ORNL-1868.txt

 

ORNL-186 -
C-84 —~ Reaoctors-Special Featu i’o%‘?irfiaft Reactors
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AEC RESEARCH AND DEVELOPMENT REPORT

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H..J L\ N DISASSEMBLY AND POSTOPERATIVE EXAMINATION
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D 0N OF THE AIRCRAFT REACTOR EXPERIMENT
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Wy ’\: W. B. Cotirell
T. E. Crabtree
k:' : A. L. Davis
W. G. Piper

 

 

 

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OAK RIDGE NATIONAL LABORATORY
operated by
UNION CARBIDE CORPORATION
for the
U.S5. ATOMIC ENERGY COMMISSION

     
 

ORNL-1868
C-84 — Reactors-Special Features of Aircraft Reactors
M-3679 (20th ed., Rev.)

This document consists of 50 pages.
Copy}m 273 copies. Series A,

Contract No. W-7405-eng-26

AIRCRAFT REACTOR ENGINEERING DIVISION

DISASSEMBLY AND POSTOPERATIVE EXAMINATION

OF THE AIRCRAFT REACTOR EXPERIMENT

W. B. Cottrell A, L. Davis
T. E. Crabtree W. G. Piper
DATE ISSUED

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4 34 B e
AEE& T Adhder

 

OAK RIDGE NATIONAL LABORATORY
Ook Ridge, Tennessee
operated by
UNION CARBIDE CORPORATION
or the
U.5. ATOMIC ENERGY COMMISSION

  
 

      

MARTIN MARIETTA ENERGY SYSTEMS LIBRARI

- JITEEN T

3 4456 034984y 5

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 
 

     

 

 

 
 

TABLE OF CONTENTS

IR OAUCHION o v oo sscno s s o ioims s nines e TS A TR R BT TR B e R e 1
Chronology from Final Scram to Disassembly. . o o v v vn e vin i ittt . 4
Disassembly of the Reactor and Auxiliary Systems. ..o vve ity 5
Auxiliary Shielding and Sofety Precautions « v v vvv v vn i vinn vib ke e b sows D
Dismantling of the Sodium System . .. v v vt vi vttt 5
Dismantling of the Fuel System. . . v v v v s i tnneiie sttt nnanann 8
Dismantling of the Reactor. v v v v v v v v v st s s i st ie ettt taatoenasnas 8
Dismantling of the Fill and Drain Tanks « o v v v vvvnvn vttt as 16
Somples Taken During Disassembly. v v oo v v i iin it 16
Results of Examinations of Samples . ... .. s o o AR BT B R e e B 21
Beryllium Oxide Blocks ..o vvin i ST, I o S B Pl AR, e 21
Structural Materials and Valve Components . . .. ..o v § W RS e R il i o wonces BB
Nonmetallic Materials + v v v v v v v v v e v v i v o v o S R e R T A SR R 34
Balanoid NalNEE o s susssis siasrs sinivre bipbu oo B 8055 3 KEA & Gen s sl s o s S oees e 34
Fuel Pump Pressure Transmitters . ... oo ey o ami e BT Bl ¥ A 38
Scale from Main Sodium PUmp v v v e v v s s v e annosnsrsssssrsnnssansonse D il 38
Activation Analyses of Sodium System Samples . .. ... . . i i i 39
Disposition of Fission-Product Activity « .o vvvueveniinininiiiiiiiniiniaeaen 39
Disposal of Disassembled System. . ..o v evvvvvn o B RN B IR R NS R PN e SISTOR. |
{

Tr
 

 

DISASSEMBLY AND POSTOPERATIVE EXAMINATION OF THE
AIRCRAFT REACTOR EXPERIMENT

W. B. Cottrell
T. E. Crabtree

A. L. Davis
W. G. Piper

INTRODUCTION

The Aircraft Reactor Experiment (ARE) was
successfully concluded in November of 1954,
and a detailed report of the operation was
published the following year.! At that time
it was thought that an extensive examination
of the reactor and system components after
disassembly was warranted. It was realized,
of course, that the level of radicactivity of the

 

I'#. B. Cottrell et al., Operation of the Aircraft
Reactor Experiment, ORNL-1845 (Aug. 22, 1955).

o

ey
/ STANOBY FUEL PUMP

components would necessitate extensive delays
in the examinations.

Since examination of a few critical ARE samples
showed nothing unexpected, much of the planned
hot-cell inspection was postponed and complete
examination of all but a few specimens was
indefinitely suspended. The few examinations
that were completed are described in this report,
along with a description of the disassembly of
the ARE system. Diagrams of the fuel system,

sodium system, and off-gas system are presented
in Figs. 1, 2,and 3 for reference use in visualizing
the disassembly process.

       
  

 

KEY:
X0 CROSS BARS ON LINES INDICATE WELDS,
ALL FUEL PIPE LINES NUMBERED iN 100 SEMIES.

Fig. 1. Diagram of ARE Fuel System.

 
 

     

UNCLASSIFIED
ORNL-LA-DWG 6245

 
  
 

P

P / STANDBY SODIUM

 

TQ g

|l'-;‘l

  

KEY:

L ©x=3 CROSS BARS ON LINES INDICATE WELDS.
o ALL SODIUM PIPE LINES NUMBERED IN
300 SERIES.

Fig. 2, Diagram of ARE Sodium System.

 

 
 

  
 
  
   
 
  

PRESSURE THANSMITTER

S0DiUM VAPDR TRAP

 

RESERVE TANK WO |

CARRIER FILL TANK NO. 2

Fig. 3. Diagram of ARE Off-Gas System.

 
CHRONOLOGY FROM FINAL SCRAM TO DISASSEMBLY

The nuclear operation of the ARE was con-
cluded at 8:04 PM, Friday, November 12, 1954,
upon insertion of the scram rods. Circulation
of the fuel and the sodium was continued, how-
ever, until the next day in order to allow the
afterheat to decay before dumping the fuel into
the fuel dump tank (and the sodium into the
sodium dump tanks).

During the period from shutdown of the reactor
until the fuel was dumped it was necessary for
the operating personnel to wear gas masks for
several hours because of the level of the airborne
activity. As was previously noted,' a gas leak,
which permitted gaseous fission-product activity
to be released to the cell, had existed ot least
since the start of high-power operation. The
exact location of this leak has never been de-
termined, although it was known to be from
the gas volume above the main fuel pump. While
it was subsequently shown that the diaphragm
of the main fuel pump pressure transmitter was
ruptured, it is believed that this occurred during
the early morning of November 13 and that prior
to that time the activity had been leaking from
one or more of the potential leaks in or around
the main fuel pump (spark plugs, seals, fill
lines, vent lines, etc,). In order to keep these
gases from leaking from the cell into occupied
areas of the building (the cell proved to be quite
porous), it was necessary to maintain the cell
at a subatmospheric pressure (-2 to —4 in. H,0)
by the use of a jet air pump which pumped around
100 cfm of air from the cell. This air, with its
activity, was discharged some 1000 ft south of
the ARE building,! but the negligible wind
velocity that existed at times during the period
after shutdown was not sufficient to cause
odequate dispersion of this activity. (Prior
to this period the wind had dispersed the activity.)

On the morning of Saturday, November 13, the
fuel was transferred into the fuel dump tank.
Since the fuel tubes in the reactor would not
have drained completely if the fuel had merely
been allowed to drain from the system by gravity,
the fuel was forced out of the system by pres-
surizing carrier material into one end of the
system and draining the other end of the system

into the dump tank. The fuel in the dump tank
was thus diluted by this flush material. The
dilution ratio was approximately 1 to 1.

In order to determine whether the fuel in all
six parallel reactor passages was removed,
the flush carrier was heaoted to a temperature
which was 100°F hotter than the fuel system
so that the pdssage of the carrier through each
of the six parallel reactor tubes could be ob-
served by thermocouples on these tubes. Although
the individual recorder charts showing temperature
differentials across each of the six reactor tubes
indicated that one tube (No. 4) did not clear,
the multipoint temperature record of all tubes
indicated that the carrier had passed through
all tubes. The individual temperature recorder
on tube No. 4 was subsequently found to be
inoperative.

After the fuel was dumped, the increased level
of the airborne activity caused all but a few
of the operating personnel who were wearing
gas masks to evacuate the building for about
1 hr. The gas used in pressuring the fuel system
during the dumping operation was discharged
to the stack, but the activity level in the building
rose because the wind at the time was such
that the activity descended and entered the
ventilators on the top of the building.

Following the completion of the removal of
the fuel on Saturday, November 13, the sodium
was drained into the sodium drain tanks. |In
this instance gravity drainage was sufficient
and was readily effected.

Evidence that all the fuel had been removed
from the reactor was obtained the following
Monday by a test in which the three control rods
were withdrawn from the reactor and no increase
in background count was observed. It was of
especial interest that there was no measurable
afterheat in the fuel; however, the expected
amount of afterheat was only o small fraction
of the electrical heat on the tank. Subsequent
analysis of the fuel indicated the activity to
be in reasonable agreement with expectations.
 

DISASSEMBLY OF THE REACTOR AND AUXILIARY SYSTEMS

The postoperative examination of the ARE system
started on December 10, 1954, with the taking of a
fuel sample from the dump tank while the fuel
was still molten. Disassembly of the reactor
and the auxiliary systems then proceeded os
radiation levels permitted. The primary objective
of this work was, of course, the obtaining of
samples for metallurgical, chemical, and physical
examination. It was also expected that in the
dismantling of the system much equipment could
be salvaged while the cells were being prepared
for the modifications required for the forthcoming
Aircraft Reactor Test.

Radiation surveys were used as a basis for
planning the disossembly sequence and techniques.
The radiation decay curves obtained from data
taken daily ot five monitoring points are shown
in Fig. 4.

AUXILIARY SHIELDING AND SAFETY
PRECAUTIONS

In order to separate the fuel circuit from the
sodium circuit for the purpose of dismantling
the sodium system, two flat lead shields, 6 ft
high, 4 ft wide, and 2 in. thick, were suspended
on beams which ran the width of the heat ex-
changer cell. For work in the higher radiation
fields associated with the fuel system, a leod
box was built that was 2 in. thick on four sides
and the bottom. One end of the bottom had a
6-in.-wide slot, and one side had an 8-in.-square
slot, 9 in. off the bottom, through which personnel
worked. The building crane was used for moving
the flat shields and the box.

In order to offset the fire hazard associated
with the sodium system, all water lines were
cut from the water manifold in the basement below
the heat exchanger cell, and no sodium lines
were cut with flame or arc torches. All sodium
lines were cut with hack sows, and all cut
lines were immediately sealed with several
layers of masking tape. Fire-fighting equipment
was available at all times.

DISMANTLING OF THE SODIUM SYSTEM

The work of dismantling the sodium system,
shown in Fig. 2, was started on Jonuary 18, 1955,
at which time the radiation level was down to
30 to 250 mr/hr with the lead shields hanging
between the sodium system and the fuel system.

The main sodium pump was removed first,
ond it was found that the rotary element had a
radiation reading of 12 mr/hr at contact with
the impeller. After the rotary assembly had
been cleaned in a bath of 50% kerosene and
50% methyl alcohol, the impeller was removed
and submitted for examination.

The standby sodium pump was removed as @
unit with 6-in. stubs left on all lines so that
the pump could be salvaged for use in other
experiments, At the time it was removed, this
pump, which had not operated during nuclear
operation of the reactor, had a radiation reading
of 1 mr/hr at contact with the bowl and the
top flange.

The sodium-to-helium and the helium-to-water
heat -exchangers were removed next. The water
that had remained in the heat exchangers was
first drained* into containers and removed from
the cell. The sodium lines were then cut with
a hand hack saw and the helium blower duct was
cut loose with a cutting torch. The exchangers
were thus removable as complete units. After
the insulation and electric heaters had been
stripped from the exchangers, neither showed
a radiation reading on the outside. The ends
of the cut sodium lines read 2 mr/hr ot contact.

After the removal of the sodium system pumps
and heat exchangers, it was found that the
radiation field had increased to about 600 mr/hr.
The equipment had helped to shield the area
from the fuel system radiation.

Sodium lines 304 through 309, 313, and 314
were then removed in as long lengths as possible
and sealed at the cut ends from air and moisture.
Valves in the lines were left as installed. Radi-
ation levels on these lines were 2 to 10 mr/hr.
During the removal of these sodium lines the
flat lead shields were adequate to protect
personnel from the radiation from the fuel system,
as shown in Fig. 5. The remainder of the sodium
piping ran adjacent to the fuel system, however,
and, in places, over and under the fuel piping,
and therefore the lead box described above had
to be brought into use. Sodium lines 301 and
302 were removed by personnel working within
the lead box. To moke the line cutting job
easier, the heavy gage stainless steel annulus
can surrounding the sodium line was cut by using
an electric arc before the sodium line was cut
e
ORNL~LR-DWG 27129

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

100
50
20 : \
10 e
-""'--.._
LOCATION OF MONITOR

RADIATION LEVEL (¢/hr)

 

 

 

 

 

 

 

AT CENTER OF SOUTHWEST SIDE OF THE HEAT EXCHANGER CELL

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

i- AT THE ELEVATION OF THE MAIN SODIUM PUMP.
» AT CENTER OF SOUTHEAST SIDE OF THE HEAT EXCHANGER CELL
5 - ABOVE THE STANDBY SODIUM PUMP.
3. AT SODIUM - TO- HELIUM HEAT EXCHANGER NUMBER 1 IN THE 2
" STANDBY SODIUM CIRCUIT.
4. AT THE CEILING OF THE NORTH END OF THE REACTOR CELL.
5 THREE FEET NORTH OF THE MAIN FUEL PUMP ON TOP OF LINE 19, .
"WHICH WAS THE EXIT LINE FROM THE STANDBY FUEL PUMP,
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0 5 10 {45 20 25 30 35 40 45 50 55 60 65 70

DAYS AFTER NOVEMBER {2, 1954

Fig. 4. Radiation Decay Curves for Five Locations in ARE System.
 

LUINCL ASSIFIEL
PHOTO 14182

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" N T ]

 

Fig. 5. Heat Exchanger Pit After Removal of Sodium System Pumps and Heat Exchangers and Lines, Note lead shield used to protect workers from

fuel system radiation.
with a hack saw. The radiation level at the open
ends of lines 301 and 302, which were cut where
they entered the tank pit wall, was only 2 mr/hr,
but the radiation field in that region outside the
lead box was 700 to 1000 mr/hr.

Sodium lines 303 and 310, the lines to and
from the reactor, were removed next. Because
of the position of line 303, it was necessary
to sever it where it entered the reactor cell
The cutting
which was

wall by wusing an electric arc.
operation caused a sodium fire,
extinguished with Metal-X extinguishers. A small
amount of air activity, 2 divisions on the 2K
scale of the air monitor, was observed as o
consequence of the sodium fire. Both line 303
and line 310, which was cut with a hack saw
by a craftsman werking from the lead box, showed
radiation levels of 2 mr/hr at contact with the
ends of the lines.

Auxiliary equipment was removed as necessary
or convenient. Most of these items, such as
pump drive motors and lubricating oil systems,
were salvaged for further use. The radiation
level of the main sodium-pump lubricating system
was 2 mr/hr, but this was found to be surface
contamination. The oil showed no radiation,

DISMANTLING OF THE FUEL SYSTEM

A complete radiation survey of the fuel system,
made on February 14, 1955, showed 75 r/hr at
contact with the insulation on the rotameter of
the main fuel pump, 55 r/hr ot contact with
insulation on the dead leg on the bottom of the
pump bowl, and 12 r/hr ot contact with the top
flange. These high radiation levels indicated
the advisability of removing the high radiation
sources first, where possible.

The components of the fuel system are illus-
trated in Fig. 1, and a photograph of the cell
containing the fuel pumps is presented in Fig. 6.
The disassembly work was started with the
main fuel pump which was causing a high radiation
field. Removal of the rotary assembly caused
a small burst of air activity that cleared in 2
to 3 min. The radiation level of the rotary
assembly was 20 r/hr ot 5 ft.  After removal
of the rotary assembly, the lines to the pump
were cut and sealed and the pump was lifted
out of the cell, as shown in Fig. 7. The radiation
level of the main fuel pump bow!l was 900 mr/hr
at 5 ft. The standby fuel pump was removed in
a similar manner. This pump was not used during

operation at power, and therefore the radiation
level on the bottom of the pump bowl was only
16 mr/hr. This pump was salvaged for further
use.

The fuel-to-helium and helium-to-water heat ex-
changers were removed next. The radiation levels
of both exchangers were 10 r/hr at 6 in. In order
to remove these exchangers intact, the water
lines and the helium ducts were cut. The electric
wiring which ran in large bundles across the
heat exchangers was removed by hooking onto
the bundles with the overhead crane and pulling
the wires out of the cell. The fuel lines to the
heat exchangers were cut by a craftsman working
from the lead box with a hack saw. The insulation
and the electric heaters were removed from the
heat exchangers after they were removed from
the cell, oand heat exchanger No. 2 then had
a radiation level of 10 r/hr at 12 in.

During removal of the pumps and heat ex-
changers, most of the fuel lines were cut at
one end., The open ends of these fuel lines

had radiation readings of 10 r/hr at 3 in.

Lines in the standby fuel circuit were free
of fuel and showed only surface contamination
of 2 mr/hr.  The reactor supply and discharge
lines were cut with hack saws by craftsmen
working within the lead box. With the removal
of the fuel lines, all equipment installed in the
heat exchanger cell had been removed.

DISMANTLING OF THE REACTOR

On March 10, 1955, the concrete pit plugs were
removed from the top of the reactor pit and the
concrete blocks which lined the walls and bottom
of the reoctor cell were lifted out to provide
space for the lead box. A portable grinder with
a flexible shaft and an extension handle was
mounted on the side of the lead box. The crafts-
man could operate the grinder from inside the
box by using the extension handle.

The insulation was stripped from the reactor
and the sodium and fuel lines that were in the
reactor cell. The helium manifold and the control
rod chamber (Fig. 8) were then removed. The
helium manifold and the shim and regulating rod
chambers had radiation readings of 200 mr/hr
at contact.

The portable grinder was then used to cut
the fuel lines as close as possible to the reactor
can.  These lines had radiation readings of
10 r/hr at 3 in. when removed. The grinding
- " UNCLASSIFIED
PHOTO 14191

 

e et 2 B

B

- T _ PP

 

 

 

Fig. 6. Heat Exchanger Pit Showing Components of Fuel System Before Dismantling Operation,
0l

Fig. 7. Main Fuel Pump Being Lifted from Cell.

UNCLASSIFIED
PHOTO 14205

 
 

 

€

T
.o 1 -"‘!'*T

  
     
   
     
  
   
      

HELIUM /
RETENTION TANK

VIEWING PORT—_ ||

 

MG

 

'~
\\\\\\\

xxxxx

 

Fig. 8. Diagram of Reactor.

 

UNCLASSIFIED
ORNL-LR—DWG 6221

SODIUM CUTLET
PRESSURE SHELL

THERMAL SHIELD

GAS TIGHT SHELL

SAFETY CHAMBER NQ. 2
MICROMICROAMMETER

-

LOG N METER

8F, CHAMBER AND
TEMPERATURE SERVO MECHANISM

SAFETY CHAMBER NO.{

11

 

 
operation generated so much heat that it was
necessary, as before, to cut the sodium lines
with a hack saw. The sodium lines had radiation
readings of 300 mr/hr at contact.

The gas-tight shell surrounding the reactor
was then cut in quarter sections by using an
The can had a radiation

electric arc (Fig. 9).

 

 

level of 200 mr/hr at contact. The exposed reactor
pressure shell showed 6 r/hr at 6 in.

Special disassembly tools were fabricated for
dismantling the reactor. This equipment included
a steel tank with driven rollers for rotating the
reactor and a hydraulic-powered cutoff grinder
with a 14-in.-dia abrasive wheel. The reactor

. L
’ PHOTO 14186

Fig. 9. Gas-Tight Shell Being Removed from Reactor.

12
is shown in Fig. 10 positioned in the tonk for
grinding, and the cutting side of the grinder,
which was designed to be used under liquid,
is shown in Fig. 11. With this equipment, a
period of five 8-hy days was required to cut
the 2-in.-thick pressure shell. Upon removal
of the shell,- a radiation check was made. The

 

top of the reactor showed a radiation level of
75 r/hr at contact and the side showed 37 r/hr.

The next task was the removal of fuel tubes
and samples of the BeQO moderator blocks. The
fuel tubes and the BeO blocks were encased
in a stainless steel can with welded seams,
shown in Fig. 12, and the small sodium tubes

A : PHOTO 14655
. ¥
L P

Fig. 10, Reactor Positioned in Disossembly Tank for Grinding.

13
I UNCLASSIFIED
-  PHOTO 14659

 

Fig. 11. Cutting Side of Grinder for Dismantling the Reactor.
662

PHOT

— -

 

Fig. 12. Top of Reactor After Removal of Pressure Shell and Some of the Sodium and Fuel Tubes.

15
that penetrated the BeO blocks were welded to
the top and bottom of the can. In order to free
the top of the can, the welds holding the sodium
tubes were drilled out and the outer edge of
the can was cut. The reactor was then placed
in a horizontal position and the fuel tube bends
were cut off flush with the bottom of the can.
After removal of the bottom tube bends, the
reactor was set upright and the crane was used
to pull out the fuel tubes and to place them in
drums, as shown in Fig. 12.

With all the fuel tubes removed, only the BeO
blocks remained in the can. The required samples
of BeO were obtained, and the can containing
the BeO blocks was taken to the storage area.
The radiation reading at that time was 4.5 r/hr
at 2 ft. The reactor cell was then cleared of
all contaminated material left from disassembly
of the reactor.

DISMANTLING OF THE FILL AND DRAIN
TANKS

The fuel dump tank was removed first because
it presented the only radiation in the tank pit,
which also contained three sodium tanks and
three fuel-carrier tanks. (One of the fuel-carrier
tanks was not used.) A radiation survey showed
18 r/hr at contact with the top of the tank,

990 r/hr in one of the cooling tubes through
the tank, and 84 r/hr at contact with msulntwn
on the side of the tank.

In order to get the lead box into position
for disassembly work, it was necessary to remove
steel framework, o space cooler, and the stack
at the top of the tank. With the lead box in
position, the fuel lines were cut with the grinder
and then toped to prevent loss of material and
the spread of contamination. The tank was then
removed to a storage area. After the dump tank
and fuel lines had been removed, the radiation
level in the pit was 20 to 50 mr/hr.

The lines to the three fuel carrier tanks and
the three sodium tanks were then cut, and these
tanks were removed from the pits with insulation
and heaters intact. After removal of insulation
and heaters, the sodium tanks were found to
be free of radiation. One of the fuel carrier
tanks, however, had a radiation level of 100 mr/hr
and was therefore sent to storage. The sodium
and the fuel carrier were salvaged from the non-
radioactive tanks.

The tank pit was then cleared of small gas
lines, valves, and wiring, and the dismantling
of the tank pit was complete. General cleaning
of the entire area then completed the disassembly
of the ARE reactor and auxiliary systems.

 

SAMPLES TAKEN DURING DISASSEMBLY

Many samples were token during disassembly
of the ARE, and much re-usable equipment was
salvaged. The samples taken are listed in
Table 1, which also gives the type of examination
requested and the status of the samples.

Sodium system samples were taken after the
components had been removed from the cells.
After all the samples had been taken, the re-
maining equipment that could not be salvaged
was cleaned of sodium by immersing it in water
until the sodium-water reaction was complete.
Most of the equipmént cleaned in this manner
was loter disposed of in the burial ground because
of radiation levels of about 2 mr/hr.

Fuel system and reactor samples were taken
in the reactor pit after the reactor had been dis-
assembled and removed, except the sample from
fuel supply line No. 120, which was taken before
dismantling was storted. After removal of the

16

reactor, a horizontal cutoff saw was set up in
the reactor pit for cutting the samples, which
were moved to the reactor pit from the radicactive
storage facility for sampling. All work involved
in obtaining the fuel system and reactor samples
had to be done from within the lead box with
the use of improvised long-handled equipment,
the portable grinder used for disassembly work,
and the cutoff saw.

A sample of the fuel was taken from the fuel
dump tank while the fuel was still molten. In
order to obtain the sample, o hole had to be
drilled in the top of the dump tank. Since the
tank was at a temperature of between 1200 and
1300°F and the radiation level was high, the
drilling had to be done from atop the cell plugs.
A pipe extension which would reach to the top
of the tank was adapted to the drill chuck, and
the drill bit was welded to the pipe. Several
Ll

Taoble 1. List of Samples Taken from the ARE During Disassembly

 

 

Sample Type of Examination
M. Description Requested Status of Examination
Sodium System Samples
s Section of moderator coolant return line 303 Metallographic Sample to be examined
S2 Section of moderator coclant supply line 310 Metallographic Somple to be examined
s3 Section of main No pump impeller Metallographic Sample to be examined
S4 Seat from sodium valve U-23 Sterecphotographic Results of examination given in this report
S5 Plunger from sedium valve U-23 Stereophotographic Results of examination given in this report
S6 Bellows frem sodium valve U-23 Metallographic Results of examination given in this report
57 Bellows from sedium pressure transmitter Metallographic Sample to be examined
in line 303
58 Section from cold dead leg No. 931 Metallegraphic Request rescinded, CF-56-6-24; sample to be disposed of
59 Section from cold dead leg No. 932 Metallographic Request rescinded, CF-56-6-24; sample to be disposed of
Sample of scale in main Na pump Spectrographic analysis Spec Lab Report No. 3450, and this report
Section of line 308 to main Na pump Activation analysis CF-55-10-35, and this report
Section of line 305 to Na pressure Activation analysis CF-55-10-35, and this report
transmitter No. 2
Sample of sodium Activation analysis CF-55-10-35, and this report
Fuel System Samples
F1 Section of fuel supply line 120 Metallographic and CF-55-2-58, CF-55-2-36, and this report
activation
F2 Section of fuel return line 111 Metallographic Request rescinded, CF-56-6-24; sample to be disposed of
F3 Section of main fuel-pump impeller Metallographic Request rescinded, CF-56-6-24; sample to be disposed of
Fd-1 Main fuel-pump bowl Visual Sample examined visually and found to be in goed condition;
sample to be disposed of
F4-2 Main fuel-pump bottom baffle plate Visual Sample examined visually and found to be in goed condition;
sample to be disposed of
F4-3 Main fuel-pump socket-weld joint Visual Sample examined visually and found to be in good condition;
sample to be disposed of
F4-4 Main fuel-pump flange outside bowl Visual Sample examined visually and found to be in good condition;
sample to be disposed of
F4-5 Main fuel-pump spark plug Visual Sample examined visually and found to be in good condition;
sample to be disposed of
F4-6 Main fuel-pump channel Visual Sample examined visually and found to be in good condition;
sample to be disposed of
F5 Section of inlet to fuel heat exchanger Metallegraphic Results of examination given in this report

No. 2

 

 

 
8l

Table 1. (continued)

 

 

Sample Type of Examination
o Description Rsquested Status of Examination
Fuel System Samples
Fé Section of inlet to fuel heat exchanger Metallographic Request suspended, CF-56-6-24; sample to be held in
No. 2 storage
F7 Section of outlet to fuel heat exchanger Metallographic Request suspended, CF-56-6-24; sample to be held in
No. 2 storage
F8 Section of outlet to fuel heat exchanger Metallographic Request suspended, CF-56-6-24; sample to be held in
No. 2 storage
F9 Section of middle of bends in heat ex- Metallographic Results of examination given in this report
changer No. 2
F10 Section of middle of bends in heat ex- Metallographic Request suspended, CF-56-6-24; sample to be held in
changer No. 2 storage
F11 Seat from fuel valve U-1 Stereophotographic Results of examination given in this report
F12 Plunger from fuel valve U-1 Stereophotographic Results of examination given in this report
F13 Bellows from fuel valve U-1 Stereophotographic Request suspended, CF-56-6-24; sample to be held in
storage
Fl14 Seat from fuel valve U-2 Stereophotographic Request suspended, CF-56-6-24; sample to be held in
storage
F15 Plunger from fuel valve U-? Stereophotographic Request suspended, CF-56-5-24; sample to be held in
storage
F16 Bellows from fuel valve U-2 Stereophotographic Request suspended, CF-56-6-24; somple to be held in
storage
F17 Tapered plug from main fuel rotameter Metallographic Request rescinded, CF-56-6-24; sample to be disposed of
Sample of fuel from dump tank Activation analysis CF-55-1-128 (results analyzed in CF-55-2-36), and this
report
Hot fuel dump tank with fuel Fuel to be reprocessed See ANP Quar. Prog. Reps.
Reactor Samples
R1 Section of serpentine fuel tube close Metallographic Request rescinded, CF-56-6-24; sample to be disposed of
to pressure shell
R2 Section of serpentine at }'2 radius Metallographic Request rescinded, CF-56-6-24; sample to be disposed of
R3 Section of serpentine in center of core Metal lographic Results of examination given in this report
R4 Serpentine bend from region near outer Metallegraphic Request rescinded, CF-56-6-24; sample to be disposed of
surface
R5 Socket weld joint near center Metallographic Request rescinded, CF-56-6-24; sample to be disposed of
Ré6 Socket weld joint near outer surface Metallographic Request rescinded, CF-56-6-24; sample to be disposed of
R7 Serpentine bend from center of reactor Metallographic Results of examination given in this report
RB Sample of pressure shell wall Metallographic Results of examination given in this report
61

Table 1. (continued)

 

 

Sample Type of Examination
Kk Description Requested Status of Examination
Reactor Samples
R9 BeO block from outer region of core Metallographic CF-56-6-113, and this report
R10 BeO block from central region of core Metallographic CF«56-6-113, and this report
R11 BeO block cut prior to installation Metallographic CF-56-6-113, and this report
R12 BeO block cut prior to installation Metallegraphic CF-56-6-113, and this report
Reactor inlet mainfold Metallographic Sample in storage to be disposed of
Reactor outlet manifeld Metallographic Sample in storage to be disposed of
Miscellaneocus Samples
M1 U-belts from main fuel pumps Visual CF-55-7-27, and this report
M2 O-ring from main fuel pumps Visual CF<55-7-27, and this report
M3 Qil sample from main fuel pump Vi;uul plus viscosity CF-55-7-27, and this report
check
M4 Qil sample from standby Na pump Visual plus viscosity CF-55-7-27, and this report
check
M5 Oil sample helium blower Visual CF-55-7-27, and this report
M6 Electric wire from main fuel pump Visual CF.55-7-27, and this report
region
M7 Thermocouple wire from main fuel pump Visual CF«55-7-27, and this report
region
M8 Thermal insulation from main fuel pump Visual CF-55-7-27, and this report
region
M9 Concrete from wall near main fuel pump Activation anal ysis Wall since decontaminated

Rubber diaphragm from Hzfi valve B139

Thermal insulation wrapping paste on

line 303
Thermal insulation on reactor
Thermocouple lead wire at reactor
Braided thermocouple wire on top of reactor
Main fuel pump vapor trap and lines

Standby fuel dump tank pressure trans-
mitter 7

Main fuel dump tank pressure transmitter 6

Hot fuel dump tank pressure transmitter 10

Visual plus viscosity
check

Visual

Visual
Visual
Visual
Visual

Performance test and
visual

Performance test and
visual

Performance test and
visual

CF-55-7-27, and this report
CF-55-7-27, and this report

CF-55-7-27, and this report
CF-55-7-27, and this report
CF-55-7-27, and this report
Results of examination given in this report

Results of examination given in this report
Results of examination given in this report

Not examined; sample to be disposed of

 

 
0C

Table 1. (continued)

 

Sample
No.

Description

Type of Examination

Requested

Stotus of Examination

 

Main fuel pump vent solencid U-19

Main fuel pump vent solenoid U-86

Main fuel pump He supply seclencid U-82
Standby fuel pump vent solenoid U-185
Main fuel blower He supply solenoid B-233

Pit-activity=monitor isolation solencid

B-418

Off-gas system monitron inlet solencid
B-124

Off-gas system monitron inlet solenocid

B-125
Standby fuel pump vent solencid U-89

Standby fuel pump emergency vent solenocid
U-159

Main fuel pump emergency vent solencid
U-10

Sample of inlet emergency off-gas line

Sample of off-gases from cell

Miscelloaneous Samples

Leak test and visual
Leak test and visual
Leak test and visual
Leak test and visual
Leak test and visual

Leak test and visual
Leak test and visual
Leak test and visual

Leak test and visual

Leak test and visual
Leak test and visual

Activation analysis

Activation analysis

Results
Results
Results
Results
Results
Results

Results
Results

Results

Results

Results

of examination given
of examination given
of examination given
of examination given
of examination given

of examination given
of examination given
of examination given

of examination given

of examination given

of examination given

CF-55-2-36, and this report

Undocumented memo 12-22-55, results described in

CF-55-2-36

in this report
in this report
in this report
in this report
in this report

in this report
in this report
in this report

in this report
in this report

in this report

 

 
drill bits, made of different metals, were tried
before the hole was finally completed. To remove
the fuel sample, a 3/8-in.-dia Inconel tube was
passed through the opening in the tank, and a
small vacuum pump was used to pull fuel into
the tube. The tube was then removed from the
tank ond allowed to cool. A section of the tube

containing fuel
desired sample.

In addition to the sodium system, fuel system,
and reactor samples, numerous samples were
taken from other parts of the various systems
during the dismantling operations. The samples
token are described in Table 1.

was then cut to provide the

 

RESULTS OF EXAMINATIONS OF SAMPLES

BERYLLIUM OXIDE BLOCKS?

Beryllium oxide blocks were taken for exami-
nation from the outer region, the central region,
and the core region of the ARE. The stacking
arrangement may be seen in Fig. 13, which shows
the top of the reactor during assembly. The blocks
surrounding the serpentine fuel tubes were split
to facilitate assembly, but the blocks with small
holes for sodium-coolant tubes were not cut. The
small spaces between the blocks were filled
with slowly moving sodium.

The blocks that were removed are shown in
Figs. 14, 15, and 16. It was found that draining
the reactor had left the surfaces of the BeO
blocks essentially free of sodium. Since it was
not necessary to strip sodium from the blocks
it could be assumed that any damage that was
found during the initial examination had occurred
during operation or was present in the as-tab-
ricated material. The post-test handling could
not have caused further damage.

The block shown in Fig. 14, which had sur-
rounded a fuel tube in the central region of the
reactor, has many visible cracks and one half of
the block had fractured. Previous tests of the
as-received blocks had revealed that nearly all
the blocks had ot least one crack and some of
the blocks had several cracks.® Comparison
with photographs of the as-received blocks showed
that slight erosion of the edges occurred during
reactor operation. The presence of small flakes
of BeO adhering to the surface of the block
indicated some spalling.

 

2Materiol abstracted from a report by R. J. Gray and
L. Long, Jr., Examination of BeO Blocks from the
ARE URNL CF~56-6-113 (June 18, 1956).

3‘L. M. Doney, Structure of BeO Block with the 1Y g1
Central Hole, RNL CF-52-11-146 (Nov. 17, 1952).

A BeO block that surrounded a fuel tube in
the outer region of the reactor is shown in Fig.
15. The droplets visible on the surface are
sodium hydroxide produced from sodium that
remained after removal from the reactor. There
were no complete fractures of this block, but
many cracks are visible.

Three cut blocks from the core are shown in
Fig. 16, Two of the three blocks fractured
during operation of the ARE; however, no cracks
are visible.

Only the BeO block taken from the outer region
of the core (Fig. 15) was examined to determine
the depth of penetration of the sodium. The
dark areas in the transverse sections shown in
Figs. 17 and 18 indicate high sodium concen-
trations in porous material, whereas the light
areas, which were not penetrated by sodium,
are dense material. The dark lines indicate
sodium in cracks and crevices. As may be
seen, the core of the block is more porous than
the periphery. Cracks did not propagate through
the dense material. In general, the BeO blocks
withstood the operation of the ARE reasonably
well.

The remainder of the blocks were stored in
place in the reactor. When final disposition of
the reactor was to be made in October 1957, it
was found that, as a result of atmospheric action
on the adsorbed sodium, the blocks had deteri-
orated to the point where they had lost their
structural strength. Since the blocks could serve
no further useful purpose, they were buried along
with the reactor. The results of attempts to
remove blocks for examination prior to the
disposal decision are shown in Figs. 19 and
20, and the reactor may be seen in Figs. 21,
22, and 23 just prior to burial.

21
UMCLASSIFIED
PHOTO 11208

22

 

(Secret with caption)

Fig. 13. ARE Reactor Core During Assembly.
UNCLASSIFIED
Y-15743

 

Fig. 14. BeO Block Which Had Surrounded a Fuel Tube in the Central Region of the Reactor.

UNCLASSIFIED
Y-15742

-

 

Fig. 15. BeO Block Which Had Surrounded o Fuel Tube in the Outer Region of the Reactor.

23
UNCLASSIFIED
Y-15744

 

UNCLASSIF IED
Y-18872

Fig. 17. Transverse Section of Block Shown in Fig. 15.

24
 

UNCL ASSIFIED
Y-18873

 

 

UNCLASSIFIEL
PHOTO 40727

Y :-.-r*J - .

h s
T ke

   

Fig. 19. Condition of Reactor When Removed from Storuge in October 1957. The broken pieces of BeO
blocks on the floor resulted from ottempts to remove blocks for examination.

< - Vo, ¥

25
26

Fig. 20. View of Deteriorated BeO Blocks in Reactor.

    

Fig. 21. Reactor Being Removed from Storage for Transfer to Burial Ground.

UHCL ASSIFIED
PHOTO 40738

 

UNCLASSIE ED
PHOTO 41773
. i

 

Fig. 22. Top View of Reactor During Removal from Storage.

UNCLASSIFIED
PHOTO 41778

URCLASSIFIED
PHOTO 41779

 

 

 

Fig. 23. Bottom View of Reactor Ready for Transfer to Burial Ground.

 

27
STRUCTURAL MATERIALS AND VALVE
COMPONEMNTS

Specimen F1 from fuel supply line 120 was
found? to have general surface attack to a depth
of 1 to 2 mils (Fig. 24), and the surface was
rough, as shown in Figs. 25 and 26. Stereo-
photographs were taken of the seats and plungers
from valves U-1 and U-23 (specimens F11, F12,
S4, and S5).°> As shown in Figs. 27 aond 28,
there were dark deposits on the Stellite seat
and plunger from valve U-1. The Stellite plunger
from valve U-23 was scored, as shown in Fig.
29: the seat of valve U-23 is shown in Fig.
30.

Metallographic examinations were completed of
specimens R3, R7, R8, F5, F9, and $6.% Specimen
R8, a section from the pressure shell wall, in-
cluding a weld, is shown in Fig. 31. A crack

 

M. J. Feldman, ARE — Line 120 (lnner Surface),
ORNL CF-55-2.58 (Feb. 11, 1955).

5A. E. Richt et al., ANP Quar. Prog. Rep. June 30,
1957, ORNL-2340, p 267.

SA. E. Richt et al, ANP Quar. Prog. Rep. Sept. 30,
1957, ORNL-2387 (in press).

 

may be seen at the weld junction, and there
are several voids in the weld area. A section
from a serpentine bend in a fuel tube in the
center of the reactor, specimen R7, is shown
in Figs. 32 and 33.
formation on the interior wall to a maximum depth
of 3.5 mils, but the penetration wes not uniform.
Some parts of the wall showed deeper and more
dense penetration than others. The outer wall
of the serpentine bend, which was in contact
with sodium, showed what appeared to be a mass
transfer deposit (Fig. 34) in addition to some
subsurface wvoid formation. The deposit had
plated on the wall to a maximum thickness of

There was subsurface void

1 mil.
Specimen R3, which was also taken from a
serpentine bend in a fuel tube in the center

of the core, also showed some subsurface void
formation; however, the density and depth of
penetration were not so great as for specimen
R7. The areas of attack were localized, and
some areas of the wall showed no attack, as
may be seen in Figs. 35 and 36. MNo deposit
similar to that found on specimen R7 was noted
on the exterior wall.

UNCL ASSIFIED
RMG-1028

Fig. 24. Inner Surface of Fuel Supply Line 120 (Specimen F1). Unetched. 250X.

28
UNCLASSIFIED
R

MG-1046

 

 

 

o
i
i

1 I »
IINCII-iES l

8

oor

 

 

 

 

010

 

011 / o 3

 

 

| Ao , .
| Lotz B o~
& s

e | e

 

 

 

UNMCL ASSIFIED
RMG-1047

 

 

 

 

 

10

 

011

 

 

| e ‘

Fig. 26. Another Section from Specimen F1. Etchant: electrolytic oxalic acid (10%). 250X.

 

-

 

 
4

UNCLASSIFIED
RMG-1

L

g !i_.\.

 

Fig. 27. Plunger from Yalve U-1, 5X.

RMG-1735

o
w
L
oy
V)
<<
-l
o
Z
=

 

 

Fig. 28. Seat from Valve U-1. !5)(.

30
 

-

[ AN
.b e ""z“-"r
B,

 

    

MG-1736

YSEUNCLASSIF IED
! IR F

  

'_-..

Fig. 29. Plunger from Valve U-23, 5X.

UNCLASSIFIED
RMG-1737

Fig. 30. Seat from Volve U-23. 1.5}(.

31
32

 

UNCLASSIFIED

Fig.

31. Section from ARE Pressure Shell, In-
cluding a Weld. 5X. (Secret with caption)

Fig. 32.

Section from a Serpentine Bend in a Fuel
Tube in the ARE Core, Specimen R7.
250X. (Secret with caption)

As polished.

 

Fig. 33. Section Shown in Fig. 32 After Etching. 250X, (Swwssewith caption)

 
UNCLASSIFIED
RMG-1828

 

Fig. 34. Mass Transfer Deposit on Outer Wall of Fuel Tube at Serpentine Bend. This surface was in contact
with sodium during ARE operation. As polished, 500X. (%™9W8® with caption)

UNCLASSIFIED
RMG-1829 |

 

o
& -
® . \ .
®
- " -
o = 0y - -
' 8]
» - »
Coamy
- - @
.
o

Fig. 35. Section of Specimen R3 Taken trom Serpentine Bend in Fuel Tube in ARE Core. Etched. 250X. (mmmgd

with caption)

33
 

UNCLASSIFIED
RMG-1830

Fig. 36. Another Section of Specimen Shown in Fig. 35. Etched. 250X, (@wmmmg with caption)

Specimen F5, which was taken from the inlet
of a fuel-to-helium heat exchanger, was cast
in epoxy resin so that the fins on the tube would
not be damaged during cutting. The fin-to-tube
wall junction is shown in Fig. 37. (Iln the ARE
fuel-to-helium heat exchangers the fins were
helical strips placed in grooves on the tubes.
The strips were not brazed to the tubes.) The
interior wall of the tube showed subsurface
voids to a depth of 4 mils, as shown in Figs.
38 and 39. Specimen F9, which was taken from
the middle of a bend in a fuel-to-helium heat
exchanger, also showed subsurface voids to a
depth of 4 mils, as shown in Figs. 40 and 41.
Specimen S6, the bottom bellows from sodium
valve U-23, was also cast in epoxy resin. No
cracks in the bellows folds were found, and
only a slight roughening of the inside surface
was noted, as shown in Fig. 42.

NONMETALLIC MATERIALS

Several nonmetallic specimens taken from the
ARE were examined for radiation damage effects.’
Viscosity measurements were made on the oil
specimens, but all other specimens were examined
visually and compared with unirradiated speci-
mens. No changes that could be attributed to
irradiation were noted in any of the materials.

 

?U. Sisman, Radiation Damage to ARE Nonmetallic

fiazerr'lafs, ORNL CF-55-7-27 (July 7, 1955).

34

The specimens examined included thermal in-
sulation, electrical insulation, rubber, and oil.

SOLENOID VALVESS

The solenoid valves removed from the ARE
were examined in simulated service tests and
were found to be completely leaktight. The
valve seats were found to be in poor condition
as a result of dirt and filings that were deposited
during the disassembly process. The valves
examined are listed below:

Valve Mo, Description
U-10 Main fuel pump emergency vent valve
u-19 Main fuel pump vent valve
U-82 Main fuel pump helium supply valve
U-83 Standby fuel pump helium supply valve
U-86 Main fuel pump vent valve
U-89 Standby fuel pump vent valve
U-159 Standby fuel pump emergency vent valve
U-185 Standby fuel pump vent valve
B-124 Off-gas system monitron inlet valve
B-125 Off-gas system monitron inlet valve
B-418 Pit activity-monitor isolation valve
Radiation measurements showed only valves

U-10 and U-85 to be contaminated, and the activity
was slight,

 

BResults of these examinations were reported by
R. G. Affel on May 13, 1955.
UNCLASSIFIED
1832

RMG

 

Fig. 37. Section of Specimen 5‘5 Showing Fin-to-Tube Wall Junction in ARE Fuel-to=Helium Heat Exchanger.
Fins were wound helically in grom;as and were not brazed to the tubes, Etched. 100X. (W with caption)

UNCLASSIFIED
RMG-1833

 

Fig. 38. Interior Wall of Tube Shown in Fig. 37. As polished. 250X. @y} with coption)

35
UNCLASSIFIED
RMG-1834

  

i
. ! s
‘. - JOM (T A \“'4 Ay ’Nl % , h-‘ 5 ":‘.'
" Cer 5 @ 1 . g 1'. . - » . - 5 \ -
’ l'.. » i "o, — =
L \ o A\
2 0N
i

- i

- —
- - = m—
e

Fig;h?'fl'. Section Shown in Fig. 38 After Etching. 250X. ({mmmly with caption)
*

UNCL ASSIFIED
RMG-1835

 

Fig. 40. Section of Specimen F9 Taoken from Middle of a Bend in an ARE Fuel-to-Helium Heat Exchanger.
As polished, 250X. {"'w'\fth caption)

36
UNCLASSIFIED
RMG

 

Fig. 41. Section Shown in Fig. 40 After Etching. 250X. (99 with caption)

UNCLASSIFIEC
RMG-1837

 
   

 

Fig. 42. Section of Specimen 56 Taken from the Bottom Bellows of Sodium Valve U-23. Etched. 250X. Egergt

with caption)

37
FUEL PUMP PRESSURE TRANSMITTERS’

The main and standby fuel pump pressure
transmitters (PXT-6 and PXT-7) were examined
for calibration shifts, zero shift, radiation damage,
and mechanical damage. The standby pump unit
was inexcellent condition, but the main pump
unit was inoperable because of a failure of the
glass-cloth pressure-sensing diaphragm, as shown
in Fig. 43. Replacement of the diaphragm showed
the unit to be otherwise undamaged. The ruptured
diaphragm was not the source of the initial
fission-gas leak in the ARE, inasmuch as several
checks of the unit at the time the heat exchanger
pits were closed indicated that the diaphragm
was satisfactory. It is believed, however, that
the subsequent rupture of the diaphragm was
the source of the leak that developed after shut-
down of the fuel system. As stated above, the
fission-gas leaks that existed prier to shutdown

 

Results of these examinations were reported by
R. G. Affel on February 18, 1957.

were probably due to one or more of several
potential leaks around the fuel pump.

SCALE FROM MAIN SODIUM PUMP

removed from the
sodium pump bowl showed the following:

An analysis'® of scale

Element Quantity (wt %)
Ni 61.4
Cr 4.88
Fe 2.20
Si 9.44
Mg 0.01
Ca 0.8
MNa Bal

The silicon found in the scale is believed to
be from a residue of the ‘“Conklene’’ compound
used to slush the system before it was filled
with sodium. This cleaning mixture is 99%

Na,0:Si0,-5H,0.

 

mSpecfroscopy Laboratory Report No. 3450.

UNCL ASSIFIED
PHOTO 14490

%
*
5
llS-f':

h
5

 

 

 

weaMg. 43. Main ond Standby Fuel Pump Pressure Transmitter Diaphragms. The ruptured diaphragm was removed

from the main fuel pump pressure tiansmitter.

38
 

ACTIVATION ANALYSES OF SODIUM
SYSTEM SAMPLES

Two sections of pipe containing residual sedium
ond o somple of sodium were examined for gomma
activity after a decay time of 206 days, by using
a sodium iodide gamma-ray spectrometer. The
data obtained are presented in Table 2.

DISPOSITION OF FISSION-PRODUCT
AcTIiViTY!!

Attempts have been made to determine the fate
of several fission product nuclides in the ARE.
The results leave much to be desired since no
adequate plans were made before reactor operation
for the study of this problem. Nevertheless,
several interesting lines of research are suggested
as a consequence of the investigations.

During operation, copious amounts of radioactive
gas evolved from the reactor and escaped into
the pit from a gas leak in the fuel pump. In
an investigation of this gas by P. R. Bell ez al.,'?
the presence of Xe'33, Xe'38, and Kr®® was
revealed by their own or their descendants’ gamma
radiation. It was also observed that poisoning

 

HMuferiul abstrocted from a report by M, T. Robinson,
S. A. Reynolds, and H, W, Wright, The Fate of Ceriain
Fission Products in the ARE, ORNL CF-55-2-36 (Feb.
7. 1955).

uP. R. Bell et al., Measurement of Gamma Radiation
gnm Ojf-Gases of ARE, undocumented memorandum,

ec. 22, 1954,

 

of the reactor was very much below the level
expected if Xe'3® were efficiently retained. '
After shutdown of the ARE and dumping of the
fuel, preliminary measurements indicated that
the radicoctivity of the dump tank was far below
the expected level.'¥ Therefore several samples
taken from the ARE after shutdown were examined
to determine what radicactive nuclides they
contained.

A sample of pipe taken from the intake end
of the emergency off-gas line for draining the
pit was examined in a gamma spectrometer. The
only activity clearly identified was due to Ru'%3,
which is characterized by a 0.50-Mev gamma ray
and o 40-day half life. Chemical tests showed
that this material was probably on the outside
of the pipe.

A similar study wos made of three small samples
cut from the reactor tuel inlet line (line 120).
Gamma spectrometry showed Ru'®3, Ru'%6, and
Zr?5-Nb?5 as the only identifiable activities.
The following disintegration rates were observed

at 62 days after shutdown of the ARE:

Ru™3 (1.3 + 0.3) x 107 dis/min/em? ,
Zr?5.Nb%5 (1.3 + 0.2) x 108 dis/min/em?

 

wJ. L. Meem and W, B. Cottrell, Preliminary Report —
Operation of the Aircraft Reactor Experiment, ORNL
CF-54.11-188 (Nov. 30, 1954),

4y K. Ergen, personal communication,

Table 2. Gamma Activity in ARE Sedium System Samples

 

Approximate Maximum

 

Sample Number Description of Sample Activity Disintegrations
per min per em® of No

1 10-in. section of 2-in, pipe te sodium pump Mns" e 103

containing about 10 emof sodium Zn63 1 x 104

Cob0 4 x 102

2 14-in. section of 'I%-in. pipe to heat ex= Mn¢ 7 x 10°
changer containing about 5 em3 of sodium Zn®3

No?2 (= 1 x 104
Cob0

3 Sodium sample of about 50 em> Na?2 2 x 102

Mn34 1 x 102

Zn5 2 x 102

 

* Activity indistinguishable; may be combination of one or more.

39
The expected ratio of disintegration rates for
unsegregated fission products at this age is

Rul03/Z,95.Nb% = 0.8 ,

whereas a value 10 was found. It is clear that
some ‘‘plating’’ of ruthenium onto the walls of
the fuel line had occurred.

The decay of radicactivity of a sample of solid
ARE fuel taken as liquid from the dump tank was
followed through the period from 31 to 81 days
after ARE shutdown. The total activity of the
sample was observed with the large ion chamber
of the Radioisotopes Department. These results
were combined with gamma spectra to yield both
total photon emission rates and differential decay
data. By observing gamma energy and half life,
the following nuclides were identified:

Nuclide Half Life Gamma Energy (Mev)
Ba'40. 010 13days’S 2.5, 1.60, 0.8, 0.5, 0.33
Ce 141 28 days 0.14
ZeI5.Nb%®  Very long'® 0.8

No Ru'® or 1?1 was detected. It seems likely
that the amount of the ruthenium present was
relatively Because of the long delay
before decay measurements were started, the
detection of 113! was made difficult; however, the
differential decay data were analyzed to estimate
the specific activity of the sample:

small.

Time After Specific Average
ARE Shutdown Activity Gomma Energy
(days) (curies/kg) (Mev)
31 16 0.96
79 3.5 0.73
At 79 days, the dose rate from the sample

(0.0074 g) was measured with a calibrated “‘cutie
pie.'" If the dose rate is assumed to be given
by r/hr at 1 ft = n CE, where C is the number
of curies in the sample and E is its energy, the

 

15This mixture of nuclides was at transient steady

state and decayed with the Ba'40 half life. Most of the
gamma rays were due to Lo 40

16This mixture was still approaching equilibrium; the
opparent half life was too long to meosure with limited
precision of the available equipment. Both iscbars emit
gamma rays at about 0.8 Mev,

40

results give » = 8, Since a value of 6 or 7 is
usually assumed for n, it is apparent that the
measurements of the specific activity of the
fuel are essentially in agreement with the values
obtained on small samples with a *‘cutie pie.”

Another sample of fuel was used for various
chemical and radiochemical analyses. The
chemical results are given in Table 3. The

Table 3. Results of Chemical Analyses of ARE Fuel

 

 

Fuel Before ARE High In
Component Power Operation* Dump Tank
U, wt % 13.59 5.97

Fe, ppm 25 140
Cr, ppm 420 250**
Ni, ppm 25 70**

 

*The U analysis was taken from the ARE Nuclear Log
Book; the other results were obtained from W. R. Grimes,
Jan. 4, 1955.

**Corrected to basis of undiluted fuel by multiplying
by 13.59/5.97.

uranium analysis was lower after ARE operation
because of the use of barren fluorides to flush
out the fuel system. The iron presumably re-
sulted primarily from the drilling operation used
to sample the salt.'” Aliquots of the sample
from the dump tank were analyzed for Sr89, 7,95,
Ru'99,  Cg138, €127, Ce'*, and La™Y; all
except lanthanum and the two cesium isotopes
were determined by conventional radiochemical
methods.

In order to estimate the efficiency of retention
of typical fission products, the radiochemical
analyses reported on ARE fuel were compared
with similar results obtained on a sample of
solid fluoride fuel (NuF-ZrF‘-UF , 8.5 wt % U)
irradiated in hole 12 of the ORNl Graphite Re-
actor. The irradiation time approximately matched
the high-power operating time of the ARE. The
ratios were then obtained between
reported for the ARE fuel and those
the standard sample (designated

following
analyses
reported for
as MR-1):

 

7w, E. Browning and H. L. Hemphill, Solid- State
Semiann. Prog. Rep. Feb. 28, 1955, ORNL-1851, p 18.
Nuclide Ratio (ARE)/(MR=1)
587 6.1

795 3.3

Ry 103 1.5 x 10~4
La'40 14

Ce 4l 18

The ratios were corrected to the ends of the
respective irradiations. Apparently Sr8?, a de-
scendant of 2.6-min Kr®?, was reduced by a
factor of about 2 below the expected level in
the ARE, presumably due to partial escape of
its noble gas ancestor. The low value of Zr?s
has not been explained. It is of interest to
point out that the amount of Zr?5 reported radio-
chemically was in good ogreement with the
amount deduced from the decay data on solid
fuel. It must also be mentioned that the amount
of Zr?5-Nb? found on the walls of the fuel circuit
was apparently negligible compared with the
amount found in the fuel (about 1 port in 2000
if the wall activity was token to be uniform
over the entire surfoce). The very low Ruy'03
value in the ARE sample moay possibly have
been due to a faulty analysis, but it was felt
to be real. The agreement was fair between
radiochemical results from the ARE dump-tank
material for Ce'¥' and La'4? and the results
deduced from decay of the solid sample of ARE
dump-tank meterial. The cerium and lanthanum
results in the experiment and in the decay study
oppear to be in reosonable agreement with what
would be expected from the ARE power history.

Gamma spectrometry was used to determine the
ratio of amounts of Cs'3® and Cs'37 in the two
samples. The former isotope is ‘‘shielded’’;
that is, it must be formed directly in fission

—— e e ——— — = - — — — — " pp— -

since Xe'2% is stable and has a very low thermal-

neutron absorption cross section (0.15 barns).
On the other hand, Cs'37 is the daughter of
3.9-min Xe'®7. Thus, a difference between the
ratios of the amounts of these isotopes in the
two samples was a measure of the escape from
the ARE fuel of Xe'37. The results indicate
that not over 20% of the Xe'?7 escaped from the
fuel, and that, possibly, none did.

It has become customary'®/'% to describe the
release of xenon from the fluoride fuels in terms
of a quantity )l.p, defined by

Rate of Xe escape =.?-lp x amount of Xe in fuel.

From the ARE poisoning data it is roughly es-
timated that for Xe 135

Ap = 5 % 104 sec—! .

This valve is consistent with the observed be-
havior of Cs'37 also. From the Sr87 dote reported
above it appears that krypton isotopes have larger
values of A, than do xenon isotopes.

Several lines of investigation ore suggested by
the results reported here. In particular, the
effects of ruthenium on the physical properties
of Inconel and on corrosion by the fluoride fuels
should be studied. If all Ru'®? is removed from
the fuel by the walls, and if the ruthenium “‘plate’’
is uniform over the entire reactor, the approximate
rate of deposition of Ru'0% is 0.7 (P/A) w/hr,
where P is the total reactor power in Mw and A

is the surfdce area in ecm?.

 

by 1 Meem, The Xenon Problem in the ART, ORNL
CF-54-5-1 (May 3, 1954).

19\, T Robinson, Release on Xenon from Fluoride
Fuels: oposal for an Experimental Program, ORNL
CF-54.6- 4 { une 2, 1954).

 

DISPOSAL OF DISASSEMBLED SYSTEM

The radioactive disassembly debris from the
ARE will eventually be treated for recovery of
unknown amounts of fuel believed to be adhering
to these materials. The reactor and the BeO
blocks have been buried, as described above.
The fuel is awaiting reprocessing in the fused-

salt fluoride-volatility process being developed.

The various samples not yet examined will be
token up as time permits and disposed of as
appropriate at that time. Reports of subsequent
examination will appear in periodic progress
reports.

41
 

 

 
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