower region of the core
the beryllium reflector is adjacent to a 1" layer of Plexiglas. No
folls were placed in this layer of plastic during these traverses
and the interpolation is based on exposures in a slngle cell which
will be described later. A curve of a zero order Bessel function
of the first kind 1s also included in this figure for comparison.

Flux Distribution in A Unit ggl}

 

Inhomogeneous loading of a reactor produces local variations
in power and flux distributions which are of importance in re-
actor calculations. Bare and cadmium covered traverses were made
in & unit cell where both fuel thickness and -orientation could be
varied. The unit cell used was always near the center of the re-

“actor where the flux:;lis relatively.constant Over a 3".test.section.

42
RELATIVE ACTIVITY

1.0

o
©
o

0.90

0.85

0.80

PLEXIGLAS

1.O

FIGURE

18. POWER

Av.=893 %

I7 mil ' PLEXIGLA
STAINL

STEEL

 

LS ‘ 2.0 2.5 3.0
INCHES FROM BOTTOM OF UNIT CELL

DISTRIBUTION IN UNIT CELL USING 2mil U-DISCS
(STAINLESS STEEL AND U-DISC NOT DRAWN TO SCALE)

8yelZ "OMa

ey
ACTIVITY

RELATIVE

FIGURE

I9.

GROSS

9 12

BARE INDIUM ACTIVATION

BE REFLECTOR

CADMIUM COVERED INDIUM ACTIVATION

BARE INDIUM ACTIVATION MINUS
CADMIUM COVERED INDIUM ACTIVATION

 

15 18 21 24 27 30

INCHES FROM AXIS OF REACTOR

INDIUM TRAVERSE

HORIZONTAL IN PLANE OF |INTERFACE (M-12 THROUGH U-12)

33

6¥EILZ "OMA
RELATIVE ACTIVITY

BARE INDIUM ACTIVATION

CADMIUM COVERED INDIUM ACTIVATION

BARE INDIUM ACTIVATION

MINUS CADMIUM COVERED
INDIUM  ACTIVATION

3 ' 6 9 12 15
INCHES FROM REACTOR MID
FIGURE 20. GROSS INDIUM TRAVERSE ALONG

DWG, 21350

GRAPHITE REFLECTOR

I8 21
PLANE

REACTOR AXIS (M-i2)

 

24

S
ACTIVITY

RELATIVE

FIGURE 21

GROSS

BARE

x)

CADMIUM COVERED

BARE
COVERED

9 12 15

INDIUM ACTIVATION

CURVE

INDIUM ACTIVATION

INDIUM ACTIVATION MINUS CADMIUM

INDIUM  ACTIVATION

18 21

INCHES FROM AXIS OF -REACTOR

INDIUM TRAVERSE VERTICALLY DOWN

IN PLANE OF

BE REFLECTOR -

24 2r 30

INTERFACE (M—I2 THROUGH M-22)

 

33

1612 "OMaQ
In a normal unit cell the fuel disk is located 2" from the
bottom of the test cell and 1" above the Plexiglas, producing

a flux in the cell which is symmetrical above and below the fuel
disks. The vertical flux traverses are, thersfore, symmestrical
about the fuel plane.

Figure 22/18 aidrawing of a:unit dellwilth: thetstainless
steel removed to show the relative positions of the fuel disks
and the outline of the Plexiglas in the bottom of the cell. The
lines A-A, B-B, etc. show the locatlons of various traverses
made through the cell. Figures 23, 24and 25! ghow thres vértical
flux traverses, perpendicular to the fuel plane, taken through a
unlt cell along the paths A-A in Cell M-12, B.B in Cell N-12 and
C-C in L-12 respectively. The reductions in neutron flux which
occur at the fuel disks are the results of shadow shielding by the
fuel and the neutron sinks produced by lumped fuel. The reduction
1s less pronounced in the A-A traverse, Fig. 23, because the indium
foil, in this traverse, was located over the center hole of the fuel
disk and the shadow shleldling was greatly reduced.

Three vertical neutron flux traverses were also made in a
unit cell between the fuel disks and the results are shown in
Fig. 26, 27 and 28. The increase in flux occurring at the fuel
plane along traverge D-D, Fig. 26, was verified experimentally,
although its cause is not fully known. Filgures 27 and 28 . indicate
that the flux along E-E and F-F, between the Plexliglas layers,
is essentially constant. -

Determinations of neutron flux distributions in a unit cell
vere also made with one 10-mil fuel disk replaced by five 2-mil
thick disks. Only fifteen 2-mil disks were available so a large
test volume was lmpossible. The thin disks were substituted in
cells L.12, M-12 and N-12 near the reactor midplane with their
planes parallel to the planes of the Plexiglas. The thin disks
vere equally spaced between the Plexiglas layer and the top of
the cell. The meutron flux distribution in cell M-12, along path
A-A Fig., 22, 1s shown in Fig. 29. Thie- change caused the average
neutron induced activity in cell M-12 to decrease by about 8 percent.
In these traverses the indium foils were placed 0.25" from the center
of the fuel disk to avold the effect of the hole. The ratioc of the
average thermal neutron activity in the Plexiglas to that in the
space between two adjacent layers of the plastic, with the Z2-mil
disks loaded is 1.82 which 1s not significantly different from the
ratio wheén a 10-mil disk is used., The results of an experiment -
measuring the ratio of thermal neutron flux inside and outside the
Plexiglas will be glven later.

Another set of experiments was done with five 2-mil fuel dilsks
each in Cells L-12, M-12 and N-12 placed perpendicular to the Plexi-
glas layers and parallel to the axis of the core. The thin uranium
disks were equally spaced between the vertical sides of the unit cells
and extended into narrow slots cut into the Plexiglas. Vertical trav-
erses, centrally located in the fuel disk array, were made in cell M-12
with the indium foils perpendicular to the Plexlglas and repeated wilth
the foils parallel to it. The activity of the folls which were ‘

7.
 

 

 

48
*
\ *
\
‘. * “ \
II \ fa
\
\ \ %
\ * \
A \ .
\ \ \
\
\ \ D
\ 1 \ »
\ 1 \ \
\ 1 \ \
A | \ .
\ 1 Y \
. 1 . .
\ i / N
\
. .— \ . /
1 \ ! '\ \
/ * ! \ \ *
\ \ ] “ N\
— —— — — i, S e— — — p—— pp— — —— —
w —_ ~ —— - - J
\ . / / \
11 \ 4 \ N
~ \ \\\ \\I.l.lll. a- \
oQ——— llllll.l'll\’fll. A IL/..II"}H'II'I - —_— 0
- g e o /
N ’ A \ N
3 s\ 5 * /
vl_d v 7 / lt .
« . \ \ “
— 2! 1 . .
o™~ A ) -a "\
- [ ¥ . '\
o i s \ \ .
3 w—_—— l[.ll.%lll._sl-.ll I..ll..__.lu..ll.lla'llll. e o
\ . \
a -q If * at "
— . “ of ao
\ .
_p 1’ -— / —/
\
\ / 1 * \
4 \
\ \ i / .
\ \ ! Y \
\ z- 7 Y //
A \ ‘ \
v
\ 3 ! \ N\
- '\ / . 1
\ L 7 * '
tl / / A a-
S 4\\\ \\llll/ \ %
I-lllll.l\\\ \ \\ . -— K
A /7 N\ . .
) 7 x \ .
Y ! \ " Y
- ! \ .
Xs / " s,
1y _- / *
.
1 L)
—§ f— ! pa /
.
i \ ! \ AN
o — —— III|.+]|I',| —_— e — - —_—
1 \ | v .
\ \ 1 . \
\ \
L X
\ \
\ A
\ * DY
\ v A
QO — e ey c— — < — — —— —_—
\
a2
»
a

FIGURE 22. TYPICAL CELL { WITHOUT STAINLESS STEEL)
RELATIVE ACTIVITY

0.9

o8

Bare Indium Activstion

0.7

0.
o.s
' Indium Activation
0.4 Mimus Cadmiwm Covered
' Tndium Activation

wm Covered
Indium Activation

€GEIZ "OMa

0.2

0.1 Plexiglas Stainless Steel tainless Steel
and Air and Air

 

0 I 5

Distance irom Bottom of Cell i-12 in Inches
‘FIGURE 23, INDIUM TRAVERSE IN UNIT CELL LOCATION: A-A IN M-I12 (SEE FIGURE 22.)

&y
0.9

0.8

0.7

0.6

o4

RELATIVE ACTIVITY

0.3

0.2

0.l

o
0o

FIGURE 24.

Bare Tndium
Activation

C um Cov
Indium Activation

e Indium Activation
inus Cadmium Covered
Indium Activation

Plexiglas Stainless Steel tainless Steel lexiglas
and Air and Air

INDIUM TRAVERSE

 

2 3.0 |
Distance from Bottom of Cell N-12 in Inches

IN UNIT CELL LOCATION: B-B IN N-12 (SEE FIGURE 22)

¥SELE "OMd

oS
RELATIVE ACTIVITY .

0.9

0.8

Bare Indium Activation

0.7

  
  

0.6
0.5
O \ . .
: 3 Bare Indium Activation
Minus Cadmium Covered
Indium Activation
) Cadmium Covered
0.3 Indium Activation o
=
©
N
0|2 ;
o
0.1 Stainless Steel Stainless Steel Plexiglas

and Air and Air

0

o ! 2 3.0 1
Distance from Bottom of Cell 1-12 in Inches

FIGURE 25. INDIUM TRAVERSE IN UNIT CELL LOCATION: C-C IN L—I2 (SEE FIGURE 22)

n
—
RELATIVE ACTIVITY

009

0.8

0.7

0.6

05

0.4

0.3

0.2

0.l

0

o

FIGURE 26.

Tridiom
Activation

Indium Activation
Minus Cadmium Covered
Indium Activation

Cadmium Covered
Indium Activation

Stainless Steel

Alr

 

2 : 3.0 I

Distance from Bottom of Cell M~12? in Inches

INDIUM TRAVERSE

IN UNIT CELL LOCATION: D—D IN M-I2 (SEE FIGURE 22.)

9SE1Z 'OMa

A%
RELATIVE ACTIVITY

Bare Indium Activation

Bare Indium Activation
. Minus Cadmium Covered
\, Indium Activation

~#3 admium Covered
Indium Activa

Stainless Steel Steinless Steel
and Alr and Air

 

{ 2 3.0 |
Distance from Bottom of Cell N-12 in Inches

FIGURE 27. INDIUM TRAVERSE IN UNIT CELL LOCATION: E-E IN N-I2 (SEE FIGURE 22.)

LSELT "OMa

£S
RELATIVE ACTIVITY

FIGURE 28,

Stainless Steel Stainless Steel

and Air

and Air

Distance from Bottom of Cell I=-12 in Inches

INDIUM TRAVERSE

IN UNIT CELL LOCATION: F-F INL-I12

Indium Activation

Indium Actl
Minus Cadmium Covered
Indium Activation

Cadmium Covered
ndium Activation

 

(SEE FIGURE 22.)

8GE1Z "OMa
RELATIVE ACTIVITY

0.8

0.7

0.6
0.5
0.4
um Covered
Indium Activation

0.3 o

' =

®

Indium Activati ;o

- Minus Cadmium ;;

0.2 Indium Activation G

°o|‘

Plexiglas 58 Steel, Fueland Air lexiglas

 

0
0 ! 2 3.0 i

Distance from Bottom of Cell M=12 in Inches

FIGURE 29. INDIUM TRAVERSES IN UNIT CELL M-12 WHICH HAS 5-2mil U-DISCS PARALLEL &
TO PLEXIGLAS
1v,

oriented perpendicular to the Plexiglas was about 6% less

than that in those which were parallel. The two sets of data
wvere normalized at a point 2" from the bottom of the cell and
are shown in Fig. 30. The minimum in the flux distributlon
oceurs at a point 1.5" from the bottom of the cell which 1s the
location of the center of mass of the fuel disks.

A measure of the flux distribution across the dlameter of
a 10-mil fuel disk was obtained from bare and cadmium covered
indium traverses across a disk located in cell M-12 and the
results are shown In Fig.3l. No foils, however, were located
adjacent to the edge of the disk.

The ratios of the average thermal flux, i.e., below the
cadmium cut-off, in the Plexlglas to that In the space between
the Plexiglas layers for this group of unlt cell traverses are
given . in Table VITIT, i, oG drois sl

TABLE VIIT
THERMAL NEUTRON DISTRIBUTION IN A UNIT CELL

Av., Flux in Plexiglas
Location of Traverse Average Flux Average Flux Between Av. Flux between

 

 

 

in the Unit Cell in Plexiglas Plexiglas Layers Plexiglas layers
A-A 4080 2275 1.79
B-B 3610 2345 1.54
C-C 3690 2300 1.60
D-D 3980 2650 1.50
E-E 3900 0655 1.47
F.F 3700 2750 1.35

EXFERIMENTAL STUDIES. DN RECTANGULAR/"SHELL! TYPE ASSEMBLY
A. Control:Rod Calibrations In: The Rectangular Shell Type Assembly:

The control rods in the rectangular shell type mock-up were call-
brated by the super-critical period method as described previously.
The rod drop method was not used for compariscon. The reactivity values
obtailned for three of these rods are comparable to those obtalned when
the planes of all the core materials were horizontal. The fourth, Rod
A, is significantly more effective in the rectangular shell type mock-
up due to the change on the neutron distribution resulting from the
omigssion of the Plexiglas from this rod and also surrounding the rod
on all four sides by one inch layers of the plastic. It is interesting
to note that the rod calibration was not changed by a gap, in the axial
direction at the midplane of the core, of thickneas up to 0.8": .. -

Detailed data for rod A are given in Table IX and a summary for
all four rods in Table X.

-56-
RELATIVE ACTIVITY

0.7

0.6

0.5

0.4

0.2

0.l

o
o -

FIGURE 30. INDIUM

LOCATION : A—A

Stainless Steel, Fuel and Air

um

Bare Indium Activation
Bare Indium Activation
Cadmium Covered Indium

Cadmium Covered Indium

Indium Folls Parallel to Plexiglas
@ Bare Indium Activation

Bare Indium Activation
Minus Cadmium Covered
Indium Activation

 

Distzance from Bottom of Cel]? M=12 in Inches
TRAVERSE IN UNIT CELL
PLACED PERPENDICULAR TO THE: PLEXIGLAS AND REPLACE |—I10 mil U-DISC

IN M—-12 (SEE

FIGURE 2.

IN

)

WHICH 5-2 mil U-DISCS ARE

09€LT "OMa

S
RELATIVE ACTIVITY

0.8

0.7

0.6

0.5

04

0.3

0.2

0.1

O

0o

FIGURE 3l.

| 2 3

Distance from Midplane in Inches

INDIUM TRAVERSE ACROSS A 0.010" FUEL DISC

Indium Acetivation

um Covered Indium
Activation

Bare Indium Activation
minus Cadmium Covered
Activation,

 

4

LOCATION: M~— |2

19€1Z "OMQ

8¢
TABLE IX

CONTROL ROD "A" CALIBRATION

 

Position of Rod "A"

 

 

from Core Mid-plane Integrated Average Sensitivity
in Inches Value in Cents In Cents per Inch
0,00 _ 0.00 --
3.23 3.42 | 1.06
4.93 6.80 1.98
6,43 9.81 2.01
8.35 13.08 1.79
10.30 16.95 1.99
13.73 ' 21,22 1.25
17.01 . 24,58 . 1.02
21.39 Lo 27.39 ‘ 0.64
26,09 | 28.79 , 0.30
TABIE X

SUMMARY OF CONTROL ROD CALIBRATION DATA -

Reactivity Value

 

 

Control Rod - | Lengthe in Inches ~ in Cents
A 26, 28.8
B 26, 17.1
C 254, ’ 15.6
D 254, 17.0

Temperatire Effects in Rectangular Shell Type Assembly:

The effect of temperature on reactivity in the rectangular shell
type assembly was measured by the same procedurs and over approxi-
mately the same temperature range as was used in the shelf type
assembly.  The ihcreéase .in:.reactivity with increasing température
ébeerved was 1.43 cents:per Fahrehheltndegree, compared.with the.
sarlier value of 1.25 cents per Fahrenhelt degree.

Plexiglas Thickness Reactivity Coefficient:

The thicknese of the Plexiglas moderator strips in five cells,
K-16 through 0-16, was varied and the changes in reactlvity produced
were determined from the control rod positions. In Fig. 32 the
changes in reactivity in cents referred to the reactivity with one
indh‘strips,;are;plbttéd.as.affunCtion:dftthe-Plexiglasfithibkness.

This curve shows an optimum thickness of approximately 0.9 inches

for this region of the reactor at the operating temperature.
Effect of Separation of Core at Midplane on Reactivity:

It wag desired to determine the loss in reactivity due to the
separation of the two halves of the reactor core, the limit of the
experiment to be a gap of one inch or a reactivity decrease of 100
cents, whichever occurred first.

=59~
IN REACTIVITY IN CENTS

CHANGE

-0

owc. 21362 (N

 

 

]

L | L

| |

 

 

 

0.5

FIGURE 32. CHANGE

0.6 0.7 0.8 09 1.0 1.1
PLEXIGLAS THICKNESS IN INCHES

IN REACTIVITY VS. PLEXIGLAS THICKNESS

IN TEST CELLS

09
The procedure used was to start with the two halves together
and, as they were separated in an axial direction, compensate for
the loss in reactivity by driving in control rods. The reactivity
loss for each increment of separation was then determined from the
control rod calibration curves. When all the control rods were in,
extra beryllium reflector was added to strateglc areas and evaluated,
allowing withdrawal of the control rods, and the process was repeated,
with soms overlapping in distance, until the desired magnitude of
separation had been achieved.

Table XI shows the reactivity losses for various separation
distances and the average values of the reactivity losses per unilt
distance over successlve Ilncrements of separation,

TABLE XI

EFFECT OF CORE GAP ON REACTIVITY

 

 

Separation Dilstance Reactivity Loss Average Sensitivity
at Mid-plane in Inches in Cents in Cents per Inch
0.03 1.92 61.9
0.06 L.37 Th.25
0,10 7T.39 81.6
0.1k 11.20 90.5
0.19 16.41 . 99.8
0.26 23.88 114.6
0.34 33.75 124 .2
0.43 45.83 133.2
0.46 50.00 138.1
0.26 25.30 99.4
0.43 48.17 131.4
0.51 2907 139.0
0.59 T1.43 148.2
0.63 77.13 _ 150.1
0.29 26,61 9l.1l
0.4k 16.05 . 134.1
0.55 62.50 . 150.9
0.60 71.23 . 143.9
0.6k T7.95 158.1
0.70 86.22 159.0
0.75 94 .23 . 160.2
0.79 102.2 & . - 162.3
0.81 105.7 164 .8

When separated, the facing sides of the core halves are scmewhat
non-parallel introducing an inaccuracy in the separation distance
estimated. to be about 0.02". In Fig. 33 the solid curve shows
the values of the measured reactivity losses as a function of the
distance between the two halves of the core when separated at the
midplane:. The broken line in this figure represents the calculated .

‘~values of the reactivity losses using the method suggested by'Tamor7l

7 Tamor, S.,"The Effects of Gaps on Pile Reactivity; ORNL-1320, July 1k,
1952, pages 11, 13.

~61-
REACTIVITY LOSS IN CENTS

DWG. 21363 _

 

120

100

80

60

40

20

 

 

  

   

EXPERIMENTAL CURVE

 

 

0.1 0.2 0.3 0.4 0.5 0.6 0.7
SEPARATION DISTANCE AT MIDPLANE IN INCHES

FIGURE 33. LOSS OF REACTIVITY VS. SEPARATION OF TWO HALVES

0.8 0.9 1.0
R

OF CORE AT MIDPLANE
and treating the assembly as a bare homogeneous reactor. The
reactor constants used were those recorded by Leverett®.

Reflector Studies on The Rectangular Shell Type Assembly:

1.

Stainless Steel, Plexiglas and Boral Composite Reflector

The test cells used in these reflector studies were K-21
through 0-21 and K-22 through 0-22. The beryllium reflector-
was removed from these cells, on both halves of the assembly,
and replaced by stainless steel, except for a 5/16" layer
adJacent to the core which was filled with Boral* strips 36"
long. The system was then made critical by the addition of
beryllium reflector at locations remote from the test section
and by control rod adjustment, thereby measuring the loss in
reactivity. Keeping the overall dimensiocns of the test section
constant, the thickness of the stainless steel was reduced and
Plexiglas was added between the core and the Boral in a step-
wise menner until all the steel had been removed. The changes
in reactivity, referred to the all beryllium reflector, which
were incurred by these reflector alteratlons are shown in Fig. 3k4.
An additional experimental point in this figure, designated as
2.9/16" Plexiglas and 2-7/8" stainless steel shows the effect of
removing the Boral from a Plexiglas-stainless steel composite.
Another point, measured with the Boral replaced by 5/16" of Plexi-
glas shows the effect of a 5/16" vold between the Plexlglas and
the steel to be small. The effect of the removal of all materials
from the test section is also shown on the graph. ‘

Stéinless Steel, Plexlglas and Cadmium;Composite Reflector

The experiment described above was repeated with a layer
of cadmium substituted for the Boral. Since the cadmium sheets,
which were 0.02" thick and sandwiched between 5-mil aluminum,
were thinner than the Boral, the Plexiglas-steel composite was
correspondingly thicker. Changes in reactivity were measured
as the thickness of the Plexiglas, separating the core from the
cadmium shéets, was increased from zero to 2-3/16". As in the
preceding experiment, the total reflector thickness remained
constent. .The results are shown graphically-in Fig. 35, where,
for reference, the zero reactlvity change 1s again taken as that
with the normal beryllium reflector.

Neutron Flux Distribution through Stainless Steel, Plexiglas
and Boral Composite Reflector .

Bare and cadmium covered indium traverses were made through
this composite reflector starting at a point G" inside the core

Leverett, M. C. "The Direct Cycle Aircraft Reactor", Reactor Science
and Technology, 3, 7 (1953).

The Boral was prepared from a mixture of 35 weight percent B)C and
aluminum. This mixture is held between two aluminum sheets; each about
0.04" thick, forming a 1/4" sandwich containing 250 mg of boron per square
centimeter. The Boresl etripe were wrapped in masgking tape to mlnimize

boron contamination.

-63-
REACTIVITY LOSS IN CENTS

DWG. 21364 Y

O I [ l \I [
BERYLLIUM REFLECTOR

 

50}— —
60— . . L
REFLECTOR COMPOSED OF
2-9/16" PLEXIGLAS AND
70—

2-7/8" STAINLESS STEEL |

2-7/8" PLEXIGLAS AND

5 '/2-7/8" STAINLESS STEEL

90—
100—

| {Op—

 

120— - | —

150 : —_—
REACTIVITY LOSS FOR

/_vom IN TEST SECTION

| | I I I I
60 0 | 2 3 4 5 6

PLEXIGLAS THICKNESS IN INCHES
FIGURE 34. REACTIVITY LOSSES VS. PLEXIGLAS THICKNESS
FOR STAINLESS STEEL-PLEXIGLAS-BORAL COMPOSITE REFLECTOR

 

 

 
REACTIVITY LOSS IN CENTS

— DWG. 21365

SECURITY INFORMATION 65

T | A |
BERYLLIUM REFLECTOR

715p— : —

 

2- 7/8" PLEXIGLAS AND

' 2-7/8" STAINLESS STEEL
80|— —

85— —_—
90— , ' —
S ' ‘ | —
| 00— —_
05— —_

50— -

/vom IN TEST CELL

0o I 2 3 4 5
PLEXIGLAS THICKNESS IN INCHES

FIGURE 35. REACTIVITY LOSS VS. PLEXIGLAS THICKNESS FOR
STAINLESS STEEL -PLEXIGLAS - CADMIUM -COMPOSITE REFLECTOF

 

 

 

 
and terminating at the bottom edge of the reflector. For
these traverses the composite reflector consisted of an out-
side layer of 2-7/8" of stainless steel separated from a
2-9/16" Plexiglas inner layer by the Boral sheet. Figure

36 shows the results of these traverses, in which the various
activities are ‘plotted as a function of the distance from the
bottom of the test section. Figure 37 shows the cadmium
fraction computed from the relation

An
Cadmium Fraction = B - AC
Ag

(where A refers to the activity induced in an indium foil
and the subscripts B and C indicate it to have.been bare

or cadmium covered, respectively) and plotted as a function
of the same distance. :

- Figure 38 shows the results obtained along the traverses
reported in Fig. 36, except that the Boral strips were omitted
and the space left void. Cadmium fractions were again computed
and are plotted in Fig, 39. No flux traverses notr cadmium
fraction determinations were made with the stainless steel,
Plexiglas and cadmium composite reflector.

 

F. Danger Coefficlent Type Evaluations in Rectangular Shell Type Asgembly:

1.

Poison Rods

The changes in reactivity caused by prototypes of various
control rods in design by General Electric were obtained from

- the positions of calibrated control rods-when the assembly was

critieal.

The solid strip of Plexiglas in the bottom of a test cell
was replaced by a plece through which a hole 5/8" in diameter
had been bored. The axis of the hole wae parallel to the long
axls of the Plexiglas strip and equally spaced between the top
and bottom faces. A hole was also bored through the graphite
end reflector to match the one in the Plexiglas. The poison rod
wvas inserted into the hole with one end of the rod flush with the
midplane. The reactor was then made critical. The rod was dis-
placed 18", the end of the rod being at the end reflector-core
interface, the solid plece of Plexiglas replaced and the system
again made critical.

Table XII gives a description of the rods for identification,

'the matrix cells in which the evaluation was made and the decreases

in reactivity, in cents, Incurred when the rodsg replaced the plastic.

" The three values recorded in Table XII and evaluated In cell I-12

were made after the conversion.of: one guadrant: of:erne. halfiof . the
assembly:fromthe rectdngular. shell. .type, graphite reflected, to: the
water-cone' reflec¢ted mock-upy .The.other four values were obtained
in the rectangular shell type assembly before the conversion.

~66-
RELATIVE ACTIVITY

DWG. 21366 Iy

 

 

    

 
 

 

 

   
 

 

 

" L 67
' | | | | |
0.9— —
BARE INDIUM ACTIVATION
0.8}— -
0.7 —
0.6— CADMIUM COVERED N ]
INDIUM ACTIVATION
0.5}— —
N
0.4}— —
BORAL
0.3p— —
BARE INDIUM ACTIVATION
MINUS CADMIUM COVERED
INDIUM ACTIVATION
0.2}— —_
STAINLESS STEEL
PLEXIGLAS
0.1 _
CORE ———»
0 | | |
15 12 9 6 3 0

DISTANCE FROM BOTTOM OF TEST CELL M-22 IN INCHES
INDIUM TRAVERSES

FIGURE 36. BARE AND CADMIUM COVERED

THROUGH STAINLESS STEEL- PLEXIGLAS - BORAL COMPOSITE REFLECTOR
CADMIUM FRACTION

0.8

0.6

0.4

0.2

DWG. 21367 BN ss

 

 

I I I I I I

BORAL

    
      

STAINLESS STEE

PLEXIGLAS:

CORE —

I I I | S |

 

 

15 12 9 6 3 0
DISTANCE FROM BOTTOM OF TEST CELL M-22 IN INCHES
FIGURE 37. CADMIUM FRACTION VS. DISTANCE FROM BOTTOM
OF TEST CELL FOR STAINLESS STEEL -PLEXIGLAS - BORAL

COMPOSITE REFLECTOR

 
RELATIVE ACTIVITY

 

 

     
 

 

     
 
   

 

 

 

 

 

 

1.0 - :
I l | o | |
0.9 BARE INDIUM ACTIVATION .
0.8{— \ —
0.7}— —
0.6}— \ ]
\
CALMIUM COVERED I
INDIUM ACTIVATION I
|
0.5— I _
I
|
|
04— { —
‘ VOID
0.3}— S ———— \ ]
BARE INDIUM ACTIVATION \
MINUS CADMIUM COVERED \
INDIUM ACTIVATION \
| STAINLESS  __|
0.2 STEEL
' PLEXIGLAS
g
O.l}— \ E —
- g
CORE —= N
\
N
I | I I \
15 12 9 6 3 0

DISTANCE FROM BOTTOM OF TEST CELL M-22 IN INCHES

FIGURE 38. BARE AND CADMIUM COVERED INDIUM TRAVERSES
THROUGH STAINLESS STEEL- PLEXIGLAS COMPOSITE REFLECTOR
CADMIUM FRACTION

1.0

0.8

0.6

04

0.2

DWG. 21369 _

 

l l I | | [

VOID

   
 

STAINLESS STEEL

PLEXIGLAS-\-

CORE —

  

 

   

l I I ] R I

 

 

 

15 12 9 6 3 0

DISTANCE FROM BOTTOM OF TEST CELL M-22 IN INCHES
FIGURE 395. CADMIUM FRACTION VS. DISTANCE FROM BOTTOM OF
TEST CELL FOR STAINLESS STEEL-PLEXIGLAS COMPOSITE REFLECTOR

74
TABLE XII
POISON ROD REACTIVITY CHANGES

Reactivity Loss

 

 

Rod Description Cell Number in Cents
GE Rod #1* I-12 ‘18.5
Rod Labelled "Silver" I-12 12.9
Rod Containing 85.6% Ag I-12 18.2
and lh 4% Boron
GE Rod #1 - M-8 17.5
" GE Rod #1* M-11 19.9
Rod Detailed in GE Print M-11 23.2
#B4098083-55"
Cadmium Rod Containing .8.56 M-11" 13.2
grams Cd/cm., 34-1/2" long
and 1/2" 0.D.

2. Comparison of Molybdenum and Stainless Steel

The stainleas steel sheets in three adjacent cells in one
half of the core were replaced by ones of molybdenum having the
game dimensions. The resulting change in reactivity was de-
terminéd, in terms of previously calibrated contrel rod settings,
by bringing the reactor to criticality before and after the molyb-
denum sheets were introduced. The experiment was repeated in
three other positions in the reactor, using three adjacent cells
each tims, Table XIII gives the group of cells used and the corres-
ponding gains in reactivity resulting from the substitution of the
molybdenun for stainless steel.

TABLE XTIIT

COMPARISON OF MOLYBDENUM AND STAINLESS STEEL

 

Cells Used Gain in Reactivity in Cents
L-12, M-11, N.12 | 11.9
L-9, M-9, N-9 Te9
L-7, M-7, -7 - 5.6
L-k, M-k, N-L 1.9

3. Comparison of Plexiglas and Graphite

Removal of the Plexiglas from an array of nine matrix
cells, K through M-16, 17 and 18, in one half of the assembly
incurred a loss in reactivity of h? 8 cents. When graphite was
substituted for Plexiglas in these same nine cells the loss
amounted to only 26.7 cents.

* Specifications for: this'rod are given:in Appendix B. - .

~71-
L, Effect of Arco Soil Sample on Reactivity

 

The loss in.reactivity incurred when a small sample of Arco,
Idaho, s0o1l® was introduced into the reactor was determined by
comparing the reactivity of the assembly when the soll sample
was placed in a stainless steel box 1" x 2-7/8" x 2.7/8", having
a wall thickness of 8 mils, located at the mid-plane of the reactor
core in cell M-12, to that of a void in a box of equal volume at the
same location. The results show that with 16l.L4 grams of soil in
the box there was a loss of reactivity of 0.83 cents. -

G. Power Digtribution and Neutron Flux Detérminations:in Rectangular :
Shell Type ASsomblys

 

Axial and radial power traverses were made through the rectangular
shell type assembly employing the aluminum catcher foil technique which
has already been described. Since no fuel disks were located at the
center of the core, the axial traverses were extrapolated to the mid-
plane for normalization. Relative power wae then plotted as a function
of the distance from the core midplane. The one radial power dilstri-
bution which 1s reported was taken with all catcher foils 1.8" from
the midplane.

Neutron flux distributions in the rectangular shell type assembly
were determined using indium foils. In each case the bare indium
activation the cadmium covered .indium activation and thelr differ-
ence are shown on each graph, plotted as a function of the appropriate
distance.

1. Axlal Power Traverses

 

Axial power traverses were made through cells M-12, M-8, M-6
and M-L. The small variations in relative power through cells
M-12, M-8 and M-6, do not Justify plotting separate curves for
each of these cells. There.is, however, an appreclable decrease
through cell M-4 and a separate curve is plotted. A summary of
the results 1s given in Fig. 40 and Table XIV.

TABIE XTIV

REIATIVE AXTAL POWER

 

Distance from Midplane In Inches

 

Cell Number . 0.0 . LB .. 5.k .. 9.0 . 12.6 ... 16.2
M-12 1.00 0.99 0.99 0.87 0.84 0.71
M-12 1.00 0.99 0.98 0.88 0.81 0.69
M-8 1.00 0.99 0.97 0.88 0.82 0.73
M-6 1.00 0.99 0.68 0.87 0.82 0.72
M-k 1.00 0.99 ~ 0.94 0.83 - 0,75 0.62
M-4 ~1.00 0.99 0.93 0.88 0.7k 0.65

 

* An analysis of this soll sample is given in Appendix B.

:-72-
RELATIVE ACTIVITY

DWG. 21370 _

 

I | T I 7 I I | I

M—12

O
@
A
A

 

0.4}—

o.2t—

 

0 I I I I I I | I I

 

 

2 4 6 8 10 12 14 16 18
DISTANCE FROM MIDPLANE IN INCHES
FIGURE 40. AXIAL POWER TRAVERSES THROUGH CELLS M-{2, M-8, M-6
AND M-4
Radial Power Traverse

 

One vertical radial power traverse was made 1.8" from
and parallel to the midplane starting at the axls of the
reactor and extending through the fuel plane of cell M=k,
The results of this traverse are shown graphically in Fig. kl,
The large decrease in relative power which occurs along the filrst

few inches of thls traverse 1s due to cell M-12 being surrounded

on all four sides by & one inch layer of Plexiglas which forms
the innermost: v of the series of concentric shells making up the
agsgembly.

The data represented by the last two points on this curve
were obtained from aluminum catcher foils placed in contact
with the uranium fuel disk in cell M-4, one foil being adjacent to
the top .and onecionithe bottoin of this fuel.dlsk. - The curve is. drawn
through points repréesenting data obtained from the upper surfaces
of the fuel disks.

Factors Affecting Power Production by Fuel Disks
a. Self Shielding

If the fissionable material in a reactor is of high
density the neutron flux producing fissions is depressed,
a phenomenon known as fuel self shielding. The self shield-
ing of a 0.01" fuel disk was measured before the converslon
from the shelf type to the rectangular shell type mock-up.
by forming a sandwich composed of five 2-mil fuel disks
separated by aluminum catcher folls. The results are sum-
marized in Fig. L2 where the experimental points are the
averages of catcher foil counting rates which varied less
than 1%. The distribution is assumed to be symmetrical
about the central disk and the experimental points have been
reflected about the plane of symmetry. The ratio of the aver-
age activity to that at the surface of the 10-mil disk is 88%.

b, Cadmium Fraction of Fuel Disk

The cadmium fraction for a fuel disk 1s defined as the
fraction of fissions which are produced by neutrons having
energles below that of the cadmium cut-off. It is obtained
by the relation:

Cadmium Fraction = 1BC
Fp
where F refers to the fission fragment actlvity on the
aluminum catcher foll and the subscripts B and C indicate
the uranium disk as having been bare or covered with cadmium.
Cadmium fractions for fuel disks were obtalned both before
and after the conversion from the shelf to rectangular shell
type mock-up from aluminum catcher foll measurements. All
cadmium fractions for fuel disks were mesasured at a distance of
1.8" from the midplane of the assembly. The results are sum-
marized in Table XV.
. The
RELATIVE ACTIVITY

0.2

DWG. 21371 [

 

 

I I I l | |

 

 

 

I l | 1 l l I l
6 9 12 15 I8 21! 24 27
VERTICAL DISTANCE FROM AXIS OF REACTOR IN INCHES
FIGURE 4I. VERTICAL RADIAL POWER TRAVERSE AT MIDPLANE

6L
6.

TABLE XV

éADMIUM.FRACTIONS OF FUEL DISKS

 

 

 

 

 

 

Rectangular
Shelf Type Assembly Shell Type Assembly
Cell Cadmium Fraction Cell ‘Cadmium Fraction
M-12 0.92 M-12 0.94
U-12 0.92 M-8 0.92
M-L4 0.90 M- 0.90

The plastic encloging Cell M-12 in the RS mock-up appears
to increase the fraction of thermal fissions there.

Absolute Power Determination

Absolute power in the rectangular shell type assembly was
determined by normalizing the power traverses to absolute fission
rates by comparing & one-mil gold foll activation in the critlcal
aggembly with that induced by the Oak Ridge National Iaboratory
Standard Pile. From this normalization the absolute power during
foll exposures was found to be approximately two watts.

Radial Flux Distributions
a. Traverses not in Plexiglas

Bare and cadmium covered indium activatlions were measured
along a vertical radius starting at the axis of the reactor in.
Cell M-12 and extending downward through the beryllium reflector,
terminating at the bottom of cell M.22. Data for these traverses
were taken with the indium folls located as near the mid-plane
as poseible and not in the Plexiglas. Results are shown graph-
ically in Fig. u43.

These curves indicate that the flux in cell M-12 1is high.
This flux distribution agrees with the peak found in the relative
radial power distribution shown in Fig. 41. The flux peak in
the beryllium reflector also corresponds to that found for the
shelf type assembly as shown in Fig, 21.

b. Traverses Through Plexiglas

Bare and cadmium covered indium traverses were made through
the Plexiglas starting at the axls of the reactor in cell M-12
and extending through the beryllium reflector along a horizontal
radius. The results are shown graphically in Fig. L4. The rela-
tively high values of the flux in cell M-12, which 1s surrounded
on four sides by one inch thick layers of Plexiglas, and the peak
in the beryllium reflector, which have been noted previously, are

again observed.

Axial Flux Distribution

Bare and cadmium covered indium traverses were made through cell
M-12 along the axls of the reactor, starting at the midplane and

=76-"
DWG. 21372 7

1,00

O Experimental Points

D Reflected Points
0095

0.88

Relative Activity
o
0
o

Q
.

(@]
AVa ¥

0.80

 

0 2 L 6 8 10
Uraniuwm Thickness, Mils

FIGURE 42. POWER GENERATION THROUGH FUEL
RELATIVE ACTIVITY

DWG. 21373

 

  
 
   
    
 
  
   
   

-0
I

0.8}—

0.6p—

X MINUS CADMIUM COVERED
*\\/—— INDIUM ACTIVATION

BARE INDIUM ACTIVATION

O

 

BARE [INDIUM ACTIVATION

CADMIUM COVERED

 

 

 

 

™~
_ J*BE REFLECTOR

 

 

| fis_ o INDIUM ACTIVATION
04— ~
0.2}—
NN
0 | | | | I I N >SS S JJ
4 8 12 16 20 24 28 32

DISTANCE FROM AXIS OF REACTOR IN INCHES

FIGURE 43. VERTICAL BARE AND CADMIUM COVERED INDIUM TRAVERSE AT MIDPLANE

FROM CELL M-12 THROUGH M-22

8L
0.8

 

    

RELATIVE ACTIVITY

  
   

 

 

 

 
    
 

 

 

 

DWG. 21374 Oy
| | M~ NI~ 1
~ N>
NN DY ~N
~ > N>
- ~ NN
BARE INDIUM ACTIVATION ~ ™~ Y
<> h
~ N
~
- \ ~ ~N ™ _
~ O e
< D | REFLECTOR
™~ ~ > ~
~
~ ~ q
\ \ \ N
e~ BARE INDIUM ACTIVATION Y \\\
“w¢———MINUS CADMIUM COVERED ~ NN
N\ INDIUM ACTIVATION ~ N ~
N ~N
M AN
- ~ ~ ~ —
~ ~N
~—_ SO~
CADMIUM COVERED T~ ~ ~
/_INDIUM ACTIVATION =~ ~ N~
n ~w ~
O ~ ~
\\
~N
N
“~
porte— \ e
\ ~
\ NG 0
NN
| | | 1 | l IS
0 4 8 12 16 20 24 28 32 36

DISTANCE FROM AXIS OF REACTOR IN INCHES
FIGURE 44. RADIAL BARE AND CADMIUM COVERED INDIUM TRAVERSES TAKEN HORIZONTALLY
| AT MIDPLANE THROUGH PLEXIGLAS
extending through the graphite end reflector. The results
are shown graphically in Fig. u45.

7. Neutron Flux Dlstribution at Plexiglas Intersection

 

The conversion from the shelf type to the rectangular
shell type of mock-up produced intersectlons of the ver-
tical and horizontal layers of the Plexiglas strips ex-
tending radially from the four corners of cell M-12 at
azimuthal angles of about 45° both above and below the
horizontal. The obJect of the measurements noted here
is to study the flux distributlon in the vicinity of one
of these intersections. The region selected 1s 1llustrated
in Fig. 46. The vertical colums of the matrix are indicated
by the notation at the bottom of the figure and the row se-
lected 1s noted on the right of the figure. The arrangement
of the Plexiglas may be clearly seen from the loading dlagram,
Fig. 5.

Bare and cadmium covered indium traverses were measured
along a horizontal in the midplane with the indium folls
occupying the positions indicated by the small triangles
in Fig. 46. This figure also shows graphically the results
of the traverses.

 

8. Unit Cell Traverses

Bare and cadmium covered indium traverses were made
vertically across cell M-12 and extending through the one
inch layers of Plexiglas adjacent to the top and bottom of
this cell., This central cell contains stainless steel and
uranium fuel disks but no Plexiglas. It 1is, however, sur-
rounded by a one inch layer of Plexiglas on all four sides.
The results of these traverses are shown graphically in Fig. L7,

G, ' Neutron Flux Distribution Around a Poison Rod

Two flux distributions,'each mitually perpendicular to
the axis of one of the experimental poison rods supplied by

the General Electric Company, and designated as Rod.#l*y were
determined. One traverse was in the plane of the Plexiglas

layer and the other perpendicular to 1it.

The poison rod occupied the position in cell M-8 which
had been used during the reactivity evaluations and which
has been previously described in this report. This position
is 1llustrated in Fig. 49. The foil positions for the hori-
zontal and vertical traverses are illustrated in Figs. 48 and
49 respectively, and the diagonally cross-hatched lines in °
these two filgures represent sections through the Plexiglas.

The horizontal flux distribution is represented graph-
ically in Fig. 48 and the vertical distribution in Fig. 49.

All the flux distribution curves taken in the vicinity
of the poison rod show a discontinulty at the surface of the
rod since it was not possible to make neutron flux determin-
ations inside the rod. In Fig.. 48 those foils whose distances

¥ The specifications for this rod are glven in Appendix B.

=80~
RELATIVE ACTIVITY

0.6

0.5

0.4

0.3

0.2

0.1

 

 

 

 

 

 

DISTANCE FROM MIDPLANE IN INCHES

FIGURE 45. BARE AND CADMIUM COVERED INDIUM TRAVERSE ALONG AXIS OF THE
REACTOR THROUGH CELL M-12

DWG. 21375 -
T T <~
| | | | ~ J\ S N | | |
NN N NN
S NN N
\\ \ ™~
- ~ NN
~ SN
~ 0 ~_ \[~ GRAPHITE REFLECTOR
SO O
O ~
— N OO O
N SO N Y
U ~ N ~ N
e ™~ ~ ~ ™~
~N \\ ~
— ~
o ~ N
~
@ NG ~ ~
o OIS
~
BARE INDIUM ACTIVATION ~ WU
— MINUS CADMIUM COVERED NN Yo
/ INDIUM ACTIVATION N N LN
Q \\ ~
T -~ - , 0 ~
= - -—.__.-_. ™~ ~
O o "‘\\ > ~N d
J \\\
) ~
CADMIUM COVERED S BARE INDIUM ACTIVATION
INDIUM ACTIVATION I (H
e . U
~
\ \ 9
N ™~ ™~ @ Q
~N 0
I I _ l I ™~ \1\\ ™~ \ 1 I I
4 8 12 16 20 24 28 32 36

18
RELATIVE ACTIVITY

DWG. 21376

 

BARE

INDIUM ACTIVATION

 

 

 

 

 

 

 

 

A FOIL POSITIONS

 

 

 

 

 

 

 

 

 

 

 

 

 

BARE INDIUM ACTIVATION f\
MINUS CADMIUM COVERED —O——0—
INDIUM ACTIVATION
——————————— - O ROW 9
0.3} — = T ~\ O —
lA”'Z/A A A A A alla alala afa l
_ {
| |
l /0
. P
0.2f=© O O o —4ol,_ __ o ! / ]
CADMIUM COVERED _0_0_’/0~
INDIUM ACTIVATION L
LNM _‘-—-—J
O.lp— —_—
t— CELL N —>t®— CELL O -—>=but— CELL P —3}je— CELL Q —%4<— CELL R —»j«— CELL S —j
I I . | L] | ]
2 3 | 2 3 | 2 3 3

0o I 2 3 I 2 3 I

CELL DIMENSIONS IN INCHES
FIGURE 46. HORIZONTAL BARE AND CADMIUM COVERED
PLEXIGLAS

INDIUM TRAVERSES AT

INTERSECTION. TAKEN AT MIDPLANE

é8
RELATIVE ACTIVITY

0.6

0.5

04

0.3

0.2

0.1

DWG, 21377 — 83

 

, Eg INDIUM ACTIVATION
___™

 

| I I I I I

BARE INDIUM ACTIVATION

BARE INDIUM ACTIVATION
- MINUS CADMIUM COVERED  _
/7 \/ INDIUM ACTIVATION // \
/ \

T e . m— T —— — T — — —— — o—

| CADMIUM COVERED

 

PPLEXIGLAS%STAINLESS STEEL AND FUEL—*}-PLEXIGLAS-l

 

 

l I | I I 1
2 3 4 5 6 7

DISTANCE FROM BOTTOM OF CELL M-i3 IN INCHES
FIGURE 47 VERTICAL BARE AND CADMIUM COVERED INDIUM
TRAVERSES ACROSS CELL M-12 AT MIDPLANE
RELATIVE ACTIVITY

0.5

o
D

o
w

0.2

0.1

DWG., 21378 -

 

 

| I [ I | I I

 

AIIRIITIRRIRZRIRIRINRINRNNY

hyp

    
   
 

O

 

I

BARE INDIUM ACTIVATION

BARE INDIUM ACTIVATION

MINUS CADMIUM COVERED
/ INDIUM ACTIVATION
/
/ / i CADMIUM COVERED
/ INDIUM ACTIVATION

 

 

IN REGION OF POISON ROD # |. ROD LOCATED IN CELL M-8

/
- / —
/
I | I [ | I I | I
8 6 4 2 0 2 4 6 8
DISTANCE FROM AXIS OF ROD IN iNCHES
FIGURE 48. BARE AND CADMIUM COVERED INDIUM TRAVERSES THROUGH PLEXIGLAS

g
DWG. 21379 Ay

 

 

1 l [ | [

 

 

NN

 

BARE INDIUM ACTIVATION
MINUS CADMIUM COVERED
INDIUM ACTIVATION

%\\\\\\\\\\ P S

0N
- /
n / {
/ CADMIUM COVERED

M-8 , — ,’ ?j* INDIUM ACTIVATION
|
)I
O\
\

 

 

 

g

 

e (N

N NN Z

1 I S $ 0 P

——— M - § —

BARE INDIUM
ACTIVATION

 

 

h-(—-|———b-l:——
A

 

 

 

l I I | I |

 

 

0 0.1 0.2 0.3 0.4 0.5
RELATIVE ACTIVITY

FIGURE 49. VERTICAL BARE AND CADMIUM COVERED INDIUM TRAVERSES IN
MIDPLANE THROUGH POISON ROD #1|. ROD LOCATED IN CELL M-8

¢8
from the axis are indicated as zero were actually attached
to the circumference of the rod by a very narrow strip
of tape.

V. EXPERIMENTAL STUDIES ON WATER-CONE REFLECTED ASSEMBLY

 

A, Traversese in The Water-Cone Reflected Assembly:

 

Bare and cadmium covered indium neutron flux traverses were
made through cells R-12 and U-12 parallel to the axis, in the
quadrant of the reactor in which the reflector had been altered
to represent a mock-up of the water-cone reflected assembly. The
traverses began at the mid-plane and extended through the simu-
lated water cone. Both the bare and the cadmium covered folls
were located in the Plexiglas strips one-half inch from the bottom
of the cells. The results of these traverses are shown graphically
in Figes. 50 and 51. '

Bare and cadmium covered indium traverses were made through
cell H-12, in a quadrant of the assembly in which the reflector
was unaltered, for comparison with those of R-12 and U-12. Cell
HE-12 is the same horizontal radial distance from the central rec-
tangular shell cell, M-12, as is R-12, Alsc safety rod No. 8
occuples the same relative position with respect to cells H-12
and M-12 as No. 7 occuples relative to celle R-12 and M-12. The
bare and cadmium covered folls were again located in the Plexiglas
one-half inch from bottom of cell. The graphic results of these
traverses are shown in Fig. 52.

"No satisfactory explanation for the neutron flux peak observed
in the vicinity of the midplane in cell H-12 is immediately available.
Most other axlal flux curves taken through this assembly follow the
general shape of the zero order Bessel function of the first kind,
especially near the midplane. The dotted portion of the bare indlum
activation curve shows this relation.

-86-
RELATIVE ACTIVITY

DWG. 21380 Ay

 

0.3

0.2—

O.4p—

 

| ! I | | I I I I

 
  
 
  
   
    

BARE INDIUM ACTIVATION

—— END OF STAINLESS STEEL AND FUEL

" BARE INDIUM ACTIVATION
MINUS CADMIUM COVERED
INDIUM ACTIVATION

        

CADMIUM COVERED
INDIUM ACTIVATION

/END OF PLEXIG|LAS EXTENSION

pt—— WATER CONE — 3

 

 

 

 

5 10 15 20 25 30 35 a0 45
DISTANCE FROM MIDPLANE THROQUGH CELL U-IZ2 IN INCHES

FIGURE 50. BARE AND CADMIUM COVERED INDIUM TRAVERSES THROUGH CELL U-I2

   

 

50

8
RELATIVE ACTIVITY

0.6}—
0.5|—
BARE INDIUM ACTIVATION
. D)
0.4
<—— END OF STAINLESS STEEL AND FUEL
J
0.3 BARE INDIUM ACTIVATION
~ Q MINUS CADMIUM COVERED
INDIUM ACTIVATION
—-—.___\
S
™~ o~ |
0.2 ™ ——END OF PLE |XIGLAS EXTENSION
. \\\ /‘-
Q N 4 S 0 f<—— WATER CONE ——»=
@ N\
01— >
' CADMIUM COVERED A o
INDIUM ACTIVATION e J
A e . __7 S
U
J
I | I I | 1 = O====0
0 5 10 15 20 25 30 35 40 45
DISTANCE FROM MIDPLANE THROUGH CELL R-12 IN INCHES
FIGURE S5I. BARE AND CADMIUM COVERED INDIUM TRAVERSES THROUGH CELL R-12

DWG. 21381

 

 

 

| l I I

 

 

 

 

 

 

50
 

    
 
 
 
   

RELATIVE ACTIVITY

DWG. 21382 Ay
| | l | | N NN |
\ N\ NN\
NN YN
. 2p— \ AN
NN \\ N
\ \
N N
NN
N N\
S8ARE INDIUM ACTIVATION \ N N\
GRAPHITE
- \ \\ \\\ ““REFLECTOR
BARE INDIUM ACTIVATION
0.6 p— MINUS CADMIUM COVERED
— / INDIUM ACTIVATION
04 T
CADMIUM COVERED
INDIUM ACTIVATION
0.2}—
I 1 l I | |
0 3 6 9 12 15 27

 

 
  

 

  
  
  
  

 

 

 

DISTANCE FROM MIDPLANE THROUGH CELL H-I12

IN INCHES

FIGURE 52. BARE AND CADMIUM COVERED INDIUM TRAVERSES THROUGH CELL H-12

68
VI CONCLUSIONS

Results of the Critical Experiments on the various mock-ups of the
Direct Cycle Aircraft Reactor indicate satisfactory agreement between ‘the
theoretical and experimental values desplite the complexity of the assembly.9
The agreement obtalned between the measured and computed reactivity losses
due to separation of the two halves of the core at the midplane, based on a
homogeneous thermal reactor, gives added confidence to the application of
methods of reactor calculations to complicated assemblies.

The simllarity of the results obtained for power distributions from
catcher foll activities and the counting of fission product radiations
directly from the fuel disks indicates that this method of measuring fission
rates and for normalizing indium foll exposures between different runs is
satisfactory for the neutron spectrum studied here. Thle eliminates the
necesslity for counting the fuel element back-ground prior to exposure.

The fair agreement between the results obtained for rod calibrations by
the stable period and the rod-drop methods indicates that the latter may be
used with confidence. Since the rod-drop method is much less laborious,
it may be used to check the integrated value of control rods and to quickly
evaluate safety rods.

Some effects observed during the course of the critical experiments
have not been thoroughly studied. Among these are the peak observed on
the control rod sensitivity curve; the peak 1n the reactivity vs Plexiglas
thickness curve; the increase in reactivity with an Ilncrease in temperature
of the assembly. Some of the changes in reactivity which accompany alter-
ations of the reflector in different localities have not been explained.

 

9 1Ibid. Table 8, p. 70. .

G0 -
APPENDIX A
By E. L. Zimmerman
CONVERSION FROM SHELF TO RECTANGUIAR SHELL MOCK-UP

During the early stages of the critical experiments on the Direct
Cycle Aircraft Reactor, considerable controversy arose over the proper
method of including neutron streaming in the calculation of age. One of
the purposes of the comparison of the shelf and rectangular shell con-
figurations was to estimate this streaming effect. Whatever the magni-
tude of the streaming effect, it was anticipated that reducing the radial
straight line void distences would .result in an Increase in multipllication.
Contrary to the expected effect, converting from the shelf to the rectangu-
lar shell configuration caused a substantlal negative change in multlpli-
cation.

While the  original question was not answered, a partlal explanation
for the negative change is given on the basis of a net shift of fuel away
from the region of greatest importance in the assembly.

To affect the change from the shelf to the rectangular shell type
assembly, approximately half the cells were rotated through 90° in such
a way that, in these cells, the fuel was moved 1/2" or 1.27 cm away from
vertical longitudinal midplane of the assembly, Call-this dlsplacement a.
The component of the displacement a along a radius at an angle © 1s given
by a cos 83 b represents the distance of a particular cell from the axils.

Averaging the radial displacement over interval 0= © < 17

 

 

 

 

 

T
/’Tr/h
gives - o P CZB: a8 | .
cos8” ‘(-—" X
, 1 - L -
' bde :
. Y2
o "cos2 © |

Approximately half of the elements were rotated, 80 the overall average:dis-
placement amounted to
A - as 0.56 cm |

2
For a bare reactor of radius R having the aéymptotic flux distribution of
the actual assembly, the radial part of the flux is given by ¢ (r) = J5(By ),
where B, = 2.405 , .1s the radial bucking component. Since, in the asymptotic

R

approximation, the flux is self ajoint, the lmportance function is [¢ (fl]e.
The change in miltiplication due to increasing the fuel denslty by a small .
amount &m at v is c &m @2 (r). Integrating over the entire radius one has
R _
Skzzfcéhjg ¢2(flrdf . (2)
This change in multiplication may also be found by adding the perturbation

-91-
(l*'éé) to the fuél cross sections in the critical equation:
‘ m

Lrox VB 148 pin TE
ST o(14 G2 Y4 o DB

m
Wherqzza refers to the thermal absoprtion cross-section in materials
other than the fuel and) Y is the thermal absorption cross-section in

the fuel. Solving (3) for 6k and neglecting second order term glves
|

Korp = (3)

 

. m 2 - ‘
DB '
8}{ = Za+ §I_D._ - 0.)4'7)4_ éE o (’-l-)
Shonw |

The term in the brackets is evaluated from the unperturbed reactor con-
stants.

 

The change in multiplication caused by shifting an element of fuel
a distance = 18

ag

C[¢2 (r+d) -¢2(’f'):' m=cmn2/ ¢ (r) drly  (5)

Integrating over the entire radius and substituting the value of ¢ found
by comparing equations (2) and (4) gives:

’ R . troN g,
b = 2 |t D" -le”(*““")*d* (e

Sathat ™ | [y Forirar

Substituting J, (Bg? for ¢ (r), the values from (L),

and By = 0,0355 cm™t

J
glves the result
§x = - 0.00938, ;

or & loss in reactlvity of about 130 cents

;92;
APPENDIX B
ANALYSES OF REACTOR MATERIALS
1, Arco Soll Sample

Principal Constituents
Silicon, Aluminum, and Iron

Present 1n Spaller Quantities
Sodium, Calcium, and Potassium

Quantitatively Measured

Boron 0.05%
Lithium 0.19%
Magnesium 0.02%
Maganese 0.05%

Elements not Detected
Cadmium and Copper

2. General Electric Company Experimental Poison Rod #1.

Volume of By,C as Packed 50.77 cmd
*Length of Rod Packed with ByC 28.8 inches
Weight of B)C . .. . 81.0 grams
Packed Density of B) 1.595 grams/cm3
*¥Boron per cc 1.248 grams
Length of Lavite Plug 3/8 inch

*¥Ineslde Measurement
**Agsuming Pure ByC

3. General Electric Comcany Experimental Poison Rod Detalled in GE Print
No. B 4098083 - 55.

This poison rod consisted of a cadmium portion 20" long whose
crogs-sectional area was a Maltese cross, having arms 0.082"
thick and a total length of 3/4", formed from O.04L1l" cadmium
sheet stock. This cadmium rod was inserted into an aluminum
tube 30" long, 0.835" inside diameter and 0.083" wall thickness.
The cadmium cross was held in position by a sultable plug.

4, Stainless Steel, Type 302

Maganese 1.11%
Chromium 17.3
Nickel 7.6
Boron 0.001

Note: The 7.6 percent of nickel 1s below the specifications for type
302 stainless steel. This value was determined six times by
three methods.

-93-
 

 

 

Spectrographic Analysis of Stalnless Steel Type 310

5.
Ag < 0.04% Cd < 0.04% Hf < 0.5% Ni - 10.0%4 Ti < 0.005%
Al < 0.04%  Co 0.63 In < 0.15 Pb < 0.08 v 0.08
B < 0.004 Cr = 10.0 Mg< 0.02T Si 0.63 W <« 1.3
Ba < 0,02 . Cu- 0.15 Mn ~ 0.63 Sn < O0,04T Zn < 0.31
Be <« 0.0003 . Fe =10, Mo 0.31 Sr < 0.06 Zr < 0,15
Ca < 0.80 Ga < 0,04 Na <« 0.63 Ta < 1.3
6. Analysis of 28 Aluminum Used as Catcher Foils
QUANTITIES PRESENT IN PERCENT
Ag< 0.04 Co< 0.08 Mn - 0.04 - Sn <0.04
Al7 5.0 Cr 0.08. Mo« 0.08 Sr< 0.1
B « 0.01 Cu 0.0k Na <« 0.01 Ti< 004 T
Ba<=0.04 Fe 0.3  Ni=< 0.08 V < 0.08
Be =< 0.001 K. 0.02 Pb< 0.08 Zn< 0,3
Ca < 0.08. Mg< 0.02T Si. 0.3 " Zr«< 0.15
7. Analysis of Square Aluminum Tubing Used in Matrix
QUANTITIES PRESENT IN PERCENT
Ag< 0.04 Co<0.08 Mn 0.0k Sn < 0.0h4 °
AlZ 5.0 Cr 0.08 Mo=0.08 Sr<0.1
B< 0.0l Cu -0.15 Na<=0.01 Ti<€0.04T
Bg < 0,04 Fe 0.6 Ni< 0.08 V =< 0.08
Be< 0.00l K < 0.02 Pb< 0.08 Zn< 0.3 -
Ca<= 0.08 Mg 0.02. S1 0.2 Zr=< 0,15,
8. Analysis of Plexiglas Used as Moderator
Carbon 58.7% :
Hydrogen 8.15% F;
Ignited Oxides 0.0053%
Spectrographic Analysis of Ignited Oxldes
QUANTITIES PRESENT IN PERCENT
Ag < 0.04 Co. € 0.08 Mo £ 0.08. Sr< 0.1
Al 0.2 cr 0.15 .Na< 1.0. Ti 0.2
B £ 0.01. Cu 0.15. Ni 0.3 V'j( 0.08
Ba <& 0.0h4 Fe» 0.5 P <0.08 2Zn< 0.3
Be 0.001 Mg 0.2 Si 2.5 Zr< 0.15
Ca 0.08 Mn 0.3 Sn < 0.08‘ T

9. Chemical Analysis of Stainless Steel Type 310

Fe
Ni
Cr

52

.8%

19,1%

25

0%
';9#;
APPENDIX C

SUMMARY of MATERIALS in REACTCR ASSEMBLIES

 

A. BSBhelf Type Assembly

 

 

 

 

 

 

 

 

 

 

 

 

 

 

1. Core
Density Mass Mass Volume | Volume™
Materlal gm/cc kg Fraction liters | Fraction
Urenium (total) 18.7 b2 0.038 2.2 | 0.002
U-235 -- (38.1 )% -- -
'Stainless Steel ' :

Type 302 7.87 384.5 0.357 48.9 0.040
Plexiglas 1.18 460.9 0.k26 391.5 0.320
Aluminum Tubes

Type 2S 2.72 193.0 0.179 70.5 0.058
Void -- - - 708.9 0.580

1079.6 1222.0
2. Reflector
Boryliium ~T1.86 1 1096.1 0.586 589.3 0.565
Graphite (AGOT) 1.72 957.8 0.425 556.8 | 0.440
Aluminum Tubes

Type 2S 2.72 200.0 0.089 73.5 0.058

Void -- - ‘ -— 46.9 0. 037
‘ 2253.9 _ 12865 : .

 

 

 

 

 

 

 

B. Rectangular Shell Type Asgembly

 

 

 

 

 

 

 

 

 

 

 

1. Core
Uranium (Total) 18.7 L1.07% 0.038 2.2 0.002
1U-235 - (38.3)** -- -- --
Stainless Steel ' . .
Type 302 CT7.87 383.4 0.357 48.6 0.040
| Plexiglas 1.18 4L59.2 0.426 389.8 0.319
Aluminum Tubes
Type 2S 2.72 192.9 0.179 70.6 0.058
Void -- -- - 710.2 0.581
1076.7 I25T.4
2. Reflector
Beryllium 1,86 1096.1 . 0.451 589.3 0.429
Graphite (AGOT) 1,72 1119.0 0.460 650.6 0.475
Aluminum Tubes
Type 2S 2.72 216.6 0.089 79.2 0.058
Void -- - - 52.0 0.038
PI31T 1371.1

 

 

 

 

 

 

 

 

¥  These values are averages over the entire core or reflector and differ slightly
from those given in the text, page 11, because of peripheral irregularities.

*% The mass of uranium reported here .1s the amount loaded in the matrix; the mass
of U-235 is that required to make the system critical with all control rods

completely inserted. 9%