ORNL-1634.txt

 

 

 

ENTRAL RESEARVE LN |
; DOCUMEN? COLLECTION | ORNL 1634 S0 4
Al Reactors=Research and Power

TINMARIETTA ENERAGY S5YSTEMS LIBRARIES

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PRELIMINARY CRITICAL ASSEMBLY FOR
THE AIRCRAFT REACTOR EXPERIMENT

Dixon Callihan
Dunlap Scott

 
 
    
   
    

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OAK RIDGE NATIONAL LABORATORY
1 OPERATED BY - -
CARBIDE AND CARBON CHEMICALS COMPANY
A DIVISION OF Qfll'ph'i‘rc_ugmq‘t’fnu CARBON CORPORATION | .
©  POST OFFICE BOX P .
' .OAK RIDGE. TENNKSSEE

 
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ORNL 1634

This doéument contains 59 pages
B 1s copy 30 of 155, Series A.
CLassI™aTiny CHANGED To-
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Ject Category: Reactors-Research

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Vi el T e S “and Power.
£ a&aaaaaaezza___;2:5:él¢24£ZJ,

PRELIMINARY CRITICAL ASSEMBLY
for the
ATRCRAFT REACTOR EXFPERIMENT

Work by: Dixon Callihan
J. F, Ellis (Now in U. S. Army)
E. V. Haake (Now at Consolidated Vultee
Aircraft Corp.)
J. J. Lynn
E. R. Rohrer
Dunlap Scott
D, V. P, Williams

Preparation by: Dunlap Scott
Dixon Callihan

DATE ISSUED
0CT 238 1953

PHYSICS DIVISION

A, H, Snell
Director

Contract No. W-7k05, Eng. 26
OAK RIDGE NATIONAL LABORATORY
Operated by
CARBIDE AND CARBON CHEMICALS CCMPANY
A Division of Union Carbide and Carbon Corporation
Post Office Box P
Oak Ridge, Tennessee NAARIET T ENERY SYSTEMS LERARES

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ORNL 1634
Reactors-Research and Power

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ABSTRACT

 

A zero power mock-up of the Aircraft Reactor Experiment or ARE was
constructed in the Oak Ridge National Laboratory Critical Facility. The
asgembly was BeO moderated and reflected and used a powder mixture of Zr,
Ne.,, and enriched uranium, simulating the reactor fuel, packed In stainless
steel tubes as fuel elements. The clean critical mass, without the ARE
regulating and safety rods, was 5.8 kg of U-235 to be compared to the
predicted value of 5.5 kg. The addition of the regulating rod guide assembly
at the center of the reactor increased the critical mass to 6.68 kg U-235.
The value of the ARE regulating rod was 125¢ or 1-1/h times the effective
delayed neutron fraction for this reactor. The calibration of one ARE
safety rod gave a value of approximately 550¢.

Neutron flux distributions vwere meagsured by comparing bare indium and
cadmium covered indium foll activations at various points in the reactor.
The power distribution was found by measuring the fisslon fragment activity
on aluminum catcher folls placed in contact with a fused uranium bearing salt
contained in an Inconel tube of ARE specifications placed at various polnts
in the reactor. The spatial distribution of those neutrons capable of pro-
ducing fission was measured in the moderator and reflector using the same
catcher foll method with a metallic uranium digk.

The reactivity contributlions by a number of materials, primarily con-
stituents of the ARE structural componentis, were measured. The rasdial im-
portances of fuel and fuel containers were determined by the reactivity
coefficient method as was the contribution of a partial mock-up of a steel
pressure shell.

 
1I.

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IV,

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Abstract .
List of Figures
List of Tables -

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INTROTUCTION « = o

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OF CONTENTS

 

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DESCRIPTION OF CRITICAL ASSEMBLY

A,
B.
C.
D.

FIRST LOADING -

A,
B.
c.
D,
E,
F.
G.

SECOND LOADING

A.
B.
C.
D.
B,
F.
G.
BE.
I.
Jd.
K.

SUMMARY

Moderator and Reflector

Fuel and Coolant
Mechanical Equipment « « « o
Ingtruments and Power Interlocks

Critical Mass

Control Rod Calibration

¢ &

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¢

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Reactivity Value of Pressure Shell

Radial Importance of Fuel and Fuel Elemant Container
Reactivity Value of Reflector Coolant and Tubege .« »
Attempt to Calibrate ARE Regulating Rod
Meagurement of Neutron Flux

Critical Mass

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¢

*

® & o o o b

¢ o o o s a <@ o &0 o 8

. & . «

Regulating Rod Calibration « » « o
Fuel Tube Reactivity Coefficients

Reactivity Value of Reflector Coolant and Tubes
Reactivity Coefficlents
Evaluation of Fuel Tube Type Safety Rod

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a o ¢ & & B0 ¢ o

Reactivity Value of End Reflector -

Neutron Flux Distribution

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Fission Neutron Flux Distribution
Power Distribution é¢ o « oa o o
Evaluation of ARE Type Safety Rod -

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ACEKNOWLEDGEMENTS

APPENDIX

Analysis of Materials

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26,

LIST OF FIGURES

Top Surface of BeQ Columms o « o -

Fuel and Coolant Tubes with Be(Q Blocks

Critical Agsembly Structure . . . .
Details of Critical Assembly . . .
Loading Chart, First Loading . . .
Control Rod A Calibration » « » « »

Control Rod A Sengitivity . » « « o

Radial Importance of Fuel and Fuel Tube

o

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Indiu.m Tr&verse, Radial s 8 + ece @ O & @

Indim TI‘&'VGI‘SG, Axial * o « @ ¢ 8 0 ¢ .

*

Indium Traverse, Longitudinal at 8.38" Radius e« ¢ « « o o

Indium Traverse, Longitudinal at 15.88" Radius

Loading Ch&r‘b, Second. ].-Oad.ing ® & o e o e o . ¢ ¢« & & B8

ARE Regulating Rod Calibration e« o o o o « ¢« o o ¢« o o »

ARE Regulating Rod Sensitivity . .

Radial Importance, Stainless Steel Fuel Tube vs Void

Radial Importance, Low Density Fuel and Inconel Fuel Tube vs

s & © @ & & s 8 8 &

Stainless Fuel Tube o ¢ o ¢ o o o o o ¢ o«

. o * * . » - -

Indium Traverse, Longitudinal‘at 12.06" Radiug « » o o o » o

Indium Traverse, Radial at Midplane (Regulating Rod In) =« «

Indium Traverse, Radisl at Midplane (Regulating Rod ‘Out) « .

Fission Flux Distributlon, Radial at Midplane « + « « ¢ o o &

Fission Flux Distribution, Longitudinal at 10.09" Radius

Power Traverse, Longltudinal at Three Radll « o ¢ o o o o o o

Power Traverse, Radial at Five Elevations

- Indium Traverse, Longitudinal in Fuel « o o ¢ ¢ o ¢ o s o o @

Fuel Self-Shielding e e o % © © o ® & s 6 e ¢ 8 8 o o @ & @

-7-

Page

 

11
15
16
17
19

.21

22
2k
27
28
30
31
33
35
36
37

39
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45
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49
50
52
54
25
56
iI.
II1.
1V,
V.

VI.

VII.

VIIT.

I1X.

XL,

XII.

XIII.

XIvV.

XVII.

XVIIT.

Composition of Core and Reflector . . « ¢« v o &« &« + &

LIST OF TABLES

Reactivity Value of Fuel Tubes « + ¢« ¢ ¢ o ¢ o o & o

Radial Importance of Fuel and Fuel Container . . . . .

Neutron Flux Traverse Radial at Reactor Midplane . .

Neutron

Flux Traverse, Axial .« + ¢« ¢ o 4 ¢ ¢ o & « &

Neutron Flux Traverse, Longitudinal . . » « + « ¢ + &

Radial Importance of Stainless Steel Fuel Tube

-
-
-

Radial Importance of Fuel and Container Material . .

Reactivity Value of Fuel Tubes « ¢ o « ¢ o« ¢« o ¢ « &«

ReactiVity CoeffiCientS * ¢ 4 & e & o ® € s & e = e »

Summary of Fuel and Tube Reactivity Coefficients.. . . .

Neutron

Neutron

Neutron

Neutron

Fission

Fiesion

Flux Traverse, Longitudinal at 12.06" Radius. .

Flux Traverse, Redial at Midplane (Regulating Rod In)

Flux Traverse, Radial at Midplane (Regulating Rod Out)

Flux Traverse, longitudinal in Fuel . . . . .« . « . .

Neutron Flux Distribution, Radial at Midplane

Neutron Flux Distribution, Longitudinal at 10.08" Radius

Power Traverse, Longitudinal at Three Radii . . . . « . « . . .

8-

Page

 

10
18
23
26
29
29
38
38
40
41
k2
143
46
46
48
51
51
53
1. INTRODUCTION

The Aircraft Reactor Experiment being planned at Oak Ridge National
Laboratory is a high temperature, intermediate power reactor having a
beryllium oxide moderator and reflector and a liquid fuel-coolant. This
fuel is designed as a mixture of the fluorides of zirconium, sodium and
enriched uranium in the proportion required to meet the requirements at
the operating temperature and to include sufficient U-235 to insure a nuclear
chain reacting systeml. A series of experiments has been performed at room
temperature and essentially zero powsr on a mock-up of the reactor constructed
in the Oak Ridge Critical Experiments Facility. The main purpose of the ex-
periment was to provide a comparison of the experimental value of the critical
mass and of the power and neutron flux distributions with those predicted for
this mock-up. In addition, an evaluation was medé of the reactivity coeffi-
clents of the regulating and safety rods designed for the ARE and of samples
of certain ARE structural materials. The reflector and moderator used in
this critical assembly were the BeO blocks which had been prepared for the
ARE., The fuel was a mixture of powdered fluorides having essentially the
nuclear properties of the ARE fuel except those dependent upon density and
was packed in stainless steel tubes., A similar powder, but without uranium,
was contained 1n tubes located in the BeQ reflector to simulate the fluid
reflector coolant of the ARE.

The range of reactivity required to first build an essentially "clean"
Just critical system and then to evaluate the rather large amount of poison
in the centrally located ARE safety rod necessitated two fuel loadings of
significantly different uranium concentration. The experimental results,
for purposes of reporting, are divided between these two loadings.

1t was necessary to release the BeO for preparation of the ARE reactor
before the experiments reported here could be logically concluded. It
was not possible, for instance, to measure all of the ARE regulating and
safety systems simultaneously. Nor was it possible to evaluate the data
before disassembly of the equipment so doubtful results could not be re-
investigated. There were, for example, several neutron flux Vvalues which
were considered not representative and are not reported. The construction
and operating temperatures of the ARE will not permit a further study of the
microscopic nuclear characteristics of the reactor for which these experi-
ments were degigned.

No attempt at theoretical analysis has been made in this report. It
ig intended to present a description of the mock-up and the experlimental
procedure, where necessary, with a presentation of the data obtained.

II. DESCRIPTION OF CRITICAL ASSEMBLY

A. Moderator afid.Reflector

 

The core was a right cylinder, with ite axis vertical, 32.8" in
diameter and 35.6" in length and consisted of fuel tubes and hexagonal
faced BeO blocks. These blocks were approximately 6" long and 3-3/4" across
the hexagonal flats with a 1-1/4" diameter hole, parallel to the long axis,
through the center for the fuel tube. The average density of the BeO in the

1 "Reactor Program of the Aircraft Nuclear Propulsion Project”, ORNL 123k
June 2, 1952.
9-
blocks was 2.76 gm/cc. The BeO side reflector was 47.5" in diameter outside
and also 35.6" in length. Theae blocks had the same dimensions as the core
blocks except that the central hole was 1/2" in diamster. Some of the periph-
eral reflector blocks were cut to approximate a cylindrical outer surface.
There was no Be0O end reflector.

The BeO blocks had been hot pressed to shape with a rather close limit
on the hexagonal dimensions but a large permissible variation (X 0.1") in
the length. Although it was possible to make the top and bottom surfaces
of the asgsembly plane by selective stacking, the ends of the blocks, in
succeggive layers, were not, in general, coplanar. The uniform top sur-
face of the BeO and the ends of the fuel tubes are shown in Fig. 1. The
large hole, 2" in diameter, in the BeO block in the left foreground is for
one of the ARE fission chambers. Kerosene was used in the machining of the
blocks and all of it was not removed by a final wash with trichlorethylene
ag evidenced by the strong odor remaining. A quantitative measure of the
residue was not made. A chemical analysis of the BeO powder (taken prior to
pressing) is given in the appendix as sample #19.

A summary of the core constituents 1g given in Table I. The two welghts
of fuel mixture given are the contents of 70 fuel tubes in the first (#I) and

second (#II) loadings, respectively, with the second loading being used to
caelculate the weight percent compositiion.

TABLE I

COMPOSITION OF CORE AND REFLECTOR

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Core
Volume Volums Weight Welght
££3 % 1b
Be0 15.50 89.05 2660 89.26
Stainless
Steel 0.32 1.84 153.45 5.15
Fuel #I 165.79
(70 tubes) 1.L42 8.16 #IT 166.68 5.59
Voild 0.17 0.95 - -
Total 17.41 100,00 2980.1 lOO.bO
Reflector
Volume Volume Weight Weight
£43 % 1b %
BeO 18.09 Ok, 70 3104, 98.36
Inconel 0.0k43 0.23 22.86 0.72
Coolant 0.24 1.26 29.17 0.92
Void 0,60 3.13 - -
Void 2"
Instrument
Hole 0.13 0.68 -- —
Total 19.10 100,00 3156.0° 100..00

 

 

 

 

 

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Fig. 1. Top Surface of BeO Columns

L1
The BeO blocks were stacked on a 1" thick Inconel plate which, in turn,
was supported by elghteen Inconel "legs", 1" in diameter. The support plate
was machined to match the fusl and reflector coolant tube matrix. The blocks
were contained by a 47-1/2" diameter, 1/16" thick, open end Inconel can. Since
the reactor design prescribes a small clearance between adjacent BeO block
colums the can did not fit snugly and consequently the blocks were rather
loose. Subsequent tightening by shimming with aluminum strips between the can
and blocks gave only a small change in reactivity. (The dimensional change was
not measurable since the block shift was not uniform but the effect on the re-
activity was only about 9¢.) This tightening was done to reduce day to day
shifts in reactivity. :

B. PFuel and Coolant

 

The fuel mixture used in the critical assembly wag patterned after
the ARE fuel mix #27 which prescribed

50 mol % ZrF),
L6 mol % NaF
~4 mol % UF),

Since insufficient ZrF) was available at the time of the experiments, ZrO,
wag substituted and sufficient carbon, as powdered graphite, was added to
compensate for the difference between the thermal neutron scattering by the
mixtures.

The mixture used in the critical experimente consisted of

65 .6 wt % ZI‘OE

23.7 wt 9 NaF
10.5 wt % C
0.3 wt % Hpp

with UF) added to give the specified uranium density, For the initial loading
this density was prescribed by Mills® to be 0.1632 gm U-235/cc or a fuel
composition of:

12.4 wt % UF), (U enriched to 93,2% in U-235)
57.4 wt % Zroo
20.8 wt % HNaF

9.2 wt % C

0.26 wt % HpO

The average packed density of the fuel mixture was 1.87 gm/cc.

Since no theoretical prediction was made of the amount of uranium required
to provide sufficient reactivity to evaluate the ARE regulating and safety rods,
the fuel for the second loading was empirically selected to have a U-235 density
about 25% greater than the filrst one. The compositlon of the mixture prepared
was

2 Mills, C. B. and D. Scott, "The ARE Critical Experiment", Y-F10-108,
August 8, 1952,

-12.
16.2 wt % UF),

55.0 wt % ZrOp
19.9 wt % NaF
8.8 wt % C
0.25 wt % HyO

The uranium density was 0.2138 gm'U-235/cc and the average packed density
of the mixture was 1.88 gm/cc.

The ARE reflector coolant, at the time of the experiment, was defined
asg the same base mixture as the fuel but without the UFy,. The blend
actually used was

69 2 Wh % ZrOE

22,7 wt % NaF
9.1 wt % C

which differs from that above because of batching irregularities. A specto-
graphlc analysis of the materials used for these blends is given in the
Appendix, samples 10, 20, and 21,

A small quantity of material having properties more like those of the
liquid fuel was prepared from a ligquid mixture of ZrFu, NaF, and UF) by
casting it into cylindrical slugs. Its composition was

Th.0 wt % 2rF)

17.2 wt % NaF
8.8 wt % UFA

The bulk density of the solid was 3.75 gm/cc and the uranium density was

0.235 gm U-235/cc. These slugs were 1.06" in diameter, varied in length
between 1" and 5" and were of sufficient quantity to fill one fuel tube

when packed end to end. There were small volds between adjacent slugs which
reduced the loaded density to 3.54 gm/cc or 0.224 gm U-235/cc. Loaded in this
manner there was 114.2 gm U-235 in the 35.6" core length to be compared with
the average 123.6 gm U-235 in the packed fuel powder of the second loading.

The ARE fuel tubes are to be of Inconel 1.235" OD and 0.060" wall and
are connected above and below the Be0O to provide flow channels. Inconel was
not available at the time of this experiment, so type 302 stainless ateel
was substituted. Spectrographic analyeis of this stainless steel is given
in the appendix, sample 17. The tubes were 41-1/4" long overall with a fuel
filling length of 39.86". This provided 2.1/8" of fuel both above and below the
core to simulate the tube bends of the ARE. The tubes were sealed with threaded
stainless gteel caps 1-3/16" long, the threads being coated with Glyptal paint
to prevent powder leakage and water absorption. The tubes were packed by vibrat-
ing longitudinally as the specified amount of fuel mixture was added. X-ray
photographs taken of the filled tubes and examined on a densitometer indicated
a variation in the linear uranium density near the tube ends which was not
measured quantitatively. The method of loading gave less than 1% variation
in the overall density from tube to tubs.

The reflector coolant tubes were of Inconel, 1/2" outside diameter, 0.020"
wall thickness and 36" long, and were packed to an average density of 1.95 an/cc
by the same method as the fuel. They were sealed by welding small Inconel plugs

-13-
into the ends. These tubes occupied 66 holes in the Be0 reflector and
were centered vertically. The neutron source was located in the 67th hole
in the reflector. The two types of BeO blocks with the fuel and reflector
tubes are shown In Fig. 2

C. Mechanical Equipment

 

Figure 3 shows the reactor table, safety rod, and source drive support
structure. The control rod drives are shown mounted below the reactor
table. The location of the safety and control elements are indicated in
Fig. 4. The control elemenis were fuel tubes mounted on a lead screw driven
by a three-phase motor. The rod position indication was transmitted to the
panel in the control room by selsyn motors for the intermediate positions
and llmit switches for the extremes of travel.

Each safety rod consisted of seven fuel tubes arranged in a hexagonal
pattern mounted on a common base plate and supported by a cadmium lined rod
connected above the center tube. This array was drawn into the reactor by a
magnet mounted on a drive similar to that of the control rod, with selsyns
and limit switches for remote indication, the latter also beilng integral
with the Interlock system. Panel lights indicated the "in" position &f the
safety rods, the "out" position of the magnets, and contact of the magnets
and rods; selasyn-driven Veeder Root counters indicated intermediate positions
of the magnets. The two safety rods are shown in the "out" or least reactive
position in Fig. 4. In the normal or "in" position during operation the fuel
tubes of the safety rods were centered vertically in the reactor.

The Po-Be neutron -source was located in the reflector and was moved by a
remotely controlled motor between the midplane of the reactor and & paraffin-
boron carbide shileld mounted above the Be0. The source strength was about
b x 10 neutrons per second.

D. Instruments and Power Interlocks

 

The reactivity of the agsembly was regulated by positioning the control
and safety rods from the control panel and certaln interlocks were provided
In the system to avoid lmproper sequence of operationsg during start-up. It
was necegsary that the source be all the way in the reactor and the control
rods In their lesast reactive position before the safety rods could be raisged.
The control rods could be moved only after both safety rods were in contact
with their magnets, and one was completely raised and the other partially raised.
In the event that the BeO blocks of the array were dieplaced vertically the up-
ward motion of rods was stopped by interlocks.

Neutron and gamma sensitive detecting devices were placed about the
agsembly for measurement of the reactor power level with assoclated amplifiers,
scalers and recorders located in the control room. Two BF; proportional
counters supplied information to scalers through linear ambPlifiers for the
determination of neutron multiplication during the initial approach to critical.
A record of the relative power and period of the assembly was provided by a
log N amplifier. One D-C amplifier and two vibrating reed electrometers fed
by BFy ionization chambers and a ¥=sensitive scintillation detector provided
additionel relative power level Information. ZXach of these last four instru-
ment channels contained relays which were used to actuate the safety system.

The purpose of the safety devices was to shut down the reactor quickly
in the event of unusual rises in reactivity. Actuation of the safety system.

~1h-
 

 

-

RESTRICTED
Y=12 PHOTO 11700

 

Fig. 2. Fuel and Coolant Tubes with BeO Blocks

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Fig. 3. Critical Assembly Structure

16

 
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o

 

 

 

 

            

 

 

 

 

 

 

 

 

 

[ o A O RS T R S SN
@ PN ALY S LA L
290000092020 b S e
5 %<0 »(0 y~0 »<® )0 )~ > o e e
1 w g 2 K SOENISSUNSISISSSSOSNANSN
. M a W_M_ Sl XL L LY LA gy
(0 y<0)y~§y—<0)~0)<=) 2255 ¢ e ————
T / &, R O S SR RSN SRS S
P IINIIIIIN I A T .

 

-

 

47.5

T L7 Ll XL L L L LY L LLL AL
SO R R RSO RN S SR N ERNENAN

1
S S e N e N S S NN NI N RN N

 

 

 

 

 

 

 

a
=
xS
b
<L

 

 

 

s

Bl
FUEL TUSBE
SUPPORT
PLATE (AL}

 

133HS
3NGONI &

 

 

FIGURE 4. DETAILS OF CRITICAL ASSEMBLY

 

 

 

Y0001 rwamu
L 8 — Sl

 

 

 

 

 

SAFETY
ROD DRIVE
INCONEL
JL_. PLATE
i
T

8
T 5E

 

<

 
either manually or by instruments at a preset radiation level, de-energized
the magnets allowing the sbfety rods to drop thereby removing fuel and adding
poison to the reactor. Also the control rods were driven down removing addl-
tional fuel. Another magnet releass, actuated by instruments and operating at
a slightly higher flux level, dropped the safety rods in case of failure of
the lower level trip.

ITI. FIRST LOADING

 

A, Critical Mass

Successive increments of fuel were added, starting at the center of the
agsembly by first loading the safety and control rods, until the system be-
came critical, The reflector coolant tubeg were in place at this time. The
aystem was f£irst critical with a loading of 5.85 kg U-235, contained in 62
fuel tubes, and with the single fuel tube of control rod A almost completely
inserted. (The remaining eight fuel tube positions were empty.) With Rod A
completely inserted the excess reactivity corresponded to about 18¢, or 18%
of that due to the effective delayed
neutrons--the evaluation being made
jn the manner described in Section TABLE TII
III-B below. Figure 5 shows the load-
ing arrangemsnt for the critical system*-

 

 

 

It was shown in a similar way that the Reactivity vValue of Fuel Tubes
excess reactivity, when the assembly Position | Value (&)
was loaded with a full complement of
seventy fuel tubes, was 268¢ or about 11-3 64 .6
2% in Ak/k. The contribution to this 6-11
excess reactivity by individual pairs
of tubes was also measured, allowing 10-11 65 .7
considerable latitude in the loading, 523
so changes could be made from time to
time throughout this series of exper- 13-5 62.0
iments. Table II gives the reactivity 3-5
value and position of these tubes.

6-3 56.1
B. Control Rod Calibration 5.10

 

 

 

 

 

The control rods were calibrated
by the method described essentially
by Gladstone and Edlund3 in which
a measurement is made of the period resulting from a change 1n reactivity
produced here by the displacement of a rod from its position when the assembly
ig critical. This period is related to the change in reactivity by the well

 

known inhour formula: 5
;Ej 8 7 S
e~ I+ Al ;
63::' éig' X200 For f%;§_6% <</
[.".‘

2 8
(=1

 

*  The method of position-designation, which will be used throughout this N
report, is illustrated, for example, by 11-3 - the 11th row from the botitom
of Fig. 5, third block from the left,

3 Gladstone, S. and M. C. Edlund, "The Elements of Nuclear Reactor Theory"
(Van Nostrand, New York, 1952), p. 302.

-18-
 
Where
the change in reactivity in cents
the fraction of fission neutrons in the ith delayed group

the corresponding decay constant
the reactor period resulting from the reactivity change

produced by the control rod displacement

2w
1t 1o

Since this reactor assembly contained a large quantity of beryllium,
recognition must be made of the production of photoneutrons by fission product
gamma radistion which may be considered as equivalent to additional groups
of delayed neutrons. Their immediate effect is to make the control rod cali-
bration sensitive to previous reactor operation. An attempt to minimize the
error was made by the following procedurs. With the reactor critical at some
lower power an increment of reactivity was added by moving & rod other than
the one bging calibrated and the resulting reactor period was allowed to
stabilize. The assembly was then brought back to critical, at a power an
order of magnitude greater than the lower level, by adjusting the rod being
calibrated. The incremental change in the control rod position corresponded,
therefore, to the period with which the power had increased. The power was
then returned to the lower level without disturbing the rod being calibrated
and the procedure repeated, using successive increments of the rod to override
the corresponding periods. Summation of these increments gives the value of
any section of the rod. It is believed that the above method minimizes the
error due to the photo-neutrons since the rod increments are taken at the upper
limit of the range of operating power levels where the ratio of fission neutrons
to pseudo-source (Y-n) neutrons from previous high level operation is a maximum,

The effective total delayed neutron fraction for this reactor was cal-
culated”® to be~0.8% which includes the increased importance due to their lower
energy of emission but does not include the photoneutrons from the beryllium.

. Figure 6 ig the calibration curve of control rod A, consisting of one
fuel tube, located in position §9-7 shown in the loading plan, Fig. 5. In the
zero position the fuel tube is centered in the reactor X, In Fig. 7 1is plotted
the sensitivity of the rod, in cents/inch, against the rod position. These
data were obtained by dividing the incremental displacement of the rod into
the resultant change in reactivity. The lengths of the horizontal lines show
the displacements and their ordinates give the corresponding average sensi-
tivities. It is to be recalled that while rod A was being withdrawn during
calibration, a second nearby rod consisting of two fuel tubes, (Control Rod B),
was being inserted simultaneously, but in smaller increments. It so happened,
from the experimental arrangement, that the ends of the rods passed each other
when A was withdrawn about 18". There was possibly some disturbance of the
flux distribution when the rod ends with their stainless steel end caps were
proximate, a probable cause of the dip in the sensitivity curve near the center
of the range of calibration.

 

¥ The magnitude and stability of this period was measured on a previously cali-
brated logarithmic neutron level recorder. The period at the higher end of
the power range was stable within the limlts of measurement.

*% The ordinate of the calibration curve may be interpreted as the k-excess
available in the critical reactor for a given rod position.

4 €. B. Mills, private communication.

-20-
BWG. !'!17

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

100 | ]
CONTROL ROD A CALIBRATION
80 /
ZERO POSITION CORRESPOND /
CONTROL ROD IN NORMAL FUEL
TUBE POSITION /
70 v
3 60 e
W
3
£ 50 7 —
a /
O
® 49
30
20 /
'0 ///
L/—/
oo 2 4 8 10 12 14 16 I8 20 22 24 26 28 30 32 34 36

ROD POSITION (INCHES)
FIGURE 6

12
ROD SENSITIVITY ¢/ INCH

N

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

ROD POSITION (INCHES)
FIGURE 7

DWG., 21518
/{ [ X CONTROL ROD A SENSITIVITY
f 8
i 3
7 X
f \
‘7/ N
\\
8 10 12 14 16 18 20 22 24 26 28 30 32 34 36

¢z
C. Reactivity Value of Pressure Shell

 

Cylindrical sheets of mild steel were added outside the Inconel can,
encircling the BeQ reflector of this preliminary assembly and extending 6"
above and below it, in an attempt to estimate the contribution to the re-
activity of the ARE by its pressure shell. The addition of a sheet 1/4"
thick Increased the reactivity 34.5¢; a second l/h" thick laysr resulted
in an additional gain of 21.0¢ or a total increase of 55.5¢ for the 1/2"
thick layer. Because of time limitations and since further additions to
the thickness would have contributed successively smaller increments of
reactivity, the measurements were not extended beyond the 1/2" value.

D. Radial Impoftance of Fuel and Fuel Element Container

 

The contribution to the total reactivity of single fuel elements was
measured as a function of the radial position of the element. These measure-
ments were incorporated with an evaluation of reactor poisoning by the stain-
less steel container of the fuel, also as a function of radius, in the follow-
ing manner. The system was made critical under sach of three loading con-
ditione in five locations along a radius. In the first of these three
conditions the fuel element position was empty, in the second it was filled
with the standard stainless tube containing the fuel mixture. For the third
meagsuwrement an aluminum tube of like dimensions and filled with the fuel
mixture was substituted for the steel one. The resgults, referred to the
reactivity with a void in the test position, are given in Table III and
plotted in Fig. 8

TABIE ITI

 

Radial Importance of Fuel and Fuel Conteiner

 

 

 

 

 

Tncrease in Reactivity ¢
Stainless Steel | Aluminum Fuel
Position Fuel Tube Tube
8-7 107.2 187.0
9.8 10L,8 173.0
10-8 79.8 130.1
11.-8 51,2 85.3
12-8 37,0 57 .9

 

 

 

The difference between the two gets of data 1s a measure of thc gteel
poisoning and its shielding of the fuel. Egoept near the reflector the im-
portance of the fuel ig of the form Y = Adg (BR) where Jo(BR) is the zero
order Bessel function corresponding to the radial asympotic solution of the

flux distribution in a bare cylindrical reactor. The constant B was evaluated
from the neutron flux measurements described in Section III-G below using
experimental points near the center of the reactor.

E. Reactivity Value of Reflector Coolant and Tubes

A measurement wag made evaluating the loss in reactivity due to the
reflector coolant and the Inconel tubes in which it was contalned. The
assembly contained sixty-six tubes, each 1/2" in diameter, filled with the

-23-
GAIN IN REACTIVITY

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

RADIAL POSITION (INCHES)

FIGURE 8

DWG. 21519
200‘-— | I
s - RADIAL IMPORTANCE OF
o, FUEL AND FUEL TUBE
" |
™~
@ 2 O STAINLESS STEEL FUEL
— N Jo (BR) TUBE Vs VOID
~ @ ALUMINUM FUEL TUBE
|00) Q ~ Vs VOID
~N
~ ©
- o ~
~
50 O— S Z
- ~0O
I | | ! | | I |
¢ 4 8 10 12 14 16

¥e
baslic mixture of ZrOp, C, and NaF but without UF,. The location of these
tubes is described in Section II-B. The removal of fifty-three of these
tubes resulted in an increase in reactivity of 66¢. Linear extrapolation
glves 82¢ as the total poisoning effect of all the tubes and their contents.

F. Attempt to Calibrate ARE Regulating Rod

 

One of the purposes of this preliminary assembly was to evaluate the
regulating rod designed for the ARE. The rod is a segmented array of small
cylindrical annuli, containing a poison, having their axes colinear, which
1s to be moved vertically along the reactor axis. Two such poison rods of
different values were fabricated for test. Each segment is 1.86" OD and
1.26" ID by about 2" long*. The "weak" rod is made up of fourteen segments
packed with AlnO; impregnated wilth B)C and containing 0.005 gm.BuC/cc. A
cylinder of staifless steel, 2-13/16" long, is located above and below these
segments., The "strong" rod was of identical construction with the poison
sections filled with a mixture containing 0.023 gm BjC/cc. Preparation
for evaluation of the rod required the removal of the central fuel tube, the
central Be0O column, and the insertion of Inconel guide tubes. The removal
of the central fuel.tube lowered the reactivity 107¢ the removal of the
central BeO colum, position 8-7, lowered it an additional 263¢. It then
became apparent that the loading contained insufficient excess reactivity
to override the further loss to be incurred by the introduction of the Inconel
guide. tubes and even the weaker regulating rod.

G. Measgurenment of Neutron Flux

 

The distribution of neutrons within the reactor core and reflector was
measured by the usual method using bare and cadmium covered indium foils.
The foils were 5/16" in dlameter cut from a 0.01" thick sheet of nine weight
per cent indium in aluminum mixture. The effective indium thickness was
5.4 mg/cm® or about 3 x 10-% inches. The cadmium covers were 0.02" thick.
After irradiation the folls were counted on four Amperex Geiger counters*¥*
with associated binary scalers. Each side of each foll was counted for two
minutes in all of the four counters, a total of eight sets of counts per
foil. The exposures wers such as to give the order of 5 x 10" observed counts
from each foil. The results were corrected empirically for counter dead-time
and background and normalized for the indium decay by methods described” pre-
viously.” A gold foll was irradlated at the same position 1n each run to
normalize the exposures. Some of the runs were further normalized by com-
paring the activities in indium foils located in the same position each time.
The precision of the neutron flux measurements due to errors in this normal-
1zation is estimated to be¥ 24. The location of the flux traverses made in
this egsembly is shown on the loading chart, Flg. 5. Table IV shows the

 

*¥ The rods are described by ORNL Drawings A-2-5-1, A-2.5-1-1, A-2-5-2, &nd
A-2.5-2-1,

#% Amperex Type 120C with 5.6 mg/cm® mica end window.

5 E. L. Zimmermanm "A Graphite Moderated Critical Assembly-CA-i4",
Y-881, December 7, 1952.

-25-
counting date and Fig. 9 shows the radial flux distribution at the midplane,

 

 

 

 

TABLE IV
Neutron Flux Traverse
Radial at Reactor Midplane
Indium Activity in Arbitrary Units

Cd Bare-Cd Cadmium

Radius Bare Covered Covered Fraction
0.875" 48,500 30,950 . 17,550 0.362
6.625 42,290 26,920 15,370 0.347
10.375 31,900 21,280 10,620 0.333
13.250 26,070 15,880 10,190 0.391
14,125 23,620 14,020 9,600 0.406
17.000 18,520 8,900 9,620 0.520
17.875 16,230 6,980 9,250 0.570
20.125 12,570 4,670 7,900 0.628
21.125 7,360 2,350 5,010 0.680

 

 

 

 

 

 

 

In addition to the bare and cadmium covered activations the thermal neutron
flux (Bare Indium Activation - Cd Covered Indium Activation) and the Cadmium
Fraction are plotted. The cadmium fraction is here defined as

Bare Indium Activation - Cd Covered Indium Activation
Bare Indium Activation

or the fraction of the absorptions in indium lying below the energy of the
cadmium cutoff. The dashed lines on the figure are Jyu(BR) curves plotted using
the two points furtherest from the reflector to determine the constant B.

The values of R for Jo(BR) = O for the three curves agree very well being 21.6"
for the total neutron distribution, 21.2" for the epi-cadmium, and 21.6"

for the thermal neutrons with an average of approximately 21.5". It was

this value of R which was used for the fuel importance curve Fig. 8.

An axial traverse was made in position 827 by placing the foils between

the BeO blocks. Table V shows the counting data and Fig. 10 is a plot of
the flux against the distance from the bottom of the assembly.

-26-
ACTIVITY (ARBITRARY UNITS)

DWG. 21520

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

INDIUM TRAVERSE RADIAL
MID-PLANE OF REACTOR
50
—\
\\\
\
X
N\ z
\\ & FRAGTION =
Cd COVERED INDIUM N / g
o
o
20 \C\ RE_INDIUM_ACTIVATION a
— > N\
——— «
BARE INDIUM (BRI
ACTIVATION MINUS >\\ Jo(BR U
Cd COVERED INDIUM BR
10|~ ACTIVATION N 2
ruet o ONOE
[ TUBESSf | ; | !
' ~
_ i | | ~R |
¢ 6 12 18 24
DISTANCE FROM AXIS OF REACTOR (INCHES) %

FIGURE

9
50

ACTIVITY (ARBITRARY UNITS)

-
DWG. 21521

 

INDIUM TRAVERSE, AXIAL

 

-

<« BARE INDIUM ACTIVATION

B

 

/

 

]

Cd COVERED INDIUM ACTIVATION \

\

 

 

 

INGONEL
SUPPORT
PLATE

 

/

 

N

 

Cd FRACTION

Cd FRACTION

 

 

 

 

/

 

 

 

 

ACTIVATION

 

BARE INDIUM ACTIVATION
MINUS Cd COVERED INDIUM”/,\\

 

 

/

 

 

 

 

 

 

 

 

 

 

6 12

18 24 30

DISTANCE FROM BOTTOM OF REACTOR (INCHES)

FIGURE

i0

36

8¢
TABILE V

 

Neutron Flux Traverse

Axial

 

T Distance from

Indium Activity in A:

rbitrary Units

 

 

 

 

 

‘Bottom of cd Bare-Cd Cadmium
Reactor Bare Covered Covered Fraction

o" 4,870 3,670 1,200 0.2L5

6 29,400 18,930 11,470 0.356

12 4l ,580 28,600 15,980 0358

18 48,500 30,950 17,550 0.362

2k 42,100 30,070 12,030 0.286

30 26,420 15,970 10,450 0.396

36 2,440 1,480 960 0.396

 

 

 

 

. The low cadmium fraction and the tendency for a non-symmetrical flux
at the bottom of the reactor are to be noted, a result probably due to the
fast neutron reflection by the Inconel support plate.

Similar longitudinal traverges were made at radial distances of 8,38"
and 15.88" (Cells 10-8 and 12-8) but in the upper half of the reactor only.
Table VI and Figs. 11 and 12 show these data.

 

 

 

 

 

 

 

 

 

 

TABLE VI
Neutron Flux Traverse
Longitudinal
Distance from Indium Activity In Arbitrary Units
Bottom of Cd Bare-Cd Cadmium
Reactor Bare Covered Covered Fraction
8.30" Radiuse
18" 39,200 24,800 14,400 0.367
2k 34,100 20,890 13,210 0.387
30 20,540 13,710 6,&30 0.332
36 1,920 1,280 640 0.330
15.88" Radius
18 20, 600 11,300 9,300 0.450
24 18,290 10,250 8,040 0.439
30 11,210 6,040 5,170 0.462
36 1,030 540 490 0.473

 

 

-29.
Ty
DWG., 21522

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

DISTANCE FROM BOTTOM OF REACTOR (INCHES)
FIGURE 11

50 i l
INDIUM TRAVERSE LONGITUDINAL AT
81.38 INCHES RADIUS
—40 BARE INDIUM ACTIVATION -8
=
>
>
x
__le
S0 | 2
o Cd COVERED INDIUM |ACTIVATION: =
x ' :
< <t
:20 - 4
£ — 1 . I\ ,~cd FRACTION 9
= . \(
10 BARE INDIUM ACTIVA-TION§\ 2
MINUS Cd COVERED INDIUM
ACTIVATION
| r
0 6 2 E 24 30 36

o€
.
DWG. 21523

 

 

 

 

 

 

50
INDIUM TRAVERSE
LONGITUDINAL AT 15.88
INCHES RADIUS
40
e
z
o
E 30
< | &
(0t ] —
= ,—Cd. FRACTION 3
< | -
- 20 — : ©
> O
: - l
E BARE INDIUM ACTIVATION
Q Cd. COVERED INDIUM AGTIVATION~|
N / ' :
1O )
'BARE INDIUM ACTIVATION \
MINUS Cd. COVERED INDIUM |
ACTIVATKTN l l %_\
0 6 12 18 24 30 36

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

DISTANCE FROM BOTTOM OF REACTOR (INCHES
12 -

FIGURE

T lE
IV, SECGNQ“LQADING
A. Critical Mass

Provision was made to increase the reactivity in the available tube
positions by increasing the uranium concentration in the powder mixture,
thereby allowing further investigation of the poison regulating and safety
rods of the ARE. The U-235 content of the fuel was increased from 0,1632
gm/cc to 0.2138 gm/cc by adding UF,, the relative proportions of the other
salts remaining the same. The details of the composition are given in
Section II-B. The "clean" assembly was again made critical requiring
5.19 kg U-235 in the 35.6" length of the core and contained in 42 fuel
tubes with Rod A removed an equivalent of 15¢.%¥ The values are to be
compared with 5.85 kg in 62 tubes which were initially critical with the
fuel of lower uranium density. The lower critical mass is a consequence
of geveral factors other than the change in uraniwi density. The stainless
steel loading was significantly less since the twenty positions formerly
filled with fuel tubes were left empty in thie experiment. The reflector
wag also effectively thicker since the overall dimensions of the BeO were
the same in both cases. The critical mass of th%s loading had not been pre-
dicted theoretically but subsequent calculations™ give a multiplication
constant of 1.009. No additional experiments were done with this clean
agsembly.

Immediately after determining the critical mass, the regulating rod
aggembly with its assoclated components were mounted at the center of the
reactor requiring the removal of the center fuel element, the center 36"
column of BeO, and the insertion of three Inconel tubes. The three tubes
extended completely through the reactor and were 3.,75" OD by 0.052" wall
thickness, 2-31/32" OD by 0.042" wall thickness, and 2-15/32" OD by 0,039"
wall thickness. The tubes were approximately concentric. This unit arrange-
ment was exactly the same as that in the ARE except the insulation in the
annulus between the inner two tubes was omitted*¥,

A step-wise evaluation of this change was not done but the mass required
to keep the system critical was determined. The required loading was 55 fuel
tubes or 6.80 kg U-235 with the regulating rod inserted a distance corres-
ponding to a reactivity value of 54.5¢ which was approximately equal to
that of one fuel tube. The critical mass therefore, with the regulating
rod assembly mounted but with the rod removed was 6.68 kg U-235 in 5k fuel
tubes. Figure 13 glves the arrangement of the 55 fuel tubes. This repre-
sents an increase of 1.5 kg U-235, or 12 fuel tubes over the unpoisoned
gystem,

The large change in critical mass shows the Inconel at the center of
the core is a strong poison which greatly reduced the total rod effectiveness.

B. Regulating Rod Calibration

The weak ARE regulating rod was evaluated in the central position, 8.7,
¥ The loading pattern consisted of a hexagon of 4 columns on the slde, plus the
following tubes: 10-10, 10-L, 6-k, 6-10, 8-3.

6 Mills, C. B. ANP Quarterly Progress Report June 10, 1953, ORNL 1556, p.28

*% A subsequent evaluation of 5505, a rough equivalent to thisg insulation, mede
independently of the Inconel tubes, showed it to effect only a small change
in reactivity. The result is given in Table X, and an analysis in the

Appendix, semple 8.
-32.
 

 
and subsequently used for operational control. The method of calibration

was the same as that described in the Section III-B for the fuel tube type

rod and ylelded a total reactivity value of about 125¢. At the "zero" rod
position all of the B)C of the rod was above the reactor and the bottom of

the stainless steel section of the rod was flush with the top of the BeO.
Figure 1bh gives the calibration curve of the regulating rod. It is of partic-
ular interest that the minimum and maximum values of the rod as a reactor poison
occurred when it was somewhat displaced from the limits of its travel, being a
minimum when inserted approximately 3" and a maximum when inserted 32", The
reactivity Increase in the interval from zero to 3" can probably be attributed
to a contribution of the rod to the neutron reflection or to the reduction in
neutron channeling in this bare ended assembly. When the rod was at the 32"
position, the top of the 30" section containing boron was at the level of the
top of the BeO blocks. Further downward displacement, therefore, did not
introduce addltional poison and redistributed that already present only
slightly. It d4id, however, bring the top 2-3/16" long steel section into

the core, providing some neutron reflection and reducing the channeling. The
reactivity would have increased upon further lowering of the rod as boron was
removed at the bottom of the core. In a second installation of the rod, to be
described later, at a radial distance of T.5" from the center, the above be-
havoir at the 3" displacement was verified. It is to be noted in Fig. 15 that
the net result of these end effects is to give the rod a negative sensitivity
in these ranges. The irregularities in the data near the peak of the sensitivity
curve has been attributed to the interaction with an adjacent fuel type control
rod which was being Ingerted with the poison rod. At this position their ends
were at the same level. These effects were not investigated further.

Since the maximum reactivity of this asgembly did not occur when the rod
wag at the upper limit of its travel it is important that some consideration
be given to the operating position of the ARE gafety rods. The data of Figs. 1lb
and 15, or similar results from the ARE, should be used to ascertain the appro-
priate ralsed position of the ARE rods to insure that the reactivity is always
decreased as they are insgerted, without significant loss in total rod value. A
pogitive mechanical stop should be provided to limit the downward travel of the
rods at the points where the reactivity is a minimum. It is believed, however,
that the neutron reflection provided is the ARE by the pressure shell and the
moderator coolant will reduce these effects.

C. Fuel Tube Reactivity Coefficients

In an effort to provide additional data for comparison with the clean
reactor of the first loading, the experiment in which the importance of the
stainless steel fuel tube was measured as a function of radius was repeated
and the data are shown in Table VII and Fig. 16. Since the holes in the BeO
were 3-3/4" apart, no intermediate points have & real meaning and therefore
no curve has been drawn through the measured values.

3k
ROD VALUE ¢

DWG. !l!!

 

 

120 v

 

 

 

 

 

 

 

- / CONTROL ROD B CALIBRATION
100 / ARE REGULATING ROD (WEAK)

 

90

 

80_ /

 

 

70 /
60

 

 

40 /

 

 

/

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

—t | A ] 1 ] l ] i i 1 ]

 

 

 

 

 

 

| | 1
0 2 4 6 € o 2 4 16 8 20 22 24 2 28 30 32 4
ROD POSITION (INCHES)

FIGURE 14
2/INCH

ROD SENSITIVITY

DWG., 21526

 

I ! !

 

0

 

 

ARE REGULATING ROD
SENSITIVITY

 

 

 

 

 

 

 

 

| A

 

 

 

 

 

 

 

 

N\

 

N

 

 

 

\l

 

 

 

 

 

 

 

 

 

 

 

|

 

 

]

 

 

 

 

 

18 20

22

24

ROD POSITION ( INCHES)

FIGURE

15

26

28

 

 

1~

30

 
100

GAIN IN REACTIVITY ¢

20

@®
O

N
O

H
O

20

e
DWG. 21527

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

RADIAL POSITION (INCHES)

FIGURE

16

T T [ ] | ! |
RADIAL |IMPORTANCE
o STAINLESS STEEL FUEL TUBE
VS. VOID
0
o
o
REGULATING ROD GUIDE
‘/— TUBES (INCONEL)
| 1 : | : 1 [
2 4 6 8 0 12 4 6

L€
TABLE VII

 

Radial Importance of Stainless Steel Fuel Tube

 

 

, Reactivity
Posltion Value in ¢
9-8 131.0
10-8 9k, 1
11-8 69.9
12-8 5743

 

 

 

 

A comparison of the reactivity coefficients of an Inconel fuel tube with the
stalnless steel fuel tube was made. An Inconel tube of ARE dimensiong and
specifications was filled with powdered fuel having the same density and U-235
concentration as that contained in the stainless steel tube. This tube then
replaced the stainless steel one in successive positions along a radius. In
another experiment some evaluation of the radial effect of a variation in
U-235 concentration was indicated by substituting a stainless steel fuel

tube with the first loading (0.163 gm of U-235/cc) for one with the higher
concentration (0.21% gm of U-235/cc). A summary of these data is given in
Table VIII and Fig. 17.

TABLE VIII

 

Radial Importance of Fuel and Container Material
Compared to Normal Fuel Tube

 

 

 

Reactivity Value in ¢
Low U-235 Density! Inconel
Pogition Fuel Tube Fuel Tube
5.8 -29.0 -27.2
10-8 -21.1 -24.9
11-8 -1k ,5 -17.2
12.8 - 9.0 ~-10.1

 

 

 

 

 

The reduction of the relative value of the Inconel fuel tube at the 3.75"
radius (position on 9-8) might be explained by its proximity to the large
quantity of Inconel in the regulating rod guide tubes. In general, the
stainless steel fuel tube values and the low density fuel tube values are
consistent but do not compare very well with the flux and power distri-
butions described later in this report.

D. Reactivity Value of Reflector Coolant and Tubes

Since the center of the reactor was strongly poisoned by the Inconel
tubes of the regulating rod assembly which might cause a change in the
total reactivity value of the filled reflector coolant tubes, it was re-
measured for this asgembly. The removal of 56 of 66 tubes gave an increase

-38-
LOSS IN REACTIVITY ¢

30

N
o

o

DWG, 2]525

 

 

 

T ! i ! 1 l !

RADIAL IMPORTANCE
| l

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

FIGURE

17

e “OW DENSITY FUEL TUBE —
® VS. NORMAL FUEL TUBE
o
o NCONEL FUEL TUBE vs.
O STAINLESS STEEL FUEL TUBE
®
o)
o
0O
o
REGULATING ROD GUIDE
/TUBES (INCONEL) |
1 | | [ [ 1 L 1
2 4 6 8 10 12 14 16
RADIAL POSITION (INCHES)

 
in reactivity of 78.2¢. When this is extrapolated linearly, the value of

66 tubes becomes 92¢, an increase from the 82¢ extrapolated in the first
loading. There was not enough additional Inconel tubing available to
evaluate the containers alone but, considering the small value of the coolant
and contalners combined, this did not appear necessary. It should be noted
that the ARE has a similar coolant in the interstices of the BeQO blocks.
Because of the difficulties 1n handling the hygroscopic material, a total
evaluation of this poison was impractical. As reported below, however,
constituents of this coolant were evaluated at one point in the reactor.

 

E. Reactivity Coefficlents

The effects on the reactor of a number of different materials were
evaluated at a single point as follows: The fuel tube of position 10-8
(7%" from the center) was removed and, with the reactor critical, the
control rod poesition was noted as the base point. Thls base point was
used ag reference for the subsequent runs in which the material to be
evaluated was placed in this same position. In the case of a large change
in reactivity, additional fuel
tubes were added at selected
points on the core periphery.
The tubes were evaluated as

TABIE TX

 

 

 

added and the data are given Reactivity Value of Fuel Tubes
in Table IX. Posltion Valvue in ¢
~ The sample materials were 11-3 46,4
used in the form of rods, tubes, L5 60.5
chips, or powder the latter two 5-10 39.3
being packed into aluminum tubes 5-G 51.8
and sealed to prevent leakage and 5-h.} 99
water absorption. The reactivity .6 773
coefficlients of the aluminum tubes I k8.7
(semples 1 and 2 in Table X) were 7-4 56.5
measured and used to obtain the net 5-10} ’
result for the samples contained. 6-11 83.5
Two pamples (chromium and manganese) 5-3 '
were packed into both an aluminum h-B.} 149,7
tube and into an annulus formed by haob
two concentric aluminum tubes while

others were run twice in a single
tube using different quantities of

materials, Comparisons gave some measures

 

 

 

of the self-.shielding.

 

For cobalt

in particalar there is & marked change in the loss in reactivity per mole be-
tween the two runs. Table X glves a complete summary of these data.

There were several reactivity coefficients measured in the same position
(10-8) which pertained more directly to the fuel and contalners. Some of
these values have been given previously but for each comparison a summary
is given in Table XI. The aluminum fuel tube was packed with fuel to the
game density (0.2138 gm U-235/cc) as the stainless steel fuel tube. All
reactivity values are referred to a vold in the test cell.

=40~
 

 

L s ———— . — —— W'

 

TABLE X

 

 

 

 

 

 

Metl & Dimensions (Inches) Volume
Sample Form U. D. “Wall Thick.  Length .3 oo
1 Al Tube 1,001 0.034 40.0 4,17 68 .33
2 Al Tubes 0.545 0.092 39,05 152.1
(Annulus )*

3 Iron Rod 1.248 Solid 44.5 54443 891,95
4 Iron Rod 0.50 Solid 44.95 8.83 144.70
5 Iron Tube 0¢543 0.091 40.12 5.18 84.88
6 Nickel Rod 0.999 Solid 42-1/16 32,97 540.3
7 Nickel Tube 1,004 0.036 48.0 5.28 786f52
8 8i02(Sand) - 0.930 Solid Pack 39.0 26,49 434.1
9  KF Powder 0.930  Solid Pack  39.0 26449 434,41
10 NaF Powder 0,930 Solid Pagk 39,0 26?49 434,1
11 Cr Powder 0,930 Solid Pack 3940 26.49 434.1
12 Cr Powder 0.930 0.1925 39,0 17.39 285.0
13 Mn (Chips) 0.930 Solid Pack 39.0 26.49 434.1
14 Mn (Chins) 0,930 0.1925 39,0 17.39 285.0
15 Cobalt (Slugs) 0.930 Random Filled 39-1/8  26.58 435.6
16 Cobalt (Slugs)  0.408 Solid 39.0 5.1 83,57
17 S. Steel Tube  1.240  0.06 40.0 8,92 146.2
1e Inconel Tube 1.235 0,06 | 40,0 3487 145.3

 

Total
Weight  Density
gms gm/ce
184.5  2.70
411 2.70
6995 7.84
1129 7.80
652 7.68
478932 8.86
7705 8490
1.76
1,43
1.08
1801.6 4,15
1193.1 4,19
1273.0 2,93
769 2470
1966.5  4.51
717.3  8.58
1119.6 7,66
1200.2

B.26

 

Weight in
Core Moles in
gm Core
164.3 6,09
370 13,73
5599 100.25
894 16.2
578 10.35
4056 63,1
571 9.73
697 11.6
566 3 975
427. 7 10.18
1645 31,63
1089.8 20095
1162 21.15
702 12.78
1790 30.37
655" 11,11
997.1 18.01
106849 18,72

 

flyhole
g - in Core
~ 2.0 = 0.33
- 4,2 = 0.31
-196.7 = 1.96
- 0.8 = 3.14
- 33.8 = 3.26
~20643 =~ 2.98
- 83.9 - b5.54
+ 0.9 $0.08
= 27.6 = 2.83
- 4.5 - 0c44
=106.6 =~ 3.37
= 76.3 =~ 3.64
-308,8 - 9,87
-153.8 -12.03
=325.3 =10,T71
-19034 ~17.14
- 70,0 - 3.89
= 93,5 = 4.99

# Sample #2 consisted of two tubes assembled to form the annulus used for samples 12 & 13. The outer tube was sample 1 and the
inmner was as described under dimensions of sample 2 above, ‘

—l]w

 
TABLE XI

 

Sumary of Fuel and Tube Reactivity Coefficlents

 

 

Material Reactivity Change
Aluminum Fuel Tube + 155.5¢
Low U-235 Density Stain-
less Steel Fuel Tube + 73.0

Stainless Steel Fuel Tube + 94.1
Stainless Steel Tube

(Empty) - 70.0
Inconel Fuel Tube 4+ 69.2
Inconel Tube (empty) - 93.5
Stainless Steel Tube |
~and Cast Fuel ‘ + 87.b
Inconel Tube and Cast .
Fuel + 62,2

 

 

 

 

F. Evaluatlon of Fuel Tube Type Safety Rod

 

A reference check of the value of the fuel tube type safety rod was
made by the rod drop method”?., This method, briefly, involves making the
reactor critical with the safety rod in the most reactive position and then,
as the rod is actuated, using a high speed recorder to observe the fast tran-
sient of neutron level decay. The relation which applies in this case i=s

No - Nl

Ny

x 100¢ = the reactivity change in cents

where:

NO ig the neutron level at critical
Ny 1s the extrapolated level at the time the rod 1s dropped.

The value of one of the rods, #2, was 480¢ and that of both when removed
gimultaneously was 900¢.

G. Reactivity Value of End Reflector

To obtain some measure of the contribution to the reactivity of the
reflector coolant and pressure shell at the ends of the ARE, the following
measurements were made. A one-inch thick 6" x 9" layer of stainless steel
was placed on the top of the fuel tubes in a locatlon shown by the dotted
lines in the vicinity of position 10-8 of Fig. 13. This gave an increase in
reactivity of approximately 6¢. A like area and thickness of sodium gave a
2.9¢ increase in reactivity. This sodium was contained in 8 mil wall stainless

gteel cans.
H. Neutron Flux Distribution

The neutron flux distribution was measured with indium foils by the method
. previously described except that a remotely placed uranium metal disk and alum.
inum catcher foil were used to normalize the power level.from run to run. The

 

5 E. L. Zimmerman, "A Graphite Moderated Critical Assembly-CA-4", Y-881,
December T, 1952,
~hoo
urenium disk was approximately 1- 7/16" in dlameter and 0.0l" thick, weighed
4.5 gm and was enriched to 93% in U-235., The Al foil was against the uranium
dise during the foll exposure and was counted in an in-chamber type pro-
porticnal counter. The activities were empirically corrected for counter
dead time, background, and fissgion fragment decay and then used for power
normalization Table XII and Fig. 18 give the result of the bare indium

and cadmium covered indium longitudinal traverse at a point 12.06" from the
center (position 11-8). The zero of the abscissa is the bottom of the BeO
column where it rests on the 1" thick Inconel support plate. The scattering
by the Inconel possibly accounts for the reduction of the Cd fraction at this
point.

 

 

 

 

TABLE XII
Neutron Flux Traverse
AdJjacent to Tnconel Cast Fuel Tube
Longitudinal at 12.06" Radius
Distance
from Indium Activity in Arbitrary Units
Bottom of Cd Bare-Cd Cadmium
Reactor Bare Covered Covered Fraction
o 8,920 7,025 1,895 0.212
6 37,280 26,450 10,830 0.299
12 55,396 38,630 16,766 0,303
18 59,941 41,019 18,922 0.316
24 51,612 33,65k 17,958 0.348
30 32,202 21,062 11,140 0.346
36 2,858 1,845 1,013 0,35k

 

 

 

 

 

 

 

The radial flux traverse at the midplane of the reactor is given in Table
XIII and Fig. 19. The dashed lines on the figure are activities extrapolated
from the data obtained in the fine structure measurements taken near the
eleven inch position. There were no similar measurements made in the first
loading. The wide gap in this traverse was unexplored because of the Iim-
portance which was attached to the study of a unit core cell In the time
available. This emphasis has been at least partially contradicted by the
fission flux traverse described in the following sectlon. The strong effect
of the center regulating rod assembly is indicated in these curves. The
comparatively small flux and cadmium fractlon depressions at the center of
the regulating rod accentuates the relative inefficiency of this type of rod
and guide tube arrangament for reactor control

-43-
(ARBITRARY UNITS)

ACTIVITY

4
DWG. 21529

 

- | | | : l

INDIUM TRAVERSE LONGITUDINAL

AT [2.06

INCHES RADIUS
| I

 

 

 

50

40

30

20

10

A

|
| | |

/BARE INDIUM ACTIVATION

 

 

 

/—Q\

)\

 

Cd COVERED
INDIUM ACTIVATION

 

{

 

\
N

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

N\
Cd FRACTION =
TOP OF BeO
BARE INDIUM ACTIVATION N
MINUS Cd COVERED INDIUM
ACTIVATION
INCONEL SUPPORT PLATE |
] ] i ] ] i i | i ] |
0 6 12 18 24 30 36

DISTANCE FROM BOTTOM OF REACTOR (INCHES)

FIGURE

I8

Cd FRACTION
(ARBITRARY UNITS)

ACTIVITY

S Dwg. 21530

 

T T I | | T T T T
INDIUM TRAVERSE
RADIAL AT MIDPLANE
REGULATING ROD IN

]

!

, &S

 

 

 

 

 

 

 

BARE INDIUM ACTIVATION

   

80— INCONEL [
|/ GUIDE Tu7
I
I
.'
/
/

ol ]

 

  

 

 

 

 

 

 

 

 

Cd FRACTION

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

% Cd FRACTION
-~
SR L
% I
W T R
= I \
7 | l T/
20 :. "'\\' /M \\ \\ + L
M| /] /‘(\ 7! \ 7/ \ 4\
i~ L1 1 I s /
/| “ BARE INDIUM ACTIVATION
10 2. c MINUS Cd COVERED y—FUEL—|
s 4 ' INDIUM ACTIVATION TUBES
- ROD |
— LTy ]
0 | | ! I f | l | l [
2 4 6 8 10 12 14 16 18 20

DISTANCE FROM AXIS OF REACTOR (INCHES)
FIGURE 19

Sy
e - '

TABIE XIII

 

 

 

 

 

 

 

 

Neutron Flux Traverse
Radial at Midplane
(Regulating Rod In)
Indium Activity in Arbitrary Units
Cd Bare-Cd Cadmium
Radius Bare Covered Covered Fraction
0.00" 61,492 46,396 15,096 0.245
1.03 - 61,265 46,508 14,760 0.241
1.28 61,579 46,185 15,329 0.249
1.69 63,181 49,418 13,763 0.218
2.06 68,352 W7,24h 21,108 0.309
2.06 69,858 50,976 18,882 0.270
2.94 68,414 54,196 14,220 0.208
4,56 79,06k 54,666 24,400 0.309
5.4k 81,787 57,679 24,108 0.295
6.69 73,689 54,792 18,898 0.256
9.56 77,086 47,463 29,623 0.384
10.4k4 66,231 43,445 22,786 0.344
10.88 53,510 38,794 14,716 0.275
©11.25 52,798 39,888 12,910 0.245
11.62 53,367 38,197 15,170 0.284
12.06 57,380 38,441 18,939 0.331
12.94 61,021 37,054 23,967 0.392
22,06 12,818 5,122 7,696 0.600
| 23.56 4,716 -
24 .00 1,998 633 1,335 0.668

 

 

 

Table XIV and Fig 20 give a similar traverse at the same location but with
the regulating rod removed. The neutron flux is slightly greater than that
observed in the preceding experiment.

TABLE XTIV

 

KNeutron Flux Traverse

Radial at Midplane
(Regulating Rod OQut)

 

 

 

 

 

 

 

 

Indium Activity in Arbltrary Units
Cd Bare-Cd Cadmium
Radius Bare Covered Covered Fraction
0.00" 66,596 50,628 15,968 0.240
1.28 68,250 50,233 18,017 0.26k4
1.69 67,604 52,705 14, 899 0.220
2,06 72,671 52,689 19,982 0.275
5 oIl 85,886 59,994 25,892 0.301
6.69 76,695 55,810 23,885 0,300
9,19 81,760 55,818 25,942 0.318

 

“L6-

 
(ARBITRARY UNITS)

ACTIVITY

-.Dwo 21531

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

1 | l n v T
A INDIUM TRAVERSE ]
80 | ’/'\ A RADIAL AT MIDPLANE
'gch:IODNEEL A REGULATING ROD oOUuT
TUBES \—BARE INDIUM ACTIVATION N
70—-—-~"'_‘7H‘,P
60 — \

- o \.(‘(/—Cd COVERED INDIUM ACTIVATION —
50 A
40

— Cd FRACTION —
30 —

4]

] —fl"‘“‘—- —
2 ///gle INDIUM ACTIVATION

> MINUS Cd COVERED

— INDIUM ACTIVATION B
10

TUBES
| 'j | ’ _
l ’ l [ | | ! | | I
0 2 4 6 8 12 14 16 E 20 22 24

DISTANCE FROM AX1S OF REACTOR (INCHES)
FIGURE 20

Cd FRACTION

44
An Inconel tube, loaded with the cast fuel salt described in Section II-B
was ingerted in position 11-8 (11.25" from the center) replacing the normal
powder packed stainless steel fuel tube., A longitudinal neutron flux traverse
wvag made axlally in this tube with the folls placed between the cast slugs.
Table XV and Fig. 21 are the result of this traverse. The three points of
Fig. 19 at positions 10.88". 11.25", and 11.62" were made in the same Inconel
tube -cast fuel arrangement.

 

 

 

 

 

TABLE XV
Reutron Flux Traverse
Cast Fuel in Inconel Tube
Longitudinal at 11.25 inches Radius -
Distance
from Indium Activity in Arbitrary Units
Bottom of Cd Bare-Cd Cadmium
Reactor Bare Covered Covered Fraction
0.18" 9,487 7,317 2,170 0.229
3.7Th . 25,062 19,622 5,440 0.217
6.68 36,026 26,126 9,900 0.275
10.68 17,233 33,638 13,595 0.288
13.7h 50,90k 38,140 12,764 0.251
20.30 52,798 39,868 12,610 0.245
25.80 42,726 30,759 11,967 0,280
29.74 29,709 20,748 8,961 0.302
34,80 7,467 4,978 2,489 0.333

 

 

 

 

 

 

J. Fission Neutron Flux Distribution

A measure of the distribution of neutrons causing fisslions wasg obtained
by use of the catcher foil method. A 4-1/2 gm enriched U disk 0,01" in thickness
end approximately 1-7/16" in diameter, was_exposed, together with en Al catcher
foil, at each of the points in the reactor’ indicated on Fig. 13. The exposures
were repeated with the disks and catcher foils covered with 0.02" Cd. Table XVI
and Fig. 22 give a radlal traverse at the midplane of the reactor.

 

* This experiment is to be distinguished from the usual power digtribution
measurements in which the catcher foils are placed adjacent to the tranlum
bearing fuel in a critical assembly. This latter type of experiment’is
described In the following section.

-h8-
e
DWG. 21532

 

t T l | | T l T T ! '
INDIUM TRAVERSE LONGITUDINAL
AT {1.25 INCHES RADIUS IN INCONEL
TUBE WITH CAST FUEL

 

PN ‘

 

)/ \ BARE INDIUM ACTIVATION

 

/ /<Cd COVERED INDIUM

 

 

 

 

ACTIVITY (ARBITRARY UNITS)

 

 

/ ACTIVATION

 

 

    

Cd FRACTION

 

 

 

   

BARE INDIUM ACTIVATION

MINUS Cd COVERED INDIUM
| ACTIVATION l
|
] |

INCONEL SUPPORT ‘PLATE l |
l

L e !

—TOP O F BeO

 

 

 

 

6 12 18 24 30 36
DISTANCE FROM BOTTOM OF REACTOR (INCHES)

FIGURE 2I

Cd FRACTION

 

6y
(ARBITRARY UNITS)

ACTIVITY

S,
DWG. 21533

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

| 1 1 i | |
FISSION FLUX DISTRIBUTION
CATCHER FOIL-FUEL DISK
RADIAL AT MID-PLANE
- BARE -
ACTIVATION
®
" //_\
7 cd FRACTION {\
ABARE ACTIVATION MINUS
cd COVERED ACTIVATION
INCONEL GUIDE
TUBES
Ccd COVERED ACTIVATION
. ] |
| i , TUBE?m \'
[ 1 [ ‘ | |
0 6 12 8 24

DISTANCE FROM AXIS OF REACTOR (INCHES)

FIGURE 22

@

o

0s

Cd FRACTION
TABLE XVI

 

Figgion Neutron Flux Distribution

Catcher Foil-Uranium Disk

Radial at Midplane

 

 

 

Relative Activity
Cd Bare-Cd Cadmium
Radius Bare Covered Covered Fraction
2.59" 8,805 2,104 6,701 0.761
4.61 11,4k 1,949 9,495 0.830
10.09 11,929 1,932 9,997 0.838
13.8k4 9,450 1,L64 7,986 0.8L5
17.59 10,099 809 9,290 0.920
21,34 5,496 268 5,228 0.952

 

 

 

 

 

 

 

The flux of low energy neutrons producing fissions peaks near the reflector,
and 18 depressed by the Inconel tube at the center. ' The cadmium fraction®
follows roughly the same pattern. Similar half length longitudinal traverses
were made 10,09" from the center and are shown in Table XVII and Fig. 23.

TABLE XVII

 

Fission Keutron Flux Distribution
Catcher Foil-Uranium Disks
Longitudinal at 10.09" Radius

 

Relative Activity

 

 

 

 

 

 

, Cd Bare-Cd Cadmium
Reactor Bare Covered Covered Fraction
18" 11,929 1,932 9,997 0.838
2L 9,934 1,778 8,156 0.822
36 b1l 93 319 | 0.775

 

 

 

 

* Ag before the cadmium fraction 1s defined as the bare activatlion minus the
cadmivm covered activaetion divided by the bare activation. It 1s a measure
of the fraction of fissions produced by neutrons having energies below the
cadmium cut-off.

-51-
(ARBITRARY UNITS)

ACTIVITY

T DG 21534

 

FISSION FLUX DISTRIBUTION
CATCHER FOIL — FUEL DISK
LONGITUDINAL AT 10.09 INCHES RADIUS

 

BARE ACTIVATION

 

 

 

T
\ e FRACTION

‘\ -

 

 

 

\ \

 

 

 

 

 

BARE ACTIVATION

 

 

MINUS Cd COVERED ACTIVATION

| | |

 

1 I

1

 

 

 

CCd COVERED ACTIVATION

 

 

 

 

»

 

 

8 12 16 20 249 28 32 36

DISTANCE FROM BOTTOM OF REACTOR (INCHES)
FIGURE 23

©

N
Cd FRACTION

s
J. Power Distribution

 

Three longitudinal power distributions in this reactor aggembly were
measured by using aluminum catcher foils, 0.5" in diameter and 0.005" thick,
placed against the end surfaces of the short cast fuel slugs contained in en In-
conel tube. The size of these foils, initilally selected arbitrarily, was
such. that the counting rate from sach was higher than necessary for the
desired statistics. In the arbitrary units reported, an activity of 50
represents approximately 10- counts. Had time permitted it would have been
desirable to repeat the experiment using smaller foils, thereby improving
the resolution, particularly since the results were very sensitive to foil
location because of fuel self-shielding. Xach catcher foll was nominally
located centrally on the axis of a fuel slug. One traverse extended from
below the Inconel support plate to above the top of the BeQ; the other two
covered the upper half of the core. Table XVIII and Fig. 2L give the
data obtained in the fuel tubes at the indicated radii.

TABLE XVIII

 

Power Traverse
Cast Fuel in Inconel Tube
Longitudinal at three Radil

 

 

 

Distance
from
Bottom of Catcher FoOll Activity Arbitrary it
Reactor Core 11.,25" radius 3.75" radius 15™ radius
-2.26" 3.52
0.18 10.35
3.7k 33.39
6.68 47.38
12.43 67 .41
17.36 T0.32
18.24 T72.71 T7.07 66.98
23.17 61.26 66.01 50,03
28.92 41.85 44,03 26.43
35.42 6.36 T7.09 3.25
37.87 2,53 3.02 2.09

 

 

 

 

 

 

Also plotted on the figure for compsrison ig the normalized longitudinal indium
thermal neutron curve of Fig. 21 measured at a distance of 11.25" from the center.
The data of Fig. 2L are replotted in Fig. 25 showing the radlal power distri-
bution at several elevations in the reactor all normalized at the center., It

is to be noted that the 37.86" elevation traverse is 2" above the top of the

BeO column. -

Two addlitional experiments were performed in an attempt to msasure the
self-shislding in this fuel. In the first, a cast fuel slug was split longi-
tudinaly along a diameter, three aluminum catcher folls were placed hetween the
eplit surfaces as shown in Fig. 26a, and the slug was loaded in an Inconel fuel
tube at the midplane of the reactor about 11" from the center. The tube was
otherwise filled with cast slugs. Since the dlameters of the split eslug and
the aluminum foils were 1-1/16" and 1/2", respectively, the two "side" folls
were separated only 1/16; so the resolution of the measurements ls not high.

=53~
ACTIVITY (ARBITRARY UNITS)

“S DWW G 21535
i

 

 

 

 

 

]

POWER TRAVERSE
CAST FUEL IN INCONEL
TUBE LONGITUDINAL AT

!

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

80 INDICATED RADIUS
\/—3.75 INCHES RADIUS l
70 | | |
BARE INDIUM ACTIVATION
\\ /MINUS Cd COVERED INDIUM
60 Loy D ACTIVATION (NORMALIZED)  _|
1.25 INCHES—(], N\
RADIUS \ |
/ \
50 \ X
15 INCHES—>\ %\ \
40— |INCONEL RADIUS \
SUPPORT
PLATE \
| b
/ TOP OF BeO
20 /4 \\ \ e0 |
’ & s\\‘l
=
-4 0 4 8 12 16 20 24 28 32 36

DISTANCE FROM BOTTOM OF REACTOR (INCHES)

FIGURE 24

¥S
RELATIVE ACTIVITY

- 4
DWG. 21536

 

POWER TRAVERSE

 

 

CAST FUEL IN INCONEL TUBE
RADIAL AT 5 ELEVATIONS

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

l.o et
. \ T—18.24 INCHES FROM BOTTOM
. N OF REACTOR —
. 23.17
7 \?3286
6 \\ 28.92
.5 ‘ \35 42
.4
3
.2
R
0 8 10 12 14 16 i8 20

DISTANCE FROM AXIS OF REACTOR (INCHES)
FIGURE 25

gs
The relative fission fragment activities collected on the three folls are

noted on the flgure. Since this fission fragment activity ls strongly de-
Pendent on the uranium density iIn the slug surface, a measure was made of

the homogeniety of the uranium in the second experiment. Three aluminum

folls were again located on the split surface of the slug as shown in Fig.

26b. This arrangement, without the second half of the slug, was exposed %o

the leakage neutron flux of another critical assembly and the resultant fission
fragment activitles measured. Ignoring the differential back scattering across
the face of the slug and asguming the flux to which it was exposed to be uniform,
these activities, included in Fig. 26b, give a relative distribution of uranium
across the slug face used in the self-shielding measurements. Applying this
indicated variation as a linear correction to the values in Fig. 26a and normal-
l1zing, the values for the relative fission rates across the fuel slug becoms
72.6 and 72.4 at the sides of 65.6 at the center. These measuremsnts and those
of the neutron flux shown on Flg. 19 within the fuel tube at an abscissa of

11.5" were made at the same points.

— Cast Fuel
76 °7l+
N Aluminum _/
Catcher Folls

Fig. 26a Fig. 26b

 

 

 

 

 

 

Bd

 

 

Fuel Self=Shielding

K. Evaluation of ARE Type Safety Rod

 

The safety rod of the ARE was mounted in position 10-8 (7%" from the
center). The components of this rod assembly are identical with those of the
regulating rod except the construction of the poison rod itself. The total
length of this rod is 40-3/8" which includes a 2-3/16" stainless steel guide
on each end. The polson section consisted of elghteen annular stalnless steel
cang, each 2" long and 2" OD, containing a sintered B),C-iron mixture having
a density of 2.45 gm/cc and composed of 80% B,C by volume. The dimensions
of the sintered slugs were 1.86" 0D, 1.26" ID and 1.88" long.

In this case, the step-wise evaluatlon of the reactivity change during
the mounting of the equipment for this safety rod was made. The removal of
the fuel tube gave a loss of 9h.1l¢. The removal of the 35.6" column of BeO
gave a loss of 224¢, the insertion of the two larger Inconel thimble tubes
gave a loss of 172¢, the third or central Inconel gulde tube gave a loss of

_56--
31¢ for a total of 521¢. Since there was not sufficient reactivity avail-
able to maintain the assembly critical with the safety rod inserted, it
was necessary to agaln use the rod drop method of evaluation. It was
obgerved that the maximum reactivity of the system with the safety rod
agsembly mounted occurred when the bottom of the rod was 3.2" below the
top of the BeO, a condition comparable to that found with the regulating
rod mounted on the axis. The rod was dropped a distance of 32-1/4" from
this position and caused a decrease in reactivity of 550¢.

For comparison the two ARE regulating rods were evaluated by the same
method in this position, 10-8. Although of somewhat different construction,
thelr position of maximum reactivity was also with the lower.end of the
rod 3.2" below the top of the BeO., The rods were also "dropped" 32-1/4"
from this position. The reactivity value of the "strong" rod wag 160¢ and
that of the "weak" rod was 80¢. The "weak" rod referred to here is the one
calibrated In the center of the reactor by the period method described in
Section IV-B. The value of the rod there was 125¢.,

V. SUMMARY

The preliminary assembly for the ARE contributed information for
checking calculational methods and applicable nuclear constants for a BeQ
moderated and reflected reactor. The rather close agreement between the pre-
dicted and experimentally determined critical mass of the first loading In-
creased the confidence in the multigroup calculations. Although the first
calculated value of the critical mass was 5% lower than the measured mass,
subsequent refinements in the calculations reduced this to about 1%.

The step by step transition from the clean, unpoisoned reactor to a
low temperature mockup of the ARE was intended to aid in the understanding
of the mechanisms Involved for. calculational purposes and did so to the extent
that the transition was carried out.

The calibration of the "weaker" of the two regulating rods indicated that
1ts value was, for safety reasons, too high for use with an automatic control
device of the type designed for the circulating fuel reactor. The total
value of the ARE safety rod was found to be sufficient to accomplish 1ts
intended purpose.

The flux measurements were not as complste as could be desired, although
the radial iImportance functions and reactivity coefficients which were
measured should aid in the understanding of the flux and spectrum of this
reactor. There was essentially no work dcne in determining the neutron leak-
age spectrum from the reflector surface. The power distribution measurements
acrogs the fuel within a tube suffered from poor resolution but give some
indication of the value of the fuel self-shielding.

..5"[ -
VI. ACENOWLEDGEMENTS

 

The work of Dr. L. G. Overholser, Messrs. D. R. Cuneo and
A, B, Townsend in the preparation of the fuel is gratefully ack-
nowledged. Appreciation is expressed to Drs. R. C. Keen, and E, L.
Zimmerman for their help in preparation and review of this report.

_58-
ANALYSIS OF MATERIALS

APPENDIX

 

 

 

 

 

SAMPLE| MATERIAL FORM Ag Al B Ba Be Ca Cd Co Cr Cu Fe Gd K Li Mg Mn Mo Na Ni Pb Si Sn [ Sr| Ta Ti v W | Zn | Zy
1 Al 1in. OD Tube | % wt|< .04 <.004 | <,02|<.001 <.08{<.04 | <.04[<.04]<.08 |[<1 <.04 | <.04 < .04 < .08 <04 <.15]<.04(<.2|<1 <.04 | <.04 |1 <.3<.08
2 Al ]éin. OD Tube | % wt| <.04] <.004 | <.02|<.001 <.08[< .04 | <,04|<.04{<.08 <46 <.04 | <.04 <.04 <.08 < .04 3 <.04)<,2|<1 <.04 | <.04 |<1 <.3|<.08
3 Fe 1]4 in. Rod % wt| <.04| <.04 [<.004 |<.02( .001 <.08(<.04 | <.04|<.04(<.04T <.04 6 <.04 o1 <.04 08| <.04]<.2| <1 <.04 | <.04 |<1 <.3|<.08
4 Fe }é ine Rod % wt| <.04] <.04 |<.004 | <.02( .0O1 <.08(<.04 | <,04|<.04({<.04 < .04 b <.,04 o1 < .04 08| <.04]<.2| <1 <.04 | <04 (<1 <.3|<.08
5 Fe }é ine Tube % wt| < .04| <.,04T|<.004 | <.02| .001 <.08}<.04 | <.04|<.04|<.04 < .04 3 < .04 3 < .04 08| <.,04)<.2| <1 <.04 | <.04 |1 <3< .08
6 Ni 1 in« Rod % wt| < .04 08 | <.004 | <.02[ .001 <.08}<.04 6 | <04 .04 3 < .04 .3 < .04 < .04 .08 <.04|<,2|<1 <.04T| <.04 [ <.3(<.08
7 Ni 1 in. Tube % wt| <.04| <.,04T|[<.004 | <.02] .001 <.08! < .04 15| <.04| .04 .3 < .04 3 < .04 <.04 081 <.04| <.2| <1 <.L.04T| <.04 <1 <.3(<.08
8 $i0, Sand % wit{ <.04| <.04 [<.0004|<.02{<,001 <.,08[<.04 | <.04|<.04|<.04 < .04 <.01|<.01 {<.04 | <.02 <.04| <.01) <.08 < .04 <.02(<.2 <.04 | <.04 <.3]<.08
9 KF Powder % wt| <,04| <.04 [<.03 |<.04{<.001 <.08[<.04 | <.,04|<.,04(<.04T| <.04 04 | <.04T| <.04 <.04| <.1 <.08 <.04] <.08] <.04|<,2| <1 <.04 | <.04 (<1 <.3|<.08
10 NaF Powder % wt|<,04| <.04 |<.004 |<.02|<.001 <.08|<.04 | <.04]|<.,04|<,04 < .04 08[<.01TI< .04 | <.,02 <.04 < .08 < .04 08| <.02|<.1| <1 <.04 | <.04 (<1 <.3{<.08
11&12 Cr Powder % wt| <.04| <.04 | <.,004 | <.04|<.001 <.08(<.04 .04 .08 1 <W01<.01 1 <.04 | <.,04 <.04| <.01] <.08T| <.04 08| <.04|<.2 <.04 | <.04 <.3]<.08
13& 14 Mn Chips % wt <.04 | <.004 | <.02|<.001 <.08|<.04 | <.04|<.04|<.04 2 <..04T < .04 <.08 < .04 1 ] <.02)<.2( <1 <.04 | <.,04 <.3|<.08
15& 156 Co Solid Cylinders| % wt| < .04| <.04T| <.004 | <.02| <.001 <.08|<.04 <.04/<.04T| 2 <.04 3 <.04 6 <.04 3 | <L.04) <W2[ <1 <.04 [ <.04 |1 <.31<.08
Sinexdin.dia
17 302 Stainless| Tube 1]41 ins % wt| < .04 .04 | <,0005] <,02( <.,001 <,.05|<.002 .2 |18.4 .1 < .005 o7 .12 8.2 <.l 6 | <.02| <1 <5 <.02 02 | <.5] <.2]<.1
Steel
18 Inconel Tube 11’1 in, % wt| < .04 2 |<.002 | <.02{<.001 <,05|<,002 2 14771 14 6.41| <.01 .02 A4 <.02 78.14 23| <.05 <5 25 02 | <.5] <.2(<.1
19 BeO Blocks PPMI .1 | 275 1.6 1668 o7 154 378 636 7
20 C Powder % wt| < .04 b .004 04| <.001T 2 | <04 | <.04]<.04] L04 o3 04]<,01T| .08 | <.04T] <.04 2 | <.08T| <.04| 5 <.02|<.1| <1 .2 <4047 (<1 <3 1<.08
2] ZrO2 Powder PPM 100 <42 <.01 40 <.2 [<2 20 4 200 10 2 <10 100 <10T {100 <4 100 <4 <20

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

_69_