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OAK RIDGE NATIONAL LABORATORY
OPERATED BY

CARBIDE AND CARBON CHEMICALS COMPANY

A DIVISION OF UNION CARBIDE AND CARBON CORPORATION

(=3

POST OFFICE BOX P
OAK RIDGE. TENNESSEE

 
L
»n

ORNL 1770

This document contains 89 pages
This is copy ¢ of 197 Series A

-

Subgject Category Reactors - Research
and Power

PRELIMINARY CRITICAL ASSEMBLIES
of the
REFLECTOR MODERATOR REACTCOR

Experimentation by D V P Williams
R C Keen (Louisiana State University)

J J Lynn
Dizxon Callihan

Reported by R M Spencer, USAF

DATE ISSUED
NOV 22 1958

PHYSICS DIVISION

A H Snell
Dairector

Contract Ko W-Th05, Eng 26
OAK RIDGE NATIONAL TLABCRATORY
Operated by
CARBIDE AND CARBON CHEMICALS COMPANY
A Division of Union Carbide and Carbon Corporation

Post Office Box P
Oak Ridge, Tennessee

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“fi. 11- ORNL 1770

1

103
10L4-106
107

Reactors-Research and Power
M-3679 (14th edition)

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ABSTRACT

Five preliminary critical assemblies, planned in conjunction with

a full scale aircraft type circulating fuel reactor, were constructed
at the Oak Ridge Critical Experiment Facility The first assembly

was constructed with a beryllium island and reflector, and had enriched
uranium disks placed between sodium filled stainless steel cans arranged
1n a fuel annulus The assenmbly was critical with 15 1 kg of U-235

The second assembly was also constructed with a beryllium i1sland and
reflector, however, a powdered mixture of sodium, zirconium and
uranium fluorides packed in aluminum tubes wes used as a fuel, and
fuel end ducts were provided The assembly was critical with 7 7 kg

of U-235 The addition of structural poisons (fifth assembly) to this
assenbly increased the critical mass to 18 2 kg of U-235 In the

third assembly the beryllium island and reflector were replaced by
graphite With 17 3 kg of U-235 in the fuel reglon, this assembly was
not critical In order to make the assembly critical approximately
80% of the graphite in the island and imner six inches of the reflector
were replaced by berylliuwn The fouwrth assembly was constructed as

a two region assembly with a 1 to 1 volume ratio of graphite and
powdered fuel in the core and a beryllium reflector This assembly
was critical wvaith 15 2 kg of U-235

Neutron flux distributions were determined in these assemblies
using indium foil activation Power distributions were determined
using aluminum catcher foils in contact with wranium fuel disks

Measurements were made on the effect od reactivity of placing various
materials in the fuel and reflector regions, and an estimate of the
neutron leaksge was obtalned for two of the assemblies
II

ITI

VIII

-vi-~

TABLE OF CONTENTS

ABSTRACT

LIST OF FIGURES

LIST OF TABLES .

INTRODUCTION

DESCRIPTION OF ASSEMBLY AND MATERIALS
CA-10

Assembly Loading

Control Rod Calibrations
Temperature Effects

Neutron Flux and Power Distribution
Danger Coefficient Measurements

=11

Assenmbly Loading

Evaluation of Control Rod D

Neutron Flux Distribution

Power Distributions

Neutron lLeakage Measurements

Evaluation of Boral Around Fuel End Ducts
Stainless Steel and Boron Poison Rods
Cadmium Importance Function

Danger Coefficients

Effect of Stainless Steel Around Fuel Annulus

GHHbOQEEBEOQE > g HOawr

CA-12

CA-13

A Assembly Loading

B RNeutron Flux Distribution
C Power Distributions

CA=14

A Critical Loadings

B Importance Function of Three Stainless Steel Rods
C Thermal Neutron Leskage

D Power Distribution
APPENDICES

A Summry of Materials in Reactor Assemblies
B Analyses of Reactor Materials

 

Page

vii

1ixX
-vii-

LIST OF FIGURES

Page
II-1  Photograph of Two Halves of Aluminum Matrix 3
III-1  Assembly Loading at Mid-plane >
ITI-2 Axie)l Loading of Assenbly in Vertical Plane Containing
Reactor Axis
ITI-3 Photograph of Typical Fuel Shish 7
III-% Calibration Curve for Control Rod D 9
III-5 Average Sensitivities of Control Rod D 10
IIT-6 Change in Reactivity vs Temperature 12
III-7 Bare and Cadmium Covered Indium Traverses Through Fuel
Core 13
III-8 Bare and Cadmium Covered lndium Traverses Parallel to
Axis Through Cell 0-12 14
ITI-9 Bare and Cadmium Covered Indium Traverses Parallel to
Axis Through Beryllium Island and Fuel 15
IIT-10 Indium Praverse Radially in Mid-plane of Reactor, Cell
0-12 to X-12 18
I1I-11 Bare and Cadmium Covere d Indium Traverses Across Fuel
Annulus 19
III-12 Power Distribution in Cell Q-13 21
II1-13 Power Distribution Awross Fuel Annulus 23
ITI-14  Power Distribution Through 20 Mil Uranium Disk 2k
III-15 Effect of Stainless Steel on Power Distribution 27
ITI-16 Toss in Reactivity Due to Addition of Stainless Steel
Core Shells 28
IV-1  Assembly Loading at Mid-plane 31

IV-2 Axial loading of Assembly in Vertical Plane Containing
Reactor Axis o o 32

IV-3  Fuel Tube Arrangement 33
V-5
V-6
Iv-7
-8
Iv-9
IV-10
Iv-11

Iv-12

vi-2
VI-3
VI-b
vI-5
VII-1

VII-2

VII-3
VII-4

VII-5

-viij-

Calibration Curve for Control Rod D

Radial Flux Distraibution at the Mid-plane o
Rad1al Flux Distribution 6-1/4" From Mid-plane
Cadmium Fractions in CA-10 and CA-ll

Position of Boral Inserted Around End Ducts

Axial Power Distributions in Call N-1k

Power Distribution Across Fuel in Mid-plane
Reactivity Loss due to Poison Rods Inserted in 0-12
Cadmium Importance Function

Assenbly Loading at Mid-plane

Assenbly Loading at Mid-plane

Neutron Flux Distribution in Mid-plane .
Assembly Loading at Mid-plane

Radisl FNeutron Flux Distribution in Mid-plane
Radial Power Distribution in Mid-plane .

Axial Power Distribution Through M-12

Radial Fuel Cadmium Fractions
Assembly Loading at Mid-plane

Axial Assembly Loading in Vertical Plane Containing
Reactor Axais ( v

Position of Nickel Sheets
Thermel Flux Leakage vs. Boral Thickness .

Power Distribution Across Fuel Annulus at Mid-plane

 

Page
35
36

. 37

39
40
41
43
46
b7
52
53
Sh
o7
58
60
61
62
64

65
67
69
70
LIST OF TABLES

Page
III-1. Summery of Calibration Data for Rod D . 8
ITI-2 Summary of Calibration Data for Rods B and C 8
III-3 FNeutron Flux Traverses in Cell Q-13 16

IIT-4 Neutron Flux Traverses Through Cells 0-12 and P-12 16

II1-5 Radial Flux Traverse in Mid-plane o 17
III-6 Neutron Flux Distribution Across Fuel Annulus 20
I1T-7 ©Power Distribution in Cell Q-13 20
III-8 Power Distribution Across Fuel Annulus 22
ITI-9 Power Distribution Through 0 02" Fuel Disk 25
III-10 TFuel Cadmium Fractions for Cell Q=13 25
III-11 Effect of Stainless Steel on Power Distribution 26
IV-1 Rod D Calibration . . . 3k
IV-2 Redial Flux Distribution in Mid-plane 3k
IV-3. Radisel Flux Distribution 6-1/4" from Mid-plane 38
Iv-k Axial Power Distributions in N-1k 42
IV-5 Power Distribution Across Fuel in Mid-plane L2
Iv-6. Sumary of Neutron Leakage Measurements Wy

IV-7T Reactivity Loss Due to Poison Rods Inserted in 0-12 45

Iv-8 Cadmium Importance Function . 48
IV~9 Danger Coefficient Measurements . 49
V-1 Radial Neutron Flux Traverse in Mid-plane 55
VI-l. Radial Flux Distribution in Mid-plane 56
VI-2 Radial Power Distribution in Mid-plane 59
VI-3. Axiel Power Distribution Through M-12 29

VI-k. Fuel Cadmium Fractions in the Mid-plane . 63
-X -

Page
VIiI-1 JImportance Function of Three Stainless Steel Rods 68

ViI-2 Thermal Neutron Leakage T1

ViI-3 ©Power Distribution Across Fuel Annulus at Mid-plane 71
I INTRODUCTION

The Reflector Moderated Reactor (RMR) as presently conceived is a

high power, high temperature, circulating fuel reactor, which 1is
charscterized by a central beryllium i1sland surrounded by a uranium
bearing fluoride fuel annulus and a beryllium reflectorl Faive

critical sssemblies have been constructed in which the essential

features of the RMR were mocked-up The series of experiments, which
were carried out at room tempersture and essentially zero power, provide
experimental data on the criticel mass snd neutron flux and power distri-
butions that cen be compared to those predicted by multigroup calculation
for the assembly in question In addition, danger coefficients were
obtained for a large number of materials placed at various points in the
reactors

The first assembly (CA-10) consisted essentially of a central 12" cube
of beryllium partially surrounded by a 3" fuel annulus and a 12" thick
beryllium reflector backed by graphite The fuel annulus contained
sodium filled stainless steel cans between which were placed 0 01" thick
uranivm disks The majority of the measurements were teken with 15 1 kg
of U-235 in the reactor, which provided an excess reactivity of 1 9% in

Kepr

The second assembly (CA-11) differed from the first in that, 1) the
fuel region contained an homogeneous mixture of Zr0Op, NaF, C, and UF}
packed 1n aluminum tubes, 2) the thickness of the fuel annulus was
increased to 4-1/2", and end fuel ducts were provided, 3) the central
island region was arranged essentially as a 9" cube, and 4) the regions
were assembled to approximate concentric spheres somewhat more closely
This assembly was critical with 7 7 kg of U=235

1
The third assembly (CA-12) was originally designed to evaluste the
characteristics of a graphite moderated and reflected assembly  The
fuel region was built up as a cylindrical shell 21" OD, 3" thick and
36" long The fuel used was similar to that used in CA-11 with the
U~-235 concentration approximately doubled This arrsy was not critical
Approximately 80% of the graphite in the island and the inner 6" of the
reflector was then replaced by beryllium in order to meke the reactor
critical The fuel region contalned 17 3 kg of U-235 1In the second
modification of this reactor, the 17 3 kg of fuel were rearranged as an
ellipsoid with the beryllium and graphite surrounding the central fuel
region This two region reactor was not critical

e

1 Fraes, A P and Mills, C B , "The Fireball, A Reflector Moderated
Circulating - Fuel Reactor" Y-F10-104, June 20, 1952
The fourth assembly (CA-13) was composed of a central ellipsoid
contalning a 1 to 1 volume ratio of graphite and the fuel used 1n
CA-12 This core was surrounded by a beryllium reflector 12" thick
The reactor was critical with 15 2 kg of U-235

The last assenmbly (CA-14) was similar to CA-1l except that the fuel
with the higher uranium concentration was used and an effort was made
to simulate the poisons (core shells, coolants, coolant tubes, etc )
that will be present in a full scale reactor It was critical with
18 2 kg of U-235

ITI DESCRIPTION OF ASSEMBLY AND MATERIALS

The critical experiment assembly has been described in other reports,
and only a brief description of the components that are necessary for
an wmderstanding of the physical make up of the reactors will be
given

The assembly apparatus consists basically of a matrix of square 2S
aluminum tubes, which form a 6' cube and into which may be placed the
reactor meterials The 6' cube 18 divided into two identical halves,
one of which is stationary and the other movable by remote control
Each half consists of 576 tubes, stacked in a 24 x 24k cell array

These tubes are 36" in length with a 3" x 3" outside cross-section

and have O O47" thick walls Part of the reactor materials are

placed 1n each half and the assembly made critical by control rod
adjustment after the two halves have been brought together A
reference system 1s used to designate each cell A letter desig-
nation is given each column of cells and a number for each row  This
system 18 evident from any of the loading diagrams  The assembly, with
the two halves separated, can be seen in Fig II-1 (The materials
visible in the upper portion of the movable half have no relation to the
subject experiments )

The basic materials present in the reactors beside the fuel and aluminum
matrix were beryllium and graphite These materisls were in the form of
square blocks, either 2-7/8" x 2-7/8" or 1-7/16" x 1-7/16" and of
various thicknesses Each block had a 0 196" hole drilled through its
center, normal to the square cross-section When required, ,a skewer rod
was inserted through these holes to form a "shish-kabob™ The skewer
rods had a 3/16" diameter and were composed of aluminum excépt for those
"shishes” used as safety or control rods in which case a steel skewer
was used All safety and control rods were normal reflector or island
shishes

 

1 Bly, F T et al, NEPA Critical Experiment Facility, NEPA-1769,
April 15, 1951
Fig. ll-1

 

Table Assembly.

B
Tfi%\éfi§15h354(93*§% U-235) uranium metal disks used were approximately
O O1" thick with diameters of 2 860" and 1 430" and weighed 18 0 £+ 0 1
and 4 5 + 0 1 grams respectively The weight tolerance was obtained by
punching small holes in some of the disks Each disk had a 0 196" hole
drilled through the center The powdered fuel mixtures are described
under the appropriate reactor designation

The appendix contains quantitative information on the materials
located in each region of all the assemblies, and also spectrographic
analyses of some of the materials used

IITI CA-10

A Assembly Loading

 

The initial loading for CA=10 1s shown in Figs III=l and III-2,
which are cross-sections of the assembly at the mid-=plane of the
reactor and the wvertical plane containing the central axis of the
reactor, respectively It is seen that the assembly consists of a
12" central cube of beryllium, partially surrounded by a 3" fuel
annulus composed of sodium filled stainless steel cans (2-7/8" x
2-T/8" x 1") between which were placed the large sized O OL" thick
uranium disks The components and arrangement of a typical fuel
shish are shown in Filg III-3 As can be noted from Fig III-2,

a section 6" square at both ends of the fuel amnulus was filled
with beryllium rather than the uranium disks and sodium The

fuel region was loaded with two fuel disks between each sodium
filled can except for the 3" end fuel regions behind the central
beryllium cube, which had only one disk between each can The
fuel annulus was surrounded by a 12" thick beryllium reflector
backed by 6" of graphite on the sides and 8" on the ends of the
reactor

With the loading described above the reactor contained 15 0 kg

of U=235 which at a room temperature of 7h°F provided an excess
reactivity of 1 9% in k.py The reactor was made gritical

by removing the reflector from cells J=13 and 14 and I-13 (1 &%)
and by adjustment of the control rods Measurements that required
more reactivity than was available on the control rods were
accomplished by imserting one or more reflector shishes in the
above cells All foll and danger coefficient measurements were
taken on the opposite side of the reactdr where the effect of the
voids on the measurements was negligible

B Control Rod Calibrations

 

The change in reactivity introduced by the linear displacement
of each of three control rods was determined Rod D was cali-

 
ORNL—LR—DWG 2562

J K L M N 0 P Q R S T U \ W

H

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Fig III-1 Assembly Loading at Mid-Plane

 

 

 

 

 

 

L
IIIMITIMIINNN
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wx/ N3 N 32, NV AN AN
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NN ZANNNINZZ AN
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\\ 227N\
N\ 7722\
NN NN

Mk .

MmN

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

444444444444444444

~r—
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

t A
ORNIH—DG 2563
.
RN RN RN R R,
BN NN A
AN NN RN N NN\
NN D NN
NN 00 NN
NN Ay
o NN NN N e
NSANARNAANS
Wy
NN
NN SN
22NN NN o o 7777
A
P S
UYL )R
J i

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Fig IlI-2 Axial Loading of Assembly

 

in Vertical Plane Containin g Reactor Axis

 
 

Fig. IlI=3. Typical Fuel Shish.
-8-

 

 

 

 

o
Y
brated by the integrated period method after Whl%h rods B and ;
C were evalunted by direct compmrason with Rod D°. Detailed <
data for Rod D are given in Table III~1 A graph of reactivity !
of this rod in cents, versus its linemr displacement i1s shown in
Fig III-k, The average change in resctivity per wat dis-
placement for each inerement of displacement is plotted for the
same rod in Fig. III-5, Siftilayr data for rods B and C are shown in
Teble III-2 The delayed neutron fraction has been taken as O 0073
TABLE IIT - 1
SUMMARY OF CALITBRATION DATA FOR ROD D
Pogition of Rod Total Change Average Incremental
From Mid-Plane 1n Resctivity Sensitivaity in Cents
in Inches o Value in Cents per Inch
00 0.0 -
2+555 2 01 0 79
7« Blils 8,98 1 %2
104245 14,62 2435
11 987 19 80 2497
13 560 24 Th 3 14
14 g21 28 98 3 12
16 290 32 50 2.57
17 623 35,02 1.89
19 19% 37 T9 1 76
25 697 42 69 075
TABLE III=-2
SUMMARY CALIERATION DATA FOR RODS B AND C
Distance of Rod Reactivity in Cents Average Sensitavity
From the Mid-plane in Cents per Inch
in Inches Rod B Rod C , Rod B Rod C
00 00 00 - -
50 11 3 35 2 26 0 70
10 O 20,9 12,2 192 1 7h
1540 28 5 28 2 3 20 320
20.0 33,0 39 3 2.22 2 27
25 0 35.0 43,2 0 78 0 78

 

2 Callihan, D. "Critical Experiments on Direct Cycle Aircraft
Reactor" ORNL~1615, October 22, 1953, p 22
REACTIVITY CHANGE (cents)

60

50

40

30

20

10

ORNL-LR-DWG 2564

 

 

 

 

 

 

 

 

 

 

 

 

 

 

LINEAR DISPLACEMENT OF ROD (in)

Fig II-4 Cahibration Curve For Control Rod D

_— il
.//
~
o /.
S
/.
/
/
/O
o
/
/
"
 

IN REACTIVITY PER UNIT LENGTH (cents /in }

CHANGE

n

ORNL LR DWG 2565

 

 

 

 

 

 

 

 

 

 

 

 

 

10 15

20

LINEAR DISPLACEMENT OF ROD {imn)

Fig Il 5 Average Sensitivities of Control

Rod D

25

30

v D
C.

~11-

Temperature Effects

 

The changes in reactivity produced by changes in temperature

over a limited range were determined by changing the demand
temperature of the assembly room Several days were allowed

for the reactor temperature to reach equilibrium The reactor

was made critical before and after the temperature changes,

and the reactivity differences determined in terms of the
calibrated control rod settings The temperature of the reactor
was determined from two iron-constanteh thermocouples Figure ITI-6
shows the reactivity changes plotted as a function of the assembly
temperature A systematic study was not undertaken to determine
the sources of this temperature effect, it was determined primerily
to correct all reactivity measurements to a common reactor
temperature

Neutron Flux and Power Distraibution

Neutron flux distributions were obtained using bare and cadmium
covered indium foils The techniques used for these measurements
have been ade%uately covered in previous reports and will not be
repeated here Power distributions were obtained using the
aluminum catcher foil technique which is also described in the
same reference

l TPlux Distributions

 

Bare and cadmium covered indium traverses were made through
the fuel core, starting at the mid-plane and extending to the
edge of the graphite reflector and parallel to the axis of

the reactor through cell Q-13 Table III-3 contains data
showing the positions of the bare and cadmium covered foils
and thelr relstive activities Figure III-T7 shows the relative
activities plotted as a function of the distances of the foils
from the mid-plane

Bare and cadmium covered indium traverses were also made
through the beryllium island parallel to the axis of the
reactor, starting at the mid-plane and extending to the edge
of the graphite reflector, through cell 0-12 and P-12 The
traverses made through cell P-12 pass through the fuel core
while those through 0-12 do not Table III-4 contains data
showing the distances of the foils from the mid-plane in the
respective cells and also the relative activity obtained for
each fo1l Figures IIT-8 and III-9 show the graph of the

 

3 TIbid, p 36
IN REACTIVITY {cents)

CHANGE

20

20

Lo d

ORNL LR DWG 2566

 

 

 

 

 

 

 

 

 

 

 

55

65

75 8BS
ASSEMBLY TEMPERATURE ( F)

Fig Il 6 Change in Reactivity vs Temperature

95

105
_gt_

28

24

20

16

12

RELATIVE ACTIVITY

ORNL LR-DW! 2567

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

DISTANCE FROM MIDPLANE OF REACTOR (in)

Fig III-7 Bare and Cadmium-Covered Indium Traverses Through Fuel Core

- FUEL CORE — BERYLLIUM | e GRAPHITE — =
m, BARE INDIUM ACTIVATION
X
N CADMIUM FRACTION
CADMIUM COVERED ]
T T~ | . ./ INDIUM ACTIVATION
..I ™
T e —— o ° 0/! \ \
—"-—--._————_
.
0 5 10 15 20 25 30

10

05

CADMIUM FRACTION
 

_-bl_

RELATIVE ACTIVITY

45

40

35

30

25

20

10

ORNL-LR-DWG 2568

 

 

 

-BARE INDIUM ACTIVATION

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Fig

lI-8 Bare and Cadmium Covered

\\ o
CADMIUM-COVERED
INDIUM ACTIVATION-/ \\\(
| - —— —
10 15 20 25 30
. A

DISTANCE FROM MIDPLANE OF REACTOR {(in)

Indium Troverses Parallel to Axis Through Cell O0—-12

05

CADMIUM FRACTION
_gt_

RELATIVE ACTIVITY

32

28

24

20

16

12

ORNL LR DWG 2569

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

BERYLLIUM _ | | FUEL _ | _ -l
{SLAND T AVERTTT BERYLLIUM GRAPHITE ——
BARE INDIUM ACTIVATION
o——@ "\ CADMIUM FRACTION
. ——
N\ " .
\ / |
CADMIUM-COVERED INDIUM ACTIVATION //\.\ \
0 5 10 15 20 25 30
DISTANCE FROM MIDPLANE OF REACTOR (in)

Fig III—9 Bare and Cadmium-Covered Indium Traverses Parallel To Axis Through Beryllium Island and Fuel

05

CADMIUM FRACTION
-16-

relative acticity of the various foils snd the corresponding

cd fraction®* as a function of their distances from the mid-plane
of the reactor for the traverses through cells 0-12 and P-l2
respectively

TABLE III-3

Neutron Flux Traverses in Q-13

Distance of Foll from Relative Activity

 

Mid-plane in Inches Bare Foils Cadmium Covered Foils
00 10 338 8 08L
05 9 118 7 889
15 9 396 T 959
25 8 899 7T 663
L5 8 302 T 393
65 8 068 6 910
85 8 532 7 160
95 9 533 T 370

10 0 13 370 8 569

11 0 17 815 8 580

15 0 15 502 4 578

Z1L 5 182 0 46k

23 0 2 97k 0 341

28 0 o0 767 0 078
TABLE III-4

Neutranr Flux Traverses Through Cells 0«12 apd P<l2

Cell 0-12 Cell P-12
DistancesFrom Distances From
Mid-plane in Relative Activity Mid-plane in Relative Activaty

 

Inches ~ Bare  Cd Covered Inches Bare = Cd Covered
03 35 335 17 232 0.3 26 678 1k 736
27 33 Th6 17 702 20 25 T35 14 931
60 25 676 14 596 Lo 21 990 13 557
75 22 315 13 008 5 7 13 602 10 234
9 0 22,349 1241k 6 5 11,401 8 938

ik o 20 537 6 840 7 5 11.137 8.599

17 0 12 169 2 675 8.5 11 655 8 463

21 0 5 250 0 614 9 2 13,954 9 09k

21 7 4 790 0.551 12.0 21.436 9 662

22 T 3 601 0 428 1540 17 140 5 158

25 0 2.307 0,54 18 0 10 440 1 980

29 0 0 926 0 o5k 21 O 5 222 0 638

23 0 3 640 0 406
270 1 388 0 142

 

¥Cd fraction 18 defined as the Praction of Ihe total ac‘l::t.‘vlitly induced by
neutrons with energies less then 0 02" cadmium cut-off
-17-

Radizal bare and cadmium covered indium traverses were made starting
at the axis of the reactor in cell 0-12 and extending to cell X-12
Table IIT-5 contains the data showing the positions and relative
activities for both the bare and the cadmium covered foils Fagure
III-10 1s & graph of the relative activities plotted as a function
of the distances of the foils from the axis of the reactor

TABLE ITII-5

Radisl Flux Traverse in Mid-plane

Distance of Foill

 

From Axis an Relative Activity
Inches Bare Cd Covered
o7 38 543 19 677
23 35 595 19 826

4 5 26 173 15 230
63 12 137 9 683
75 9 889 8 500

8 7 11 861 8 898
10 5 20 692 11 568
13 5 21 483 8 509
16 5 14 764 3 930
19 5 8 310 1 317
22 5 4 125 0 515
25 5 1 675 0 197
27 3 0 542 073

In order to obtain a more detsiled pattern of the neutron flux in
the fuel region, short radial traverses were made across the fuel
in cell N-10 Table III-6 shows the positions of the folls and
their relastive activities Figure ITI-11 shows the activitles
plotted as & function of the distances of the foils from the axls
of the reactor
RELATIVE ACTIVITY

45

40

35

30

25

20

15

ORNL—LR—DWG 2570

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Fig III—-10

DISTANCE FROM REACTOR AXIS {(in)

Indium Traverse Radially in Midplane of Reactor

BERYLLIUM FUEL
SLAND AND BERYLLIUM REFLECTOR GRAPHITE
SODIUM
/——BARE INDIUM ACTIVATION
\ CADMIUM FRACTION
o\ / /
@
/ | \
\. // ~ S
\\./. \\ AN
\/ : \\
CADMIUM-COVERED INDIUM ACTIVATION —)\o_
| | \_._'—'——-—jn_):_g;
5 10 {5 20 25

Cell O—-12 to X—12

30

08

06

04

02

CADMIUM FRACTION
_el_

RELATIVE ACTIVITY

28

24

20

16

12

ORNL-LR-—-DWG 2574

 

CELL N—={4 ——

—t—— CELL N—-10 ————=

—=— CELL N-9

 

 

 

 

S~ BARE INDIUM ACTIVATION

 

 

 

\:\‘:__

 

 

 

\CADMIUM—COVERED
INDIUM ACTIVATION

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Fig III - 14

6

DISTANCE FROM AXIS OF REACTOR (in)

Bare and Cadmium-Covered

10 12

Indium Traverses Across Fuel Annulus

 
-20-

 

 

TABLE III-6
Neutron Flux Distribution Across
Fuel Amnulus
Distance from
Axis of Reactor
In Inches Bare Cd Covered
b 7 21 172 14 419
55 1h 603 9 977
6 3 10 242 8 923
67 9 554 T 733
T1 9 038 8 098
76 8 799 T 270
80 9 200 T 991
8 4 9 060 7 466
8 7 9 104 8 018
93 12 837 8 L46
10 3 17 857 10 718

2 DPower Distributions

The power distribution through the fuel was measured along the
axis of cell Q-13 The aluminum catchers were mounted vertically
between the wranium fuel disks and the adjacent sodium cans

Table III-T contains data showing the distances of the aluminum
foils from the mid-plane of the reactor Figure III-12 is a graph
of the relative fission rates plotted as a function of the distance
from the mid-plane The two values shown at 9 5" from the axis
Indicate different fission rates on the opposite surfaces of the
uranium disks which are adjacent to the sodium cans and the

beryllium, the higher fission rate occuring on the surface adjacent
to the beryllium

5
TABIE III=T7
Power Distribution in Call Q-l13

Daistance from

 

Mid-plane i1n Inches Relative Activity
05 3182
15 3066
35 2899
> 5 2563
75 2253
95 2806
95 5318
_LZ_

RELATIVE FISSION RATE

<lllPw
ORNL LR DWG 2572

 

 

 

 

 

—0

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

3 4 5 6 7 8
DISTANCE FROM MIDPLANE OF REACTOR (in)

Fig M-12 Power Distribution in Cell Q—-143

 
=00

A power traverse was also made across the fuel annulus using
cell Q=13 Two runs were made and the fission fragment coumting
rates normalized Table III-8 contains the data showing the
distances of the centers of the aluminum catcher foils from
the ax1s of the reactor and the normaslized fission fragment
activities Figure III-13 is a graph of the activity plotted as
s function of the distances of the centers of the foils from the
ax1s of the reactor The dotted portions of the curve extrapolate
the activities to the edge of the fuel

TABIE II1-8
Power Distribution Across Fuel Annulus

Distance From

 

 

Axis in Inches Relstive Activity
6 2 661
63 596
6 6 588
69 508
6 9 511
T3 485
T5 469
75 L75
77 482
81 501
8 % 562
8 L 570
8 7 629

The depression in the fassioning distribution through two adjacent
O 01" fuel disks was determined by replacing them with ten 2 mil
thick fuel disks separated by catcher foils thereby forming a
sandwich This sandwich was inserted in the normal fuel position
between two sodium filled cans The data obtained are shown in
Table III-9 and Figure III-1} Two catcher foils were inserted
between the intermnal disks giving the two wvalues reported To
facilitate drawing the curve, symmetry hms been assumed and the
points reflected
_EZ -—
RELATIVE POWER

80

75

70

6 5

60

55

50

4 5

.
ORNL-LR DWG 2573

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

BERYLLIUM ISLAND T el FUEL ANNULUS — BERYLLIUM
\
O
N 7 °
v
O
5 6 7 8 10
DISTANCE FROM REACTOR AXIS {(n)
Fig III -13 Power Distribution Across Fuel Annulus <o

 
RELATIVE ACTIVITY

30

25

20

15

ORNL LR DWG 2574

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

®
| EXPERIMENTAL POINTS
/ ~——REFLECTED POINTS
20 mils OF URANIUM
4 6 8 10 12 14 16 18 20

URANIUM THICKNESS (mils)

Fig 1II-14  Power Distribution Through 20 mil

Uranium Disk

 
L

-25-

TABLE III-O

POWER DISTRIBUTION THROUGH O 02" FUEL DISK

Distance Through

 

Sandwich in Relative
Mils* Activity
0 31 64

6 19 63

6 19 23

10 18 58

10 17 44

16 21 19

16 20 4k

20 29 90

#Catcher foil thicknesses have been neglected

Fuel Cadmium Fractions

 

Cadmiwm fractions for the uranium fuel disks were messured at
various positions axislly along cell Q-13 Table III-10
contains the wvalues of the cadmium fractions at various
distances from the mid-plane The two values listed for the
9 5" position indicate that a measurement was made on both
gides of the uranium disk

TABLE III-10

FUEL CADMIUM FRACTIONS FOR CELL Q=13

Distance From Fuel

Mid-plane in Cadmium

Inches Fractions
l5 071
55 0 68
95 0 75
95 0 85

Effect of Stainless Steel on Power Dastraibution

In order to determine the effect of a stainless steel core
shell on the power distribution across the fuel, the beryllium
reflector i1n cells R=10 through R-15 was rearranged so that a
void (5/8" x 2-7/8" x 9") existed adjacent to the fuel annulus
extending on one side of the mid-plane only A 3" longrtudinal
-26-

section in fuel cell Q=13 was rotated 900 about a vertical
axis from its normally loaded position The scdium filled
cens were replaced by aluminum spacers and some of the O OL"
fuel disks replaced by O 002" disks A radial power distri-
bution was then measured in this 3" section  Stainless
steel sheets (9" x 2-7/8" x 0 29") were then inserted in the
voi1ds in cells R=10 through R=-15, and the power distribution
remessured The data obtained from both traverses are shown
1n Table III-11 and Figure III-15

TABLE ITI-11

EFFECT OF STAINLESS STEEL ON POWER DISTRIBUTION

 

 

 

Distance of Foil Relative Activity
From Reactor Axis Without With
In Inches Stainless Stainless
Steel Steel
62 4 885 4 912
651 3 348 3 348
6 Si 2.287 2 359
T 2 1 977 1 885
T 5 1 745 1 680
T5 1 816 1 628
ToT 1 971 1 672
8 3 2 822 2 246
85 3 546 2 565
8.7 5 199 3 526

E Danger Coefficient Measurements
1l Stainless Steel

The loss in reactivity due to the presence of stainless steel
next to the fuel annulus wes determined for three different
steel thicknesses inserted* in the wvoid in cells R=10 through
R-15 described above The void itself resulted in a loss in
reactivity of 8 7 cents The loss in reactivity as a function
of the stainless steel thickness is plotted in Figure III-16

2 Bismuth and ILead

A 1-1/8" layer of beryllium adJjacent to the fuel surface and
extending 8=5/8“ on both sides of the mid~plane was removed
from cells R=10 through R=15 The introduction of this voad

 

* The fuel region extended slightly beyond the steel inserts so the
fuel disks adjacent to the beryllium were not fully shielded
RELATIVE POWER

Aniline
ORNL LR DWG 2575

 

 

 

/‘\\ WITHOUT STAINLESS STEEL

 

 

 

 

 

 

 

 

 

 

 

 

DISTANCE FROM AXIS OF REACTOR (in)

Fig I 15 Effect of Stainiess Steel on Power

®
' /
/\\WITH STAINLESS STEEL
@
/
&\3 " ///
\ " o
6 7 8 9 10

Distribution o

 
LOSS IN REACTIVITY { hundreds of cents)

no

agm»
ORNL —LR—DWG 2576

 

 

 

 

 

 

 

 

 

0075 0150 0 225 0 300
THICKNESS OF STAINLESS STEEL {in)

Fig II-16 Loss in Reachivity Due to Addition of Stainless - Steel Core Shells
-29-

resulted in a loss in reactivity of 92 cents Six pleces of
bismuth (2-7/8" x 2-7/8" x 1") were then placed in each of
these cells to partially fill the vord The addational loss
in reactivity due to the bismuth was 32 cents The bismuth
was then removed and replaced by lead, with a resulting loss
in reactivity of 121 cents

Stainless Steel and Sodium

 

Four stainless steel cans filled with sodium were removed
from cell @-13 Four aluminum spacers, each approximately
1/4" x 1-7/16" x 1-7/16", were substituted for each of the
sodiwm f1lled cans to anchor the fuel disks in place The
change in reactivity due to substituting the sluminum spacers
for the cans was determined The aluminum spacers were then
replaced by four empty stainless steel cans similar to those
contalning the sodium: thus the change in reactivity due to
the removal of the sodium alone was obtained A gain in
reactivity of 8 4 cents was obtained by substitutiom of

the aluminum spacers for the sodium filled cans, and a gain
of 2 8 cents was realized by the removal of only the sodium

IV - cA-11

Assembly Loadlng

The fuel region of the second assembly (CA-1l) had properties
that more nearly simulated those of the conceived full scale
RMR The fuel used was an homogeneous mixture of Zr0Oz, NaF,
C, UFy, and a smell amount of adsorbed moisture The wranium
was enriched to 93 2% U-235 The composition of the mixture
wes

16 2 wt % UF),
55 0 wt % Zr0g
19 9 wt % NeF
88wt%cC
0 25 wt % HoO

The wranium density was O 21 gnm U-255/cc and the average packed
density of the mixture was 1 8 gm/cc ZrOs was used in place of
ZrF) with the carbon added to compensate for the difference in the
thermsl neutron scattering properties of the mixture The
powdered mixture was packed in 1-1/4" 0 D square aluminum tubes
-30-

with & 1/16" wall thickmess These fuel tubes were arranged in an
annulus around a beryllium island with a beryllium and graphite
reflector surrounding the fuel Figures IV-1 and IV-2 show the
loading at the mid-plane and in the vertical plane containing the
reactor axis, respectively Inasmuch as there were no fuel tubes
4-1/2" long aveilable, %-1/2" shishes were made using O Ol" quarter
size fuel disks and aluminum spacers These shishes sre shown as
vertical lines in Fig IV-2 The fuel annulus wes 4-1/2" thick in
the central region with end ducts that were one fuel tube thick

It should be noted that while the thickness of the fuel region has
been reported as 4-1/2", there were three fuel tubes (3-3/4" total
thickmess) across this dimension In general the fuel tubes were
arrenged such that as small a void as possible existed between the
fuel tubes themselves TFig IV=3 shows the arrangement of the
fuel tubes The assembly was critical with 7 7 kg of U-235 1In
order to gain the additional reactivity required for some of the
mesgurements, ©.01" thick full size fuel disks were added to the
island and/or fuel region as required

Evaluation of Control Rod D

Control Rod D was evaluated by subdividing the rod into wmits of
equal reactivity The evaluation was accomplished by mesns of the
following procedure The reactor was initially masde critical with
Rod A all of the way in, Rod D all the waey out, and Rod C gppropri-
ately located Rod A was then withdrawn one inch and the reactor
returned to the ceritical condition by inserting Rod D Rod A wsas
then Inserted one inch and the reactor retwrned to the previous
critical condition by withdrewing Rod C This sequence of operations
was continued wmtil Rod D was completely inserted By this means Rod
D was divided into increments of equal reactivity over i1ts entire
length The wnit of reactivity was determined from the measurement
of the displacement of Rod D equivalent in reactivity to the removal
of a safety rod, which had been evaluated previously by the "Rod
Drop Method" The data thus obtained for the calibration ¢f Rod D
are shown in Table IV-1 and Fig IV-4. The position of the rod

i3 measured from the assembly interface

Neutron Flux Distribution

Two radial flux traverses were taken, the first in the mid-plane
of the reactor and the second 6-1/4" from the midplames The
second traverse passed through the quarter size fuel shishes used
instead of the fuel tubes Tables IV-2 and IV-3 and Figs IV-=5
end IV-6 show all the data obtained fer both the bare and cadmium
covered indium tratverses
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

I J K L M N O P Q R S T U Vv w

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

...
.
.
...
e
NN\
NN
N
N N
oo\,
.y
.
o
.
- e
«
R ;

 

 

222222222

- ~— -— -— - -~ -~ - oJ

o

Fig IV~1 Assembly Loading at Mid Plane
RAPHIT

MOVABLE STATIONARY
TABLE S TABLE

Fig IV—2 Axial Loading of Assembly in Vertical Plane Containing Reactor AXIS e

ORNL-LR—-DWG 2578

 
-Se-

 

 

 

Fig. IV=3. Fuel Tube Arrangement.

 

PHOTO 12144

 
-3L-

b

uh

TABLE IV-1 b

Rod D Calibration

 

Position of Rod Change in Average Sensitivity
From Assembly Interface Reactivity an Cents per
In Inches in Cents Inch
00 0O 00
0 530 0 88 1 66
2 930 8 48 3 17
5 660 16 13 2 80
9 26k 23 78 2 12
12 980 31 38 2 05
15 980 38 99 2 54
18 Tho L6 62 2 76
21 880 54 24 2 43
26 120 61 90 18
33 TT4 69 51 099
TABLE IV-2

Radial Flux Distribution in Mid-Plane

 

 

Distance From Relative Indium Activity
Axis in Inches Bare Cd Covered
0 00 33 569 21 200
1 25 31 873 20 380
2 75 2k 864 16 646
L 63 13 801 10 545
5 87 10 989 9 072
7 25 10 810 9 151
8 50 12 81k 9 433
10 25 19 524 10 799
11 75 21 080 9 435
13 25 20 608 7 360
15 0 16 TOk L 757
18 0 9 T12 1 647
21 0 4 585 0 469
2k O 2 288 0 216
27 O 0 505 0 O4é
LOSS IN REACTIVITY (cents)

80

76

72

68

64

60

56

52

48

a4

40

36

32

28

24

20

ORNL—LR—-DWG 2579

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

15 20

ROD DISPLACEMENT (in)

Fig IV—4 Calbration Curve for Control Rod D

25

30

35
RELATIVE ACTIVITY

40

35

30

25

20

10

ORNL LR-—-DWG 2580

 

|

 

 

et—— BERYLLIUM —:—{-t—

1
I
FUEL -]

BERYLLIUM

Y

|
GRAPHITE———-—l

 

L)

 

 

)

 

 

 

 

 

 

 

 

 

 

 

 

 

\ BARE INDIUM ACTIVATION CADMIUM FRACTION~— 5
/
\\ /
/'><
TN
CADMIUM COVERED
INDIUM ACTIVATION-"]
H\Q

0 5 10 15 20 25 30

RADIAL DISTANCE FROM REACTOR AXIS (in)

Fig IV 5 Radial Flux Distnbution at the Mid plane

05

CADMIUM FRACTION
RELATIVE ACTIVITY

ORNL LR DWG 2581

 

~————— BERYLLIUM

 

DISKS

 

- BERYLLIUM ———A =

 

-

A~

 

T —————_ BARE INDIUM ACTIVATION

 

 

 

 

CADMIUM COVERED
INDIUM ACTIVATION

 

_—

 

 

_

Kcnowum FRACTION

 

 

 

\__________/

 

 

 

 

 

 

 

4 6
RADIAL DISTANCE FROM AXIS ( )

8 10

Fig IV 6 Radial Flux Distribution 6% in from Mid plane

05

CADMIUM FRACTION
-38-

TABLE IV~3

Radiel Flux Distribution 6-1/4" from Mid-Plane

 

 

Distance from Axis Relative Indium Activity

in Inches Bare Cd Covered
o7 Te523 5 LaL
23 6.579 L 679
33 L, 563 3 589
L 3 L 1k 3 292
4 9 3 882 3 27h
5T 3 949 3.202
63 3 953 5 120
71 L 366 3 364
76 L 866 3 346
87 6 294 3 634
97 T 229 3 532
113 T.389 3 076

The associated radiasl cadmium fractions for indium obtained in the
mid-plane from CA~10 and CA-1ll are shown in Figure IV=-T.

D Power Distributions

Power distributions were obtained by placing O 002" thick quarter
size uranium disks wlth sluminum catcher foils adjacent to fuel
tubes in the deslred location. The first was measured parallel
to the reactor axis in cell N-1L The 0.002" thick uranium disks
gnd associated catcher folls were placed adjacent to the fuel
tube between the fuel tubes and the beryllium island. A second
power distribution wes taken i1n the same location with Boral#®
strips (9" x 2-7/8" x 1/4") placed sround the end fuel ducts an
the stationsxy table, replacing some beryllium and graphite as
shown in Figure IV~8 The data have been arbitrarily normalized
at a point 14 5" from the interface In order to maintain the
reactor craitical with the Boral inserted, additiomal 0.01" fuel
disks were added to the beryllium island on the movable table

The data obtained from both rums is shown in Table IV-4 and
Figure IV-9

% The Boral wis prepared fram z mixture of 35 wt % BEE and slumimam,
This mixture is held between two aluminum sheets, each about 0.04"
thick forming a 1/L" sandwich comtaining 250 mg of borom per squsre
centimeter. The boral strips wereé wrapped in tape to minimize
FPawen contamination.
CADMIUM FRACTION

to

05

ORNL—LR-DWG 2582

 

r=— CA—-1{0— BERYLLIU

|
|
|

M—————-J'--—FU

-‘—CA—H—BERYLLIUM—’J‘—- FU

|
|
|
|

EL

BERYLL UM ——————— e

EL——T-(——— BERYLLIUM—————t——

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

DISTANCE FROM CENTER OF FUEL REGION {(in)

Fig IV=7 Cadmium Fractions in CA—10 and CA—44

o w— Y
A
‘é)\< CA—1O\
cA—1
6 4 0 2 4 6 8 10 12 14 16 18
et

NN \\\\\\ \\\W TVl s i
NUNIINUINIUNIN \\

GRAPHITE

 

 

 

 

 

 

 

 

  

 

BERYLLIUM FUEL BORAL

v ol
VA Y
N7

/ N%
N
\//

NN
L /%

 

 

 

 

11

 

 

 

12

 

 

13

 

 

 

 

 

NN
\\\\7///////\\\\\

14

 

 

Fig IV-8 Position of Boral Inserted Around End Ducts
on Stationary Table

_40_
RELATIVE ACTIVITY

ORNL—LR-DWG 2584

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

BERYLLIUM REFLECTOR
FUEL
‘ } | l ‘ | BERYLLIUM ISLAND R = == = — = — 7
E+
' \/TRAVERSE WITHOUT BORAL AROUND END DUCTS
N
TRAVERSE WITH BORAL ARQUND END DUCTS
—A—
0 10 15 20 25 30 35

DISTANCE FROM INTERFACE (in)

Fig IV—-9 Axial Power Distributions in Cell N—1{4
Axial Power Distribution in N-1k

~L4o.

TABLE IV-L

 

Distance from Interface

Relative Actaivity*

 

in Inches Without Borsal With Boral
14 5 6 902 6 902
17 3 5 896 5 647
20 O 6 312 5 469
22 7 6 027 L hés
25 5 5 205 1 160
28 3 3 W7 0 248
31 0 2 602 0 098

A power distribution was also measured across the fuel annulus
in the mid-plane 6f the reactor The 0 002" thick wranium
disks were placed in the horizontel gep (approximately 1/4")
between the fuel tubes of adjacent vertical cells, with eight
5/16" catcher foils distributed across the fuel disks The
data obtained is shown in Table IV=5 and Figure IV-10

TABLE IV~

Power Distribution Across Fuel in Mid-Plane

Distance From Reactor
Axls

O3\ &
£ oW oW o\ M

Relative
Activity

6 955
5 021
L 213
3 163
3 500
3 880
5 17k
T 549

 

¥ Dats normalized at the point 1% 5" from the interface
RELATIVE ACTIVITY

10

ORNL —-LR—-DWG 2585

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

I |
| i | |
| |
g L
| l
VOID- |t VOID—=| }-—
| o BERYLLIUM
BERYLLIUM ISLAND > I FUEL > REFLECTOR
| |
/
\M
2 3 4 5 6 9 10

DISTANCE FROM CENTER OF ISLAND (in)

Fig IV—-10 Power Distribution Across Fuel in Mid-plane
E Neutron leakage Meagsurements

Measurements were obtained in an effort to estimate the ratio of
the end to side neutron leakage from the reactor In the first
set of measurements six bare and cadmium covered indium foils
were placed on the back swrface of the reactor extending from
the reactor axis outward 7" Three sets of folls were placed on
the outer swrface of the graphite side reflector at the mid-plane
These foil measurements were taken without any Boral sroumd the
fuel end ducts

Side and end fast neutron leakage meaSurements were also cbtained
using a fast fisgsion chamber Measurements were taken both with
and without Boral around the fuel end ducts, and with a 1/4" of
Boral around the fission chamber¥#. Slde leskage was determined

by placing the chamber in cell X-12 with the center of the chamber
gt the mid-plane The axls of the fisslon chamber was approximately
1-1/2" from the graphite side reflector The end leskage wWas
obtained by placing the chamber approximately 7" behind the station-
ary table with its axis parallel to the mid-plane with the center

of the chamber on the projected exig of the reasctor The dats
obtained are sumarized in Table IV-6

TABLE IV-6

Summaxy of Neutron leskage Measurements

 

 

Ratic of End to Side Ratio of End to Side
Teaksge With Boral Ieekage Without Boral
Arommd End Ducts Arocund End Ducts
Fission Chamber 6 28
Bare Indium = 23
Cd Covered In = 12
Bare-0d Covered - 11

% Slebs of Boral, 18" x 2-7/8" x 1/4", were placed around the chamber,
which was 10" lomg, thus approximstely 4" of Boral extended beyond
either end.
-45-

F Evaluation of Boral around Fuel End Ducts

The Boral sreets that were placed around the fusl end ducts as
shown in Figure IV-7 were evaluated The gain in reactivity due
to their removal was compensated for by removing fusl disks from
+he 1sland region on the movable table The loss in reactivity
due to the replacement of beryllium and graphite by Boral around
the fuel end ducts amounted to 1 &% in k oo

G Stainless Steel and Boron Poison Rods

 

Two 3/16" diameter rods were evaluated in the center of cell
0-12 1n the 1sland The first rod was a stainless steel tube

40" long filled with natural boron to a linear density of O 1 gm
B/in  The second was a stainless steel rod 38' long The loss
in reactivity in cents for each rod i1s shown as a function of the
distance of the end of the rod from the mid-plane in Table IV-T
and Figure IV-1ll

TABLE IV-7

Reactivity Loss Due to Poison Rods Inserted in 0O-12

 

Distance From Mid-Plane Loss in Resctivity in Cents
in Inches Boron Filled Rod*¥  Stainless Steel Rod
00 W 6 50
L5 29 9 30
90 17 1 -
15 C L 8 -
21 0 0k -

H Cadmium Importance Function

The importance of cadmium as a neutron absorber when placed in the
mid-plane of the reactor was obtained by taping a piece of cadmium
2=-3/L" square and O 02" thick weighing 19 09 gm, to either s block
of beryllium or a fuel tube, and recording the loss in reactivity
for each radial position The data obtained are shown in Table
IV-8 and Figure IV-12

 

¥Further insertion of the Boron filled rod from the mid-plane to the
assenmbly interface (9") resulted in an additional reactivity loss of
30 2 cents
_9-b_

LOSS IN REACTIVITY (cents)

[\
o

75

18
O

ORNL LR DOWG 2586

 

 

 

 

 

 

 

 

 

 

 

DISTANCE FROM MIDPLANE (in)

Fig IV-11 Reactivity Loss Due to Poison Rods Inserted in O-12

/ BORON-FILLED ROD
/ STAINLESS STEEL ROD
0 5 10 15 20

25
75

LOSS IN REACTIVITY (cents)
o
o

N
14!

ORNL-LR —DEG 2587

 

 

 

 

 

 

 

 

 

 

 

 

T
———— BERYLLIUM —-—|<l—- FUEL—’-I‘———— BERYLLIUM —/W—.'

| |

| |
0 5 10 15 20

DISTANCE FROM AXIS (in)

Fig IV—12 Cadmium Importance Function

25
_L8-

TABIE TV-8

Cadmuum Importence Function

 

 

Distance From Axis Loss in Readativity
in Inches B in Cents
00 72 g
2 a9 51
6.%3 16 8
7 3% 18 7
G o1te 36 1
11 9 36 3
15 1 17 5
17 9 69
23.9 12

I Danger Coefficients

 

Danger coefficlents were obtained for a number of different materials
in the reflector and/or fuel regions. Samples were lcocated in cells
Q=12 and Q=13 in the fuel region or adjacent tc the inner face of
cells S=-12 and S~13% after removal of some beryllium reflector All
samples were placed symmetrically gbout the mid-plane The changes
in reactivity, either plus or minus, obtained are shown in Table IV~9
and are referenced to ‘the veld The samples of Li, KF, NaoF, Zry Cr,
and RbF were placed In alumanum containers (5/16" x 2-1/2" x 9") wath
1/32" well thickness, It has been assumed that these containers had no
si1gnificant effect on the reactivity. The sample of sodium was
contained 1n thin walled stainless steel cans, whose reactivity

value was measured with the cans empty, thus ensbling the reporting
of the sodium value. The other samples required no containers  The
reactivity chapnges obtained for these materials are shown in Table
IV-Oa

The reactivity clzmge due to the substitution of a sample of D0 for
beryllium was measured at three locations in the assenbly The
sanmple was 99.2% D50 and weighed 2 65 kgo It was placed at the
assembly mid-plane in the island adjacent to the fuel, in the
reflector adjacent to the fuel and in the reflector 3" from the fuel
layer In each of the first two locations the gain in reactivity
due to the substitution was 12 cents, in the latter position the
reactivity was decreased by 24 cents

J Effect of Stainless Steel Around the Fuel Annulus
In order to estimate the effect on reactivity of the fuel comtaining

core shells required in the full scale reactor, two adjacent
quadrants of the fuel annulus Were partially covered with stainless
-L4g-

TABLE IV-9

Danger Coefficient Measurements

 

 

Sample Sample Weight  Sample* Reactivity Change cents/gm
om Size In Fuel Region In Reflector
LipCOz 152 L A 0 0000 -0 0k15
KF 198 0 A -0 0162 -0 0578
NaF 97 2 A -0 0058 -0 0354
Zr 287 7 A +0 0150 -0 0438
Cr 249 L B -0 0237 -0 0680
RbF 156 3 B +0 0118 +0 0026
Inconel 1956 L A -0 0276 -0 ohk8
Ni 2061 O A -0 027h -0 0k37
Pb 2544 8 C -0 0008 -0 0007
Pb k592 7 D - -0 0012
316 Stainless Steel 1990 &4 C -0 0243 -0 0385
310 " " 3195 0 D - -0 0233
316 " " 995 2 E - -0 0509
Be 753 6 D - +0 0090
Na 366 9 D - -0 0190
B 3931 8 D - +0 0006
Graphite 376 2 A +0 0108 +0 00L3
Density 1/72 gm/cc
Graphite 700 1 D - +0 006k
Densi1ty 1 72 gm/cc
Graphite 825 6 D - +0 0065
Density 2 O gm/cc
BeO 375 0 F - +0 0055

#Sample Size

A - 5/1611 x 5:1 < 9u

B - 5/16" x 2-1/2" x 9
Cc - l/,-l-" X 5_5/ ] x 9n

D - 1" x 2-7/8" x 8-3/8"
E - 1/8" x 5_3/1'_1: X 9n
F - 7/8" x 2-7/8" X 5n
-50-

steel 0 145" thick The exposed surfaces of the fuel tubes,
extending 4-1/2" on either side of the mid-plsne, both between
the fuel and the island and between the fuel and the reflector,
were covered with 9" long plates The total weight of the steel
added was 9 O kg The additional reactivity required was pro-
vided by adding O 01" thick fuel disks distributed throughout
the fuel region The addition of this steel to the quadrant
first covered resulted in a loss in reactivity of 1 45%* in kK pm
and requiring the addition of 828 gms of uranium to remske the
system critical The steel added to the second quadrant also
resulted in a loss of 1 45%* in k.pp and required the addation
of 1818 gms of uranium

 

*The effect on the control rod calibrations due to the steel and
increased fuel loading were not measured, and therefore introduces
some doubt a8 to the accuracy of these data
-51-

V = CA-l2

In this assembly three variations of the essential features of the
RMR were carried out

1 The reactor was first built up with an all graphite island
and reflector with a 3" thick 18" long cylindrical fuel
annulus separating the two regions The fuel region con-
tained 1 5' thick 9" long end fuel ducts on both ends of the
reactor The loading 1s shown in Figure V-1 The fuel loaded
in the 1 25" squave aluminum tubes, was the same as that used in
CA-11 except the Bw235 density was doubled resulting in the
following composition

N

L
02

§3333
By

TREELELE

Q=

20

The U-235 density was O 42 gm/cm’, and the mixture was packed
to an average density of 1 8 gm/cm3° This reactor loaded with
17 3 kg of U-235 was not critical and an apparent source neutron
multiplication of about six was recorded

2 In order to go critical, the assembly was modified by replacing
graphite with beryllium in 31 cells of the assembly Thirteen
of these cells were 1n the i1sland and the remaining 18 were
around the periphery of the fuel annulus

The substitution of beryllium for graphite in the inner reflector
region was then completed as 1s shown in Figure V-2  The excess
reactivity gained by this further substitution was absorbed by
adding 18 1 kg stainless steel in the form of core shells O 059"
thick around the fuel annulus and covering the central 18" fuel
region A radial flux distribution was measured in the mid-plane
of the reactor with the assembly loaded as shown in Figure V-2
The data obtained are shown in Table V-1 and Figure V-3 The
cadmium fraction obtained from the two flux curves 1s also shown
1n the figure

% The final variation of CA-12 was a brief exploration of & two
region reactor using 17 3 kg of U-235 with the fuel axrranged in
a central region The fuel was surrounded by a 3" thick
beryllium and thick graphite reflector TFor this configuration
an apparent source neutron multiplication of about four was
recorded By altering the reflector to provide a 9" layer of
_Zg_

9.n FUEL TUBES
+ 27 sn OF GRAPHITE

36 in GRAPHITE

18 n FUEL TUBES
+18 in OF GRAPHITE

36 1n GRAPHITE

GRAPHITE OUTER PERIPHERY SAME ON TOP AND BOTTOM AS ON SIDES

Fig V-4 Assembly Loading at Mid-piane

ORNL LR DWG 2588
vV W X

 
36 1n GRAPHITE

36 1n BERYLLIUM

Ojol0|C|0]0
OO
VOID O

O

OlO|0OG

OXIOB
< O|O]O1O

AlAlAlA]Y

O
O
O
O
O
O

OOOOOO

Fig V-2 Assembly Loading at Mid-plane

 

 

 

 

 

4
ORNL—LR—DWG 2589

T U V w X

Sin Be +271n GRAPHITE

i2in Be+6mn VOID 4+ 18In
GRAPHITE

{8 1n FUEL TUBE
S-in FUEL TUBE

12 in Be + 24 in GRAPHITE

36 1n GRAPHITE

VOID

Y

 
_bg_

20

RELATIVE ACTIVITY

ORNL LR DWG 2590

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

DISTANCE FROM REACTOR AXIS (in)

Fig V-3 Neutron Flux Distribution in Mid plane

— BERYLLIUM ——wiet-— FUEL -—spf—e—— BERYLLIUM o | centt GRAPHITE B

| |

I !

\ _4— BARE INDIUM
e
/<CADMIUM FRACTION
/ / .
\ \-\)\_()/
CADMIUM COVERED INDIUM \

0 5 10 15 20 25 30 35 40

10

05

CADMIUM FRACTION
-55-

beryllium surrounding the fuel with the remainder of the
reflector graphite, the source neutron multiplication was
quadrupled The assembly was not made critical

TABLE V-1

Radial Neutron Flux Traverse in Mid-Plane of Reactor

 

Distance
From Axas Relative Activaty
in Inches Bare Cd Covered
00 14 118 7 082
30 12 Til 6 903
L 5 10,385 5 862
60 5 003 3 T96
T5 3 Thk 3 197
90 3 889 3 264
10 5 4 291 3 274
12 0 7 604 4 194
13 5 8 835 4 164
15 0 7 822 3 197
18 0 5 926 1 okh
20 9 4 360 1176
2k O 3 045 670
27 O 1 891 356
30 0 1 138 179
36 0 126 017
-56-

V‘Ihnmfllé

A Assenbly Loading

Graphite has be@n suggested as a2 high temperature fluoride container
for the fissioning volume of the RMR CA-l3 was assembled to obtain
a check on some of the characteristics of such a reactor A two
region reactor was constructed with a l/l graphite to powdered fuel
volume ratio in the central core and surrounded by & 12" beryllium
and 6" graphite reflector as shown in Figure VI-l The powdered
fuel of CA-12 was used, and the reactor was first critical with

15 2 kg U-235 As additional fuel was added to obtain the configu-
ration shown in Figure VI-1l, stainless steel plates O 059" thick
were added around the fuel to take up the excess reactivity The
reactor contained 17 2 kg of U=235 ahd 11 O kg of stainless steel
when loaded as shown in the figure The core shell covered 9%

of the central fuel region (that is, 9" on either side of the md-
plene) which contained 684 of the fuel

B Neutron Flux Distribution

 

With the stainless steel core shell in place, radial bare and
cadmium covered indium traverses were taken in mid-plane of the
reactor extending from M-12 to W=12 The data are shown in Table
VI-1l and Figure VI-2 The associated cadmium fraction is also shown
on the figure

TABLE VI-1

Radial Flux Distribution in the Mid-Plane

 

 

Distance From Relative Activity
Axis in Inches Bare Indium Cd Covered
00 0 9628 0 8852
15 0 9628 0 9032
30 1 036 0 9184
6 0 1 215 1 085
T5 1 336 -
90 1 51k 1 330
10 5 2 200 1 515
12 0 3 502 1 82
13 L 3 738 1 561
16 5 2 T18 0 7605
22 5 0 7759 0 0776
28 5 0 0815 0 0072
20

21

<>
ORNL—-LR—DWG 2591

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

b E F G H J K L M N O P Q@ R s T
NUMBERS IN REFLECTOR
1313 |13 113 113 SQUARES INDICATE NUMBER
OF INCHES OF Be IN CELL
13 | 13 | 18 |18 | 18 |18 | 18 | 13 |13 BEHIND WHICH IS ENOUGH
GRAPHITE TO FILL 301n
TA
13| 18 | 18 | 18 | 24 | 24 | 24 |18 | 18 | 18 | 13 OF THE TABLE
13 | 1318 | 24| 36|27 |36 |27 |27 |24 |18 |13 |13
13 | 18 | 24 | 24 36 | 24 | 24 | 18 | 13
13 | 18 | 18 | 24 | 21 21 | 24 |18 [ 18 | 13
HoLE | 13|18 | 24| 27 27 | 24 |18 | 13
T I
& T4
& | | 1318|2427 27 |24 |18 | 13
w o
c | €
o |3 | 13]18|24]27 27 | 24 |18 | 13
13| 18 | 18 | 24 | 21 21 124 |18 |18 | 13
13 | 18 | 24 | 24 |2f ST 24| 24 18 |13
13 | 13| 18| 24 27 |27 |27 |27 36 |24 |18 | 13 | 13
13| 18 | 18 | 18 |24 |36 |24 | 18 | 18 | 18 | 13
9in FUEL TUBE + 21 in Be
13 | 13| 18 |18 | 18 | 18 | 18 | 13 | 13 o FUEL TUBE

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

|

+ 5in GRAPHITE

GRAPHITE (SAME LENGTH
AS LONGEST FUEL TUBE IN
THE SAME CELL)

 

 

Fig VI—1 Assembly Loading at Mid-plane
86

RELATIVE ACTIVITY

ORNL LR DWG 2592

 

BERYLLIUM

 

A

 

|
|
— e GRAPHITE
!
|

_ ey
I

 

 

 

BARE INDIUM ———_ |

 

/

fi/

-

\ CADMIUM FRACTION

05

 

 

.-—-—'.

/ /
CADMIUM COVERED INDIUM

 

 

 

 

 

S~

 

 

 

wn
Q

15
RADIAL DISTANCE FROM AXIS {

}

20 25

Fig VI 2 Radial Neutron Ftux Distribution in Mid plane

 

30

CADMIUM FRACTION
-59-

¢ Power Distribution

 

Power distributions were taken radially in the mid-plare from
M-12 to P-12 and axially back through M-12 The traverses were
cbtained by placing & 0.002" uranium fuel disk and its associated
aluminum catcher foil adjacent to a fuel tube The data obtained
for both traverses are shown in Tables VI-Z and VI-3 and Figures
VI-3 and VI=k

TABLE VI-2_

Radial Power Distribution in Mid-Plane

 

 

Distance From Axis Relative
in Inches Activaty
00 9 66
30 9 9k
60 13 09
9.0 2l 23
10 3 Ly 14
TABLE VI=3

Axial Power Distribution Through M-12

 

Distance From Mid-Plene Relative
in Inches Activity
10 9 66
51 8 95
90 9 18
13 0 10 3k
17 O 25 31

Three uranium foils were covered with O 020" of cadmium

and placed in radial positions in the mid-pleane The fuel
cadmium fraction, as determined from the activity of these
foils together with the activities of similarly placed foils
for the radial power distribution, are plotted as a function
of the square of the radial distance from the reactor axis 1in
Figure VI-5 Teble VI-lb shows the data from which the curve
was constructed This gives an average of very nearly 50%
thermal fissions in the mid-plane of the reactor.
09

RELATIVE ACTIVITY

OR

LR DWG 2593

 

50

 

40

 

30

 

20

 

 

40(

N

 

 

 

 

 

 

 

 

4 G
RADIAL DISTANCE FROM AXIS ( )

Fg VI 3 Radal Power Dstribution in Mid plane
¥9

RELATIVE ACTIVITY

50

40

30

20

ORNL

DWG 2594

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

6 8 10
DISTANCE FROM MID PLANE { )

Fig VI 4 Axial Power Distribution Through M 42
29

FUEL CADMIUM FRACTION

R D!I! 2595

 

 

o8

AVERAGE

=50/

 

o6

 

 

020

 

 

 

 

 

 

 

 

20

40

60 80
(DISTANCE FROM aXIS ( )2

Fg VI 5 Radial Fuel Cadmium F actions

120
-63-

TABLE VI-4

Fuel Cadmivm Fractions in the Mad-Plane

Distance From Cadmium
Axis in Inches Fraction
00 0 200
60 0 399
10 3 0 813
ViI CA=-14

A Criticel Loadings

The final assembly was bullt up with e geometry similar to that
used in CA-1l1l The aim of the experiment was to insert materiels
in the reactor that would simulate the structural polsons that
would be present in & full scale reactor in the power range of

50 to 200 megawatts ‘The assembly was loaded as shown In Figures
VII-1 and VII-2 and conteined 18 2 kg of U-235 in the 6.42 gnm
U-235/cc fuel mixture described in Section V In general the
reactor was maintained criticel with two dlfferent sets of poilsons
The first set of polsons included

1 Boral sheets (9" x 2-7/8" x 1/4") which surrounded the fuel end
ducts on the stationary table Thas Boral extended from the end
of the assembly to a point 9" from the central fuel region, thus
4-1/2" of the end ducts were covered Boral sheets (1-1/4" x
2-7/8" x 1/4") were placed around the fuel end ducts on the
moveble table These sheets extended from the end of the fuel
ducts to a point 7-3/4" from the central fuel region

2 Stainless steel sheets (0 146" thick) covered 96% of the inside
fuel surface The outslide fuel surface was uncovered from the
1nterface of the assembly back 4-1/2" on the stationsry teble
The fuel surface exposed at the assenmbly interface and the
corresponding surface at the opposite end of the central fuel
region were left uncovered The remaining outside fuel surface
was 96% covered with O 118" thick stainless steel sheets The
total weight of the stainless steel was 25 5 kg
_bg_

<o ® ~ M o B

F G H

L LR DWG 2596

| J K L M N 0 P Q R S T U Vv w X

 

%%@%fi%%%fl%fi%

 

 

o] .

.

 

 

%7

 

%

OO ///////4

 

 

 

 

NN 7

 

 

 

N BERYLLIUM AN < 7 Z
NN NN

 

SN,

 

 

777777

 

 

 

 

 

 

 

 

 

 

 

 

 

 

\ //
A // 7

 

 

 

 

 

 

.

 

 

 

 

 

 

 

o\
Q0N
N\ \5\\\\\?\/}%/(/«%//// Y
DN NN N NN
NS SN NN
NN NN
1o NN\ // NN 7
> oD\~
L1 \\i\\ SN
. SN
o
@//////%%///// ////Z// .
/

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Fig VII-1 Assembly Loading at Mid-plane (No Structure Shown)

 
_99_

GRAPHITE

BERYLLIUM

NICKEL SHEET NICKEL SHEET

NICKEL SHEET

NICKEL SHEET NICKEL SHEET

MOVABLE TABLE STATIONARY TABLE

Fig VII-2 Axial Assembly Loading in Vertical Plane Containing Reactor Axis

L
ORNL-LR DWG 2597

 
-66-

3 TNickel sheets 2-7/8" wide and 0 01¥ thick were in the
positions indicated in Figures VII=-2 ahd VII-3

In addition to the above poisons, 177 1n5 of beryllium were added
to the fuel region Of this total, 150 inJ were distributed
symmetrically throughout the outer three fuel layers frq% 0 to 18
inches back on the stationary teble An additional 6 in” were
distributed symmetrically throughout the inner 4-1/2" long fuel
layer adjacent to the assembly interface on the stationary table
The remaining 24 inJ of beryllium were symmetrically distributed
throughout the fuel region on the movable table

During this series of experiments reactivity wvalues were obtained
for one sheet of nickel (24" x 2-7/8" x 0 010") placed in a
horizontal position across the top of cell K-12, and for the
extension of the reflector from 9" to 32" of beryllium on the
sides of the reactor The insertion of the nickel sheet resulted
in a loss in reactivity of 8 cents The replacement of graphite
by beryllimm in cells G-10 through 15 and cells V-10 through 15
and the addition of graphite to cells E-1C through 15 and X-10
through 15 resulted in a gain in reactivity of 9 6 cents

The second set of poison distributions involved the following
changes

1 The entire fuel surface was 95% covered by 27 8 kg of
stainless steel

2 All the Boral around the end ducts was removed

3 The rest of the assembly was as described previously

The remaining measurements described were obtained with the
reactor containing this second set of poison distributions

Importance Function of Three Stainless Steel Rods

The loss in reactivity due to three 316 stainless steel rods
vas determined at three radial positions in the reflector and
at one position in the island The rods were 3/16" in diameter,
and 36" long and each weighed 132 gn They were inserted as
skewers in the various cells and extended through the entire
shish on the stationary table The date obtained are shown in
Table VII-1
10

114

12

13

14

15

ORNL-LR-DWG 2598

 

 

 

 

 

 

—— NICKEL
=

SHEET

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

NICKEL SHEET —]

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Fig VII-3 Position of Nickel Sheets on the Stationary

and Movable Tables

-67-
-68-

 

TABLE VII-1

Tmportance Function of Three Stainless Steel Rods

 

 

 

Average Distance of Rods Loss 1n Reactivity
From Axis in Inches in Cents
23 23 1
10 8 17 6
13 7 12 O
16 7 60

C Thermal Neutron lLeakage

Thermal neutron leakage was determined at the end and side of the
reactor, both with and without Boral shielding sround the end fuel
ducts These meagurements were taken with 0, 1/8", 1/4", and 3/8"
thicknesses of Boral swrrounding the U-235 fission chamber that
was used

The side leakage was measured with the U-2%5 fission chamber
located in cell D-12 The geometrical center of the chamber
rested 12" back of the interface on the stationary table or 3"
back of the mid-plane of the reactor The axlis of the chamber

was approximately 1-1/2" from the edge of the graphite side
reflector and was parallel to the reactor axis The Boral layers
surrounding the chamber were in the form of concentric cylindrical
sleeves that extended from the interface of the stationary table
back 24" The inside diameter of the inner most layer was 2", and
the outside diameter of the outer most layer, using 3 layers, was
about 2-3/4" The ends of the Boral sleeves were not capped

The end leakage was determined with the fission chamber placed
5—1/2" back of the end graphite reflector The axis of the
chamber was parallel to the floor and perpendicular to the axis

of the reactor, which passed through the geometrical center of the
chanmber The data obtained are shown in Table VII-2 and Figure
VII-% All measurements are shown relative to the side leakage
without any Boral shielding around the chamber or in the reactor

D Power Distribution

A radial power distribution was taken in the mid-plane across the
fuel annulus (cells P-12 and Q-12) The traverse was obtained by
Placing uranium and aluminum foils between the vertically adjacent
fuel tubes in the cells The data obtained are shown in Table
VII-3 and Figure VII-5
THERMAL FLUX (arbitrary units)

01

0 Ot

O o1

0 0001

S
ORNL-LR—0WG 2599

BACK LEAKAGE

BACK (WITH BORAL IN TABLE)

SIDE LEAKAGE

 

Ya s ¥ %
BORAL THICKNESS (in)

Fig VII—4 Thermal Flux Leakage vs Boral Thickness

5/8
L
ORNL-LR—-DWG 2600

 

 

 

N
o

—0)—
RELATIVE ACTIVITY

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

|
I
i
|
|
\
|
| 1
O O
0\—/
e
VOID VOID
- BERYLLIUM ==J| =< FUEL TUBES ——————|  |~————BERYLLIUM————3=
|
| | |
| ] .
| | |
|| |
2 3 4 5 6 7 8 9 10 T 12

DISTANCE FROM AXIS (in)

Fig VII-5 Power Distribution across Fuel Annulus at Mid-plane
-T1-

TABLE ViI-2

Thermal Neutron Leakage

 

Layers of 1/8" Boral¥* Counter Boral Around Relative
Arownd Counter Position End Ducts Actlvity
0 Side No 1 000
1 Side No 0 00635
2 Side No 0 00103
3 Side No 0 000676
0 End No o 777
1 End No 0 0466
2 End No 0,0332
3 End No 0 0301
b End No 0,0235
0o End Yes 0 352
1 End Yes 0 0197
2 End Yes 0 0160
3 End Yes 0 0125
4 End Yes 0 00992

%0 161 gm B/cm?/Layer of Boral present

TABLE VII-3

Power Distribution Across Fuel Annulus at Mid-plane

Distance From Axis

 

 

in Inches Relative Activity
kL 8 25 30
5 2 17 86
5 6 14 24
6.2 10 97
65 12 %0
T1 12 35
TP 12 63
7,8 14 87
81 19 32
8 L 28 3k
-T2-

APPENDIX "A"

Summery of Materiasls in Reactor Assemblies

 

CA-10
Average¥*
Mass Volume Mass Volume Density
Island kg em3 x 10™°  Fraction  Fraction gm/cm
Beryllium 48 1 26 0 911 918 1 699
2S Al Matrix L 5 16 085 058 0 158
24Ls Al Skewers 02 01 003 002 0 008
Stainless Steel
skewer - - - - 0 o2k
Void - 06 - 021 -

The island contained eleven aluminum and one steel skewer They were
placed in the outer island cells

Fuel Annulus (Includes beryllium region required to close annulus)

Sodium 52 4 52 5 486 T34 0 805
Uranium 16 1 09 149 012 0 265
0 161

304 Stainless
Steel Boxes 14 3 18 1%2 025 0 220
28 Al Matrix 11 1 5 4 103 075 0 158
24s Al Skewers 06 02 005 003 0 008
Stainless Steel

skewer - - - - 0 024
Al Ceps T4 27 068 038 -
Beryllium 60 33 056 oL5 1 699
Void - } 8 - 067 -

The lower uranium density reported refers to the fuel region behind
the island The higher figure refers to the remainder of the fuel
region

Reflector (Beryllium)

Beryllimm 1521 6 822 5 91k 918 1 699
25 Al Matraix 141 6 52 1 085 058 0 158
24S Al Skewers 09 03 - - 0 008
Stainless Steel

skewers 08 01 - - 0 025
Void - 20 5 - 023 -
CA-10 = Continued

-T3-

 

Average¥
Mass Volume Mass Volume Density
kg em’ x 103  Fraction Fraction gm/cm
Reflector (Graphite)
]
Grephite . ‘1 1673 5 o7k 7 907 898 1 542
2S Al Matrix 171 6 63 1 093 058 0 158
Void - W7 7 - ol -

* The average dansities reported are for a unit cell (9"2 cross-
sectionsal areg) in which the material was actually present
-Th-

CA-11
Average
Mass Volume Mass Volume Density
kg cmd x 1070 Fraction Fraction gm[cm

Island (Includes Be Region enclosed by end ducts)
—_———

Beryllium 57 1 30 9 912 918 1 699
25 Al matrix 53 20 085 058 0 158
24s Al skewvers 02 01 003 002 0 008
Voad - o7 - 021 -

The island contained 16 skewers, which were placed in the outer
beryllium shishes (9" length)

Fuel Annulus (Includes end ducts)

Fuel Mixture

Including Usrnium 6L 4 35 0 620 527 -
Uranium 8 2 - 079 - 0 123
UF)y, - - - - 0 163
Zr0o - - - - 0 553
NaF - - - - 0 200
C - - - - 0 088
2S Al Matrix 10 5 39 098 058 0 158
Al Fuel Tubes

and Spacers 28 6 10 5 275 159 0 443
Stainless Steel

Nuts 0 8 01 007 001 -
Veoad - 16 9 - 254 -

The weight of the fuel mixture includes 360 grams of uranium present
as disks
Each fuel tube had a stainless steel nut on both ends

Reflector (Beryllium)

Beryllium 1518 6 820 9 o1 918 1 699
28 Al Matrax 141 3 52 0 085 058 0 158
243 Al Skewers 08 03 - - 0 008
Stainless Steel

skewers 08 01 - - 0 024
Void - 20 5 - 023 -

Reflector (Graphite)
e e

Graphite 1626 677 o474 906 898 1 542
2S Al Matrix 166 831 61 3 093 058 0 158
-75-

 

 

CA-12 Variation I (Graphite island and reflector)
Average
Mass olume Mass Volume Density
kg em”? x 10°2  Fraction Fraction gm/cm)
Island (Includes region enclosed by fuel annulus)
Graphite 180 1 10k 9 903 898 1 542
2S5 Al Matrix 18 5 6 8 093 058 0 158
Stainless Steel
skewers 0 8 01l 00k 001 0 024
Void - 50 - oL3 -
Fuel Annulus
Fuel Mixture 75 3 40 5 622 545 -
(Includang Uranium)
Uranium 18 5 - 153 - 0 249
UF) - - - - 0 329
Zr02 - - - - 0 hh9
NaF - - - - 0 163
C - - - - 0 072
Al Fuel Tubes 3% 0 121 273 163 0 M3
2S5 Al Matrix 11 8 L 3 Q97 058 0 158
Stainless Steel
Nuts 10 01 008 002 -
Void - - - 232 -
Reflector
ettt
Grephite 6738 5 3965 1 906 900 1 542
28 Al Matrix 696 8 256 2 09k 058 0 158
Stainless Steel
Skewers 08 01 - - 0 024
Void - 185 &4 - ok2 -

The outer 128 cells were filled with AGHT Graphite in all CA-12
configurations All other graphite is AGOT

Variation ITI (Combination graph:ite and beryllium
18land and reflector)

Island
Beryllium 82 2 Ll L 515 L72 1 699
Graphite 61 8 36 1 387 383 1 542
28 Al Matrix 14 g 55 093 058 0 158
Stainless Steel

skewers o8 01 005 001 0.024
Void - 80 - 085 -
-T6-

CA-12 - continued

 

Average
Mass Yolume Mass Volunme Density
kg em® x 10°7 TFraction Fraction gm/mm3
Fuel Annulus
Fuel Mixture 82 0 bWy 1 608 415 -
Including Uranium
Uranium 19 6 - 146 - -
Al Fuel Tubes 35 0 12 9 260 121 0 443
25 Al Matrix 16 8 6 2 124 058 0 158
gtainless Steel
Nuts 11 01 008 001
Void - 42 9 - Lol
Reflector (Be and Graphite)
Beryllium 681 1 368 2 470 450 1 699
Graphite 639 T 372 6 hh] 456 1 542
2S Al Matrix 129 3 W7 5 089 058 0 158
Void - 29 3 - 036 -
Reflector (Graphite - includes only the region outside the Be
reflector)
Graphite 5674 k4 3345 6 906 900 1 542
2S Al Matrix 587 7 216 1 0%k 058 0 158
Void - 154 9 - oh2 -
Variation III - (Pwo Region)
Core
Fuel Mixture 75 3 40 5 622 545 -
Including Uranium
Uranivm 18 5 - 153 - -
Al Fuel Tubes 33 0 12 1 273 163 0 443
Al Matrix 11 8 4 3 097 058 0 158
Stainless Steel
Nuts 10 01 008 002 -
Vold - 19 % - 232 -
Reflector (Be + Graphite)
Beryllium 559 3 302 3 676 662 1 699
Al Matrix 72 2 26 5 087 058 0 158
Graphite 196 5 11k 4 237 251 1 542
Void - 13 3 - 029 -
CA-12 - Continued

-T7-

Average
Volume Mass Volume Density
cm> x 1073 Fraction TFraction g;m./cm3

 

Mass
kg

Reflector (Graphite)
Graphite 6362 0
Al Matrix 658 2
Void -
CA-13
Core
Tuel Mixture ™ 3
Including Uranium
Uranium 18 5
UFY -
Zr02 -
NaF -
C -
Graphite 120 7
Al Fuel Tubes 33 0
Al Metrix 2L 0
Stainless Steel
Wuts 10
Void -

Reflector (Beryllium)
o

Beryllium 1654 7
Al Matrix 154 0
Al Skewers o4
Stainless Steel

skewers 13
Void -

Reflector (Graphite)

Graphite 2071 4
Al Matrix 212 L
Void -

3Thé6 1 906 900 1 542
2hk2 0 094 058 0 158
174k 5 - oh2 -
4o 5 297 266 -

- 073 - -
- - - 0 164
- - - 0 224
- - - 0 081
- - - 0 036
70 3 L75 L63 0 771
12 1 130 080 0 221
8 8 094 058 0 158
01 ook 001 -
20 0 - 131 -

8oL L o1k 918 1 699

56 6 085 058 0 158
02 - - 0 008
02 001 - O 024

22 4 - 023 -

1206 4 Q07 898 1 542

78 1 03L 058 0 158

59 0 - oll -
~78-

 

 

 

CA-1k4
Island (Includes Be Region enclosed by End Ducts)
Average
Mass Volume Mass Volume Density
kg cmd x 103 Fraction TFraction gm/cm
Beryllium 57 1 30 9 905 918 1 699
Al Matrix 5 3 20 084 058 0 158
Nickel 05 01 008 002 0 028
Al Skewers 02 01l 003 002 0 008
Void - 0T - 020 -
Fuel Annulus (Including End Ducts)
Fuel Mixture 81 3 4z 7 596 Sll -
Including Uranium
Uranium 19 6 - 143 - 0 249
Al Fuel Tubes 36 0 13 2 264 165 0 k43
Aluminum Matrix 12 7 Y 7 093 058 0 158
Beryllium 5 4 29 039 ,036 -
Stainless Steel
Tuts 11 01 008 002 -
Void - 157 - 195 -
Reflector (Beryllium)
Beryllium 1559 0 82 7 913 918 1 699
Al Matrix 145 1 53 3 085 058 0 158
Nickel 18 2 001 0002 0 028
Stainless Steel
Skewers 0 8 1 000k 0001 0 024
Al Skewers 08 3 0005 0003 0 008
Void - 20 9 - 023 -
Reflector (Graphite)
Graphite 1255 1 731 0 906 898 1 542
Al Matrix 128 7 W7 3 093 058 0 158
Nickel 12 1 001 0 028

Void - 35 6 - oflu -
APPENDIX B
Analysis of Materials in Weight Per Cent

 

Material Assembly Ag Al B Ba Be Ca Cd Co Cr Cu Fe Mg Mn Mo Na N Pb Si 3n Sr Ta Ti v W Zn Zr

Al Matrix All wt% <0 04 02 <0 01 <004 0 001 0 08 <0 08 015 015 05 02 03 <008 <10 03 <0 08 25 <0 08T <01 02 <0 08 <03 <015
Beryllium All wt % 0 05 <0 02 <01 <005 <005 <0O05 005 <0O0IT <001 <002 <005 <01 <0 05 <0 02 <0 02 <02 <01
Stainless Steel

(Core Shells} 10 11 wt % <0 05 <0 02 <0001 <01 05 03 <0 O1 1 < 02 <01 2 < 02 0 05 <02 <01
Nickel Sheet 14 wt % <0 05 <0 02 <0001 <01 02 03 2 01 02 < 02 <0 1 01 <0 02 <0 02 <02 <01
Nickel Sample 11 wt % <0 05 <0 02 <0001 <01 ] 0 05 03 ] 0 05 04 <0 02 <01 02 0 05 <0 02 <02 <01
Inconel Sample 11 wi % 03 <0 02 <0 001 <01 1 03 0 02 1 <0 02 <0 1 1 03 0 05 <02 <01
Bismuth Sample 10 N wt % <0 05 <0 02 <0001 <01 <005 <005 <005 <0O05 005 <001 <0 01 <0 02 <0 05 005 <005 <002 <0 02 <0 02 V2 <01
Lead Sample (1 ) 10 11 wt% P <0 05 <0 02 <0 001 <01 <005 <005 <005 02 01 <0 0] <0 01 <0 02 <0 05 <0 05 <0 02 <0 02 <0 02 <02 <01
Lead Sample (1/4 ) 11 wt% P <0 05 <002 <0001 <01 <005 <005 <005 03 01 <0 01 <0 01 <0 05 <0 05 <0 02 <0 02 <0 02 02 <01
310 Stainless Steel

Sample 11 wt% <004 <004 <0004 <002 <QOO0003 <08 <004 <063 >100 015 >10 <0 02T <0 643 031 <063 >0 <0 08 063 <004T <006 <1 3 <0 005 008 <13 <031 <015

Sodium 10 PPM 80 <25 <10 75 2T <10