ORNL-TM-3151.txt

 
 

 

ORNL-TM=-3151

Contract No. W-ThO5-eng-26

Reactor Division

A STUDY OF FISSION PRODUCTS IN THE MOLTEN-SALT
REACTOR EXPERIMENT BY GAMMA SPECTROMETRY

A. Houtzeel F. F. Dyer

AUGUST 1972

NOTICE This document contains information of a preliminary
nature and was prepared primarily for internal use at the
Cak Ridge National Laboratory. It is subject to revision or
correction and therefore does not represent a final report.

NOTICE

is report was prepared as an account of work
sponsared by the United States Government, Neither
the United States nor the United States Atomic Energy
Commission, nor any of their employees, nor any of
their contractors, subcontractors, or their employees,
makes any warranty, express or implied, or assumes any
legal liability or responsibility for the accuracy, com-
pleteness or usefulness of any information, apparatus,
product or process disclosed, or represents that its use
would not infringe privately owned rights.

 

 

OAK RIDGE NATIONAL LABORATORY
Oak Ridge, Tennessee 37830
operated by
UNION CARBIDE CORPORATION
for the
U.S. ATOMIC ENERGY COMMISSION

S ey 15 URLIME
plsTRIBUTION or TWIS BOCUNENT 158 %fl

   
iii

CONTENTS

ACKNOWLEDGMENT « « v ¢ & o & 4 ¢« o & o o &

SUMMARY

l.

. . - . . . . . - » » - 4 » - .

INTRODUCTION TO THE MOLTEN SALT REACTOR EXPERIMENT.

1.1
1.2

Molten-Salt Reactor Concept . . .

Description of the MSRE . . . . .

OBJECTIVES OF THIS GAMMA~RAY SPECTROMETRY
DESCRIPTION AND PERFORMANCE OF EQUIPMENT.

3.1
3.2
3.3
3.4
3.5
3.6
3.7
3.8
3.9

Background . . . . .+ . ¢ ¢ . . .
General Description . . « . . . .
Detector and Amplifier . . . . .
Analvzer . + + ¢« o o s e s o0 e e
Collimator Assembly . . . . .« . .
Setup of Experimental Equipment .
Locational Equipment . . . . . .
Shielding e h e e e e e e s

Calibration Setup . « « « o« ¢ &+ &

ANALYSIS OF SPECTRA « « ¢« ¢ 4+ ¢« & ¢« o &

4b.l
4.2
4.3
4.4

Purpose and General Procedure . .

Problems . + ¢« ¢+ ¢« ¢ o o « ¢ o« &
Table of Isctopes . . + « « « =

Computer Programs . . + « « « + o

CALIBRATION . o 4 & ¢ o o o o o o o =

5.1
5.2
5.3
5.4
5.5
5.6
5.7

General . . . « ¢ 4« o o 4 e 4 e
Source .+ . ¢ o e 4 s e s e e s s
Slit Experiment — Source Strength
Single Source Experiments . . . .

Heat Exchanger Calibration . . .

Calibration of Shielding Materials

-

Calibration for Fission Gases in the MSRE

MEASUREMENTS ., , . . « « ¢« v ¢ o« o o &

Page

vii

11
13
13
14
17
17
18
20
21
21
24
27
27
28
30
31
33
33
34
35
43
b
48
52
53
. o

. -

RESULTS . « ¢« & v ¢« o « + + .
7.1 Group A Spectra . . . . .
7.1.1 Heat exchanger . .
7.1.2 Main reactor off-gas line
7.1.3 Main fuel lines . . .
7.2 Group B Spectra . . . .
7.2.1 Main reactor off-gas line
7.2.2 Beat exchanger . . . .
7.3 Group D Spectra . . . .
7.4 Group E Spectra . . . e e
7.5 Group F Spectra .
7.5.1 Main reactor off-gas line
7.5.2 Heat exchanger .
7.5.3 Drain tank . . .
7.6 Group G Spectra . . . .
7.6.1 Heat exchanger .
7.6.2 Main reactor off-gas line .
7.6.3 Fuel pump bowl and fuel
7.7 Group 4 Svoectra .
7.7.1 Heat exchanger . .
7.8 Group I Spectra . . .
7.9 Group J Spectra . . . . .
7.10 Group K Spectra .
7.11 Group L Spectra . e e e
7.12 Group M Spectra . . . .
CONCLUSTIONS o . « e e .
8.1 Data Collection and Analysis
8.1.1 Gamma spectrometer system ,
8.1.2 Calibration . . .
8.1.3 Computer analysis program
8.2 Results — General . o .

lines

8.2.1 Metal surfaces in direct contact with

fuel salt .

8.2.2 Main reactor off-gas line .

134
136
136
138
141
141
141
142
143
143

143
145
8.3 Elements and Nuclide Chains . . . .
8.3.1 Niobium . . « « « « « « o« + .

8.3.2 Ruthenium-rhodium . . . . . .

8.3.3 Antimony-tellurium~iodine . .

8.3.4 Extrapolation back to reactor

9, EPILOGUE . o &+ ¢ &+ ¢ o o o s o s o s o o &
Appendix A. TABLE OF RADIONUCLIDES . . . . .

Appendix B. PRODUCTION AND STANDARDIZATION OF

shutdown time .

11OMAs IN SILVER

TUBING FOR MSRE GAMMA SPECTROMETRY . . . . .

Appendix C.

ABSOLUTE EFFICIENCY CURVES OF THE GAMMA-RAY DETECTION

SYSTEM USED FOR THE CALCULATION OF ABSOLUTE AMOUNTS

OF NUCLIDES DEPOSITED . . . . .

Appendix D.
PRINCIPLES . & & v &+ & o & + &

CALCULATIONS OF COUNTING EFFICIENCIES FROM FIRST

Page

146
146
148
151
154
155
161

181

188

194
vii

ACKNOWLEDGMENT

The authors gratefully acknowledge the contributions of the many per-
sons who were instrumental in the proper execution of this experiment, es-—
pecially in the design of the equipment, the recording of the spectra, and
the analysis of the results. In particular, we appreciate the valiant help

of the following persons,

R. Blumberg, for the very helpful efforts in the design of the equip-
ment and the recording of the spectra.

J. R. Engel, for his wise advice during the course of the experiment.

A, F, Joseph, for his effort in making the spectrum analysis program
operable on the ORNL computers.,

J. L. Rutherford, for the help in analyzing the computer results and
preparing the figures.

L. P. Pugh, for the assistance in the design of the equipment,

J. A. Watts, for his help in analyzing the computer results and preparing

the figures.
SUMMARY

The operation of the Molten-Salt Reactor Experiment (MSRE) has demon-
strated that a mixture of fluoride salts is a practical fluid fuel that is
quite stable under reactor conditions. The chemistry of the fission prod-
ucts is such, however, that some of them leave the circulating fuel salt
and appear on the moderator graphite‘in the core, on the metal surfaces ex-
posed to the salt, and on the metal surfaces in the off-gas system. Xenon
and krypton fission gases are stripped in the off-gas system, where they
decay to daughter nuclides. Some other elements (Mo, Nb, Ru, Te, and Sb)
appear to exist in the metallic state and tend to plate out on metal or
graphite surfaces or be carried out into the off-gas system as particles.
Because it is important in the design of larger molten-salt reactor systems
to know where and in what proportion fission products are distributed
throughout the system, considerable efforts were made to obtain this infor-
mation in the MSRE.

A technique was developed at ORNL to locate and measure fission prod-
uct depositions on surfaces exposed to the salt and in the off-gas system
of the MSRE by the intensity and energy spectrum of the emitted gamma rays.
A gamma-ray spectrometer was developed, consisting of a Ge(Li) detector, a
4096-channel analyzer, and a lead collimator to permit examination of small
areas., This device was usually positioned with the MSRE portable mainte-
nance shield over the different reactor system components, with precise
alignment and location achieved by a laser beam and surveyor's transits.

Measurements were made not only with the reactor system shut down
and drained but also with the fuel circulating and the reactor at several
power levels, Altogether some 1000 spectra were taken, 257 of which were
recorded with the reactor at some power level (a few watts to full reactor
power). Another 400 spectra were taken to calibrate the equipment. Com-
puterized data handling permitted this mass of data to be analyzed quali-
tatively and quantitatively.

Most of the effort was focused on the off-gas system and the primary
heat exchanger, the latter because it contains 407 of the metal surfaces
exposed to the salt. The off-gas system contained not only fission prod-

ucts with gaseous precursors but also metallic elements with their decay
products [such as Nb, Mo, Ru, Sb, Te(I)]. The MSRE heat exchanger con-
tained mostly depositions of the same metallic elements; It was observed
that fission gases form a major source of activity in the heat exchanger
when the reactor is shut down and the fuel is drained immediately (emer-
gency drain).

A high~resolution gamma-ray spectrometer used with proper remote main-
tenance equipment and location tools proved to be very vefsatile in locating
and evaluating fission product depositions in a highly radioactive reactor

system.
1. INTRODUCTION TO THE MOLTEN-SALT REACTOR EXPERIMENT

1.1 Molten-Salt Reactor Concept’:?

The molten-salt reactor concept originated in 1947 at Oak Ridge as a
system for jet aircraft propulsion. The idea was to use a molten mixture
of fluoride salts including UF, as a fuel that could be circulated to re-
move the heat from the core. Fluoride salts looked promising because of
their basic physical chemistry: the vapor pressure of the molten salts
would be extremely low and they would not react violently on exposure to
air or water. Radiation damage would be nonexistent in the completely
ionic liquid fluorides. Since a variety of interesting fluorides (NaF,
LiF, BeF., ZrF,, UF,, etc.) were known to be stable in contact with some
common structural metals, a corrosionless system seemed attainable. 1In
addition, high specific heat, good thermal conductivity, and reasonable
viscosity made these liquids good heat transfer media. On the other hand,
the high melting point of potential fuel mixtures (400—500°C), while no
drawback during power operation, would require provisions for preheating
the piping and keeping the salt molten during shutdowns. An intensive ef-
fort on molten salt was undertaken at the Oak Ridge National Laboratory,
and by 1954 a molten-salt reactor was operating at temperatures around 800°C,
It had been recognized that the technology of the molten~-salt system was
well suited for the development of a commercial Th-22°U power reactor.
Thus, in 1956 the Molten-Salt Reactor Program was established at ORNL,

By 1960 a picture of an economically attractive molten-salt reactor
had come into focus. Its core would contain the graphite moderator in di-
rect contact with molten salt flowing through channels and there would be
either one or two salt streams. If one, it would contain both thorium and

uranium, giving a high-performance converter or even a breeder with a small

breeding ratio.

 

'P. N. Haubenreich, '"Molten-Salt Reactor Progress," Nucl. Eng, Int,
14(155), 325-29 (April 1969).

M. W. Rosenthal, P. R. Kasten, and R. B. Briggs, '"Molten-Salt Reac-
tors — History, Status and Potential," Nucl. Appl. Tech. 8(2), 107-17
(February 1970).
The basic technical feasibility of the molten-salt reactors was on a
sound footing — a compatible combination of salt, graphite, and container
metal. A salt mixture based on 'LiF and BeF, looked most attractive from
the standpoint of melting point, viscosity, neutron absorption, and freedom
from mass transfer. A nickel-base alloy, INOR-8, had been developed that
was practically unaffected by the salt at temperatures to 700°C, that was
superior in strength to austenitic stainless steel, and that was susceptible
to conventional fabrication. It was found that salt did not wet or react
significantly with graphite and that, by reducing the graphite pore size,
intrusion of salt into the graphite could be prevented. Although the ma-
terial situation was encouraging and test loops had operated successfully,
a reactor experiment was needed to really prove the technology. Therefore
the objective of the Molten-Salt Reactor Experiment (MSRE) was to demon-
strate that the key features of the proposed breeders could be operated

safely and reliably and maintained without excessive difficulty.

1.2 Description of the MSRE?®

 

The MSRE was to use essentially the same materials as the breeders.
There was no attempt to design it to be a breeder, however, since this
would have entailed added expense and complexity in the form of a large
core or a blanket of fertile material. Some of the important design cri-
teria were:

. core of bare graphite with fuel flowing in channels,
. removable specimens of graphite and metal in the core,

. provision for sampling the salt and adding uranium during operation,

. power 10 MW or less,

1

2

3

4,  fuel temperature around 650°C,

5

6 heat rejected to the air via a secondary salt loop,
7

. fuel pump rather larger than necessary (to minimize scaleup to

the next reactor),

 

3p. N. Haubenreich and J. R. Engel, "Experience with the MSRE," Nucl.
Appl. Tech, 8(2), 118—36 (February 1970).
8. simplicity and conservatism to enhance reliability,

9. zero leakage of salt in operation,

10. enclosure capable of safely containing spill of entire fuel.
The flowsheet that was arrived at is shown as Fig. 1.1.

Details of the MSRE core and reactor vessel are shown in Fig. 1.2.

The 55-in.~diam core was made up of graphite bars, 2 in. square and 64 in.
long, with flow passages machined into the faces of the bars. The graphite
was especially produced to limit pore size to 4 u to keep out the salt, All
metal components in contact with molten salt were made of Hastelloy N (a
commercial version of INOR-8), which had been approved for construction
under ASME codes., The three control rods were flexible, consisting of
hollow cylinders of Gd;03-Al.0; ceramic canned in Inconel and threaded on

a stainless steel hose. Draining the fuel provided positive and complete
shutdown.

The volute of the centrifugal fuel pump was enclosed in a tank (the
pump bowl) which was the high point in the loop. The pump suction was open
to the salt in the bowl, so that the pump bowl and the connected overflow
tank provided the surge space for the loop. A blanket of helium, generally
at 5 psig, was provided over the salt. A tube into the top of the pump bowl
connected to the sampler-enricher, which was a two-chambered, shielded
transfer box; small sample buckets or capsules containing uranium~rich salt
could be lowered from this transfer box into the pool in the pump bowl. A
spray ring in the top of the fuel pump bowl took about 4% of the pump dis-
charge and sprayed it through the gas above the salt to provide contact
between helium and fuel salt so that the gaseous fission products could
escape into the gas. A flow of 4 liters/min (STP) of helium carried, among
others, the fission gases such as xenon and krypton out of the pump bowl,
through a holdup volume, a filter station, and a pressure~control valve to
the charcoal beds. The beds operated on a continuous-flow basis to delay
xenon for about 90 days and krypton for about 7 1/2 days, so only stable
or long-~lived nuclides could get through.

All salt piping and vessels were electrically heated to prepare for
salt filling and to keep the salt molten when there was no nuclear power.

The air-cooled radiator was equipped with doors that dropped to block the
   

 

 

   

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

      

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

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ABSOLUTE 70 °F AR FLOW: 200,000 cfm
FILTERS 1200 GFM, v T ——————r
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" VENTILATION ; VESSEL POWER FREEZE FLANGE (TYF) ——
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STACK | FAN l T 1 ‘ v
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TANK
'
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N/ | :
e romen e e st i o e e i e s

 

 

Si
FLUORIDE BED

Fig. 1.1. Design flowsheet for the MSRE.

   

 
ORNL-LR-DWG 61097RtA

FLEXIBLE CONDUIT TO
CONTROL RCD DRIVES

  
   
 
 
  
   
   
 
  
  
  
  

GRAPHITE SAMPLE ACCESS PORT oy
4
/&

COOLING AIR LINES

 

ACCESS PORT COOLING JACKETS

REACTCR ACCESS PORT

SMALL GRAPHITE SAMPLES
HOLD -DOWN ROD

OUTLET STRAINER

CORE ROD THIMBLES

 

LARGE GRAPHITE SAMPLES -

CORE CENTERING GRID

FLOW DISTRIBUTOR
VOLUTE

GRAPHITE - MODERATOR ¥
STRINGER

FUEL INLET
CORE WALL COOLING ANNULUS
REACTOR CORE CAN

|

REACTOR VESSEL — T

ANTI-SWIRL VANES

MODERATOR

VESSEL DRAIN LINE SUPPORT GRID

Fig. 1.2. Details of the MSRE core and vessel.
air duct and seal the radiator enclosure if the coolant-salt circulation
stopped and there was danger of freezing salt in the tubes. There were no
mechanical valves in the salt piping; instead, flow was blocked by plugs

of salt frozen in flattened sections of certain auxiliary lines. Tempera=-
tures in the freeze valves in the fuel and coolant drain lines were con-
trolled so they would thaw in 10 to 15 min when a drain was requested. The
drain tanks were almost as large as the reactor vessel, but the molten fuel
was safely subcritical because it was undermoderated. Water-cooled bayonet
tubes extended down into thimbles in the drain tanks to remove up to 100 kW
of decay heat if necessary.

The physical arrangement of the equipment is shown in Fig. 1.3. The
reactor and drain tank cells are connected by a large duct, so they form a
single containment vessel. The tops of the two cells consist of two layers
of concrete blocks, with a weld-sealed stainless steel sheet between the
layers; the top layer .is fastened down. The reactor and drain tank cell
were kept at -2 psig during operation. A small bleed of nitrogen into the
cell kept the oxygen content at- 3% to preclude fire if fuel pump lubri-
cating oil should spill on hot surfaces. A water-cooled shield around the
reactor vessel absorbed most of the escaping neutron and gamma-~ray energy.
The 5-in. salt piping in the reactor cell included flanges that would per-
mit removal of the fuel pump or the heat exchanger. The flanges were made
unusually large and were left uninsulated so that salt would freeze be-
tween the faces.

All the components in the reactor and drain tank cells were designed
and laid out so they could be removed by the use of long-handled tools from
above. When maintenance was to be done, the fuel was secured in a drain
tank and the connecting lines frozen. The upper layer of blocks was re-
moved and a hole cut in the membrane over the item to be worked on; after
a steel work shield (the portable maintenance shield) consisting of two
parts was set in place, a lower block was removed. Then the two parts of
this portable maintenance shield were moved together, and the hole, caused
by the removal of the lower block, was covered. Through 5-in.~diam holes
in the portable maintenance shield, one could then work remotely in the re-

actor cell.
  
 

Fig. 1.3.

REMOTE MAINTENANCE
CONTROL ROOM

 

- 1. REACTOR VESSEL
. HEAT EXCHANGER
FUEL PUMP
FREEZE FLANGE
. THERMAL. SHIELD
. COOLANT PUMP

O AHGN

Layout of the MSRE.

 

 

ORNL-DWG 63-1209R

    

 

  

RADIATCR

. COOLANT DRAIN TANK
. FANS -

. FUEL DRAIN TANKS
. FLUSH TANK

. CONTAINMENT VESSEL
. FREEZE VALVE

 

 

 
10

The conventional instrumentation and control systems for the reactor
were augmented by a digital computer that was used to log data and help
analyze the operation. About 280 analog signals from the reactor were
wired to the computer.

Construction of the primary system components for the MSRE started in
1962, and installation of the salt systems was completed in mid-1964. Pre-
nuclear tests, in which first flush salt and then fuel carrier salt con-
taining no uranium were circulated more than 1000 hr, showed that all sys-

2337 was then added to the carrier salt as the

tems worked well. Enriched
UF,-LiF eutectic (61 wt % U) and on June 1, 1965, criticality was achieved.
In May 1966, the full power of 7.3 MW was reached.

The shutdown in March 1968 was the end of nuclear operation with *3°U.
Sufficient 22U had become available, and plans had been made to substitute
it for the *2°U in the MSRE fuel to measure directly some nuclear charac-—
teristics of great importance to the nolten-salt breeder design. After
shakedown of the processing equipment, the flush salt and the fuel salt
were fluorinated to recover the 218 kg of uranium in them. Uranium-233
was then loaded into the stripped fuel carrier salt, and criticality was
attained in October 1968, after the addition of 33 kg of uranium (917 233 ).
Nuclear operation with *°°U fuel continued until December 1969, when after

4 1/2 vears of successful operation, the reactor was shut down for the last

. 4
time.

 

“M. W. Rosenthal et al., "Recent Progress in Molten-Salt Reactor
Development,'" TAEA At. Energ. Rev. 9(3), 60150 (September 1971).
11

2, OBJECTIVES OF THIS GAMMA-RAY SPECTROMETRY STUDY

The operation of the MSRE demonstrated that a mixture of fluoride
salts is stable under reactor conditions and that the majority of the fis-
sion products remain with the circulating fuel salt; however, some fission
products are found on the moderator graphite in the core, on the metal sur-
faces exposed to the salt, and in the reactor off-gas system. TFor example,
some elements (Mo, Nb, Ru, Te, and Sb) appear to exist in the metallic
state and tend to plate out on surfaces in contact with the salt or to be
carried into the off-gas system as particles.

The behavior of certain fission products, especially those that vola-
tilize or deposit, is of interest for several reasons in a molten-salt
system.

1. The reactor chemists, of course, seek to understand the chemistry
of the fission products in the salt.

2. The shielding required in remote maintenance of reactor components is
strongly influenced by the total amount of highly active fission prod-
ucts deposited in those components.

3. The deposited fission products may represent several megawatts of de-
cay heat, creating a cooling problem after reactor shutdown and drain
in a large, high-power molten-salt reactor.

4, Fission products that concentrate in the core by deposition on graphite
would absorb more neutrons and hence reduce the breeding performance
of a molten-salt reactor.

For these reasons a comprehensive program of studies of fission product
behavior in the MSRE was undertaken. The objective of the study described
in this report was to determine the identity and magnitude of radiocactive
fission product deposits in certain MSRE components using the technique of
remote gamma-ray spectrometry. Particular attention was directed to the
reactor off-gas system and the heat exchanger, the latter because it con-
tained approximately 407 of the metal surfaces exposed to the circulating

fuel salt. The deposition on graphite and metal in the core and in the
12

pump bowl is being studied by others and is discussed in other reports., >

This report presents the results of the remote gamma-ray spectrometry in
a readily usable form with some interpretations that may be useful in the
overall effort of understanding the behavior of fission products in the

MSRE.

 

®F. F. Blankenship et al., MSR Program Semiannu. Progr. Rep. Aug. 31,
1969, ORNL-4449, pp. 104—9.

°C. H. Gabbard, MSE Program Semiannu. Progr. Rep. Feb. 28, 1970,
ORNL-4548, p. 13.

’F. F. Blankenship et al., ibid., pp. 104-8.

°F. F. Blankenship et al., MSR Program Semiannu. Progr. Rep. Aug. 31,
1970, ORNL-4622, pp. 60—70.
13

3. DESCRIPTION AND PERFORMANCE OF EQUIPMENT

3.1 Background

The equipment that was used to obtain the data in this report (de-
scribed below) was developed over a two-year period.

In 1967 Blumberg, Mauney, and Scott? began to study devices for lo-
cating and evaluating amounts of radicactive materials in high-radiation-
background areas. During a shutdown of the MSRE in May 1967, they mapped
the intensity of radiation coming from the fuel heat exchanger using a
gamma-ray dosimeter mounted over a collimator in the portable maintenance
shield. During the same shutdown a few data on energy spectra were obtained
with a sodium iodide crystal mounted in a lead shield with a collimating

235

hole. At the end of U operation in March 1968, they made more, better

measurements of gamma spectra by using a different collimator-shield com-

bination, a lithium-drifted germanium diode, and a 400-channel analyzer.'®

The conclusion then was that remote determination of fission product depo-

gition by gamma spectrometry of a collimated beam was feasible and would

provide useful information, but that some improvements should be made in

the equipment. Accordingly, modifications were made with the following

specific objectives:*?

1. ability to position and aim the apparatus at a selected source with
great accuracy,

2. detector resolution good enough to identify individual nuclides among
a multitude,

3. simplified data handling and analysis,

4. collimation adaptable to a wide range of source strengths,

provisions for measuring spectra from selected spots during and im-

mediately after power operation,

6. better calibration.

 

°R. Blumberg, T. H. Mauney, and D. Scott, MSR Program Semiannu. Proger.
Rep. Aug. 81, 18967, ORNL-4191, pp. 4044,

1°%. Blumberg, F. F. Dyer, and T. H. Mauney, MSR Program Semiannu.
Progr. Rep. Aug. 31, 1968, ORNL-4344, pp. 36—40, 196.

'R, Blumberg, F. F. Dyer, and A. Houtzeel, MSR Program Semiannu.
Progr. Rep. Aug. 31, 1969, ORNL-4449, p. 31.
14

By June 1969 these objectives had largely been met in the equipment de=-

scribed below.

3.2 General Description

 

As shown in Fig. 1.3, the fuel circulating system and drain tanks are
situated in underground cells which, during operation, were covered by two
layers of concrete beams with a thin stainless steel sheet between the
layers. Gamma radiation levels in the reactor cell were 40,000 to 70,000
R/hr when the reactor was at full power, dropped to 3000 to 5000 R/hr upon
a shutdown and drain, then slowly decreased.'?® Gamma radiation in the
drain tank cell ran as high as 25,000 R/hr immediately after a drain.'?
Thus, the situation dictated that any gamma spectrometry measurements would
have to be made from a distance of 10 to 20 ft through apertures in a bio-
logical shield.

Even at the top of the shield, the intensity of the gamma-ray beam
through an opening was quite high. For example, the beam above a 5-in.-
diam hole in the portable maintenance shield, about 14 ft above the primary
heat exchanger, was on the order of 500 R/hr one or two days after a shut-
down and drain. Thus the radiation to the detector had to be reduced by
collimation and sometimes by attenuation through shielding plates as well.
Of course, the collimation of the beam was necessary also to restrict and
locate the source of the gamma rays being analyzed.

Figure 3.1 is a schematic, general view of the ultimate equipment,
consisting of a collimator, a detector, and a laser alignment device.
Figure 3.2 is a front view of the equipment. In these illustrations the
equipment is mounted on the portable maintenance shield, but it could also
be mounted over small holes drilled through the concrete shield blocks
especially for this purpose. The detector was a Ge(Li) crystal connected
through appropriate amplifiers to a 4096-channel analyzer. This combi-
nation provided the high-resolution capability that was necessary. Dif-

ferent collimator inserts could be used, depending on the intensity of the

 

12\, Houtzeel, MSR Program Semiannu. Progr. Rep. Aug. 31, 1968,
ORNL-4344, pp. 22—23.
15

ORNL - OWG 68-139B0R2

LASER

LASER ALIGNMENT BEAM

GeHJ)DETECTOR_‘\\Ti

COLLIMATOR INSERT
COLLIMATOR BODY

 

 

/,/ o L PORTABLE MAINTENANCE

s s s

i g SHIELD RAIL

 

 

 

 

 

 

 

 

 

 

48in.+

 

{2 BLOCKS
REMOVED)

 

 

 

 

 

 

 

 

 

NOTE:
N ——— ‘ PORTABLE SHIELD ROLLS EAST
AND WEST; CENTRAL PORTION
ROTATES FOR NORTH~SCOUTH
MOVEMENT

VIEWING CONE —==

HEAT EXCHANGER

 

 

 

 

LINE 402

Fig. 3.1. Schematic arrangement of the remote gamma-ray spectrometer
on top of the portable maintenance shield.
 

 

 

Fig. 3.2. Front view of detector,
portable maintenance shield.

PHOTO 979004

LEAD GLASS
WINDOW

collimator assembly, and laser on

[ ]

&
-

 
17

gamma radiation. When the equipment was used with the portable mainte-
nance shield (PMS), the location from which the detected radiation was
coming could be seen by shining the low-energy laser beam down through the
collimator and looking through a lead-glass shielding window in the PMS.
The precise location of the beam could be determined with a pair of sur-

veyor's transits set up on the floor beside the reactor cell.

3.3 Detector and Amplifier

 

The detector was a coaxial, lithium-drifted germanium detector
mounted on a specially extended arm at a right angle to the cryostat. The
manufacturer, ORTEC, measured a total resolution of 1.78 keV FWHM at 1333
keV at the time of delivery. The detector was 36.65 mm in diameter by
28.5 mm long with a total active volume of 26,25 ecm®. The peak-to-Compton
ratio was 27/1. The measured efficiency (the ratio of the area under the
1333-keV photopeak recorded with this detector to the comparable area re-
corded with a 3- by 3-in., Nal crystal, with 25 cm between the source to the
detector) was 4.3%. Most of the experiments were done with an ORTEC-440A
or ORTEC-450 amplifier, both of which were adequate and posed no problems.

Although the collimated beam to the detector was very intense, it ac-
tually interacted with only a very small portion of the total germanium
crystal. Although some 1400 spectra were taken in a period of six months
under far from ideal conditions, it did not appear that the system reso-

lution deteriorated appreciably during the rugged handling.

3.4 Analyzer

A Nuclear Data 2200 series, 4096-channel, analyzer was used. All
spectra were recorded on magnetic tape. Initially several problems were
encountered in operating the analyzer and tape unit, partly due to 'bugs"
in this new system. Other problems appeared to be connected with the fact
that the analyzer was initially located in a quite hot and very humid area
and had to be shut down frequently for experimental reasons.

Once the analyzer and amplifier were placed in an air-conditioned,
humidity~controlled room and left on power continuously, most of the prob-

lems disappeared. This move necessitated a 200-ft cable between the
18

detector preamplifier in the reactor high-bay area and the amplifier lo-
cated near the analyzer; the long cable did not noticeably influence the
system performance. System gain shifts, sometimes of several channels,
seemed to a certain degree dependent on local power conditions (voltage
and possible frequency variations). We accepted this fate and had no op-
portunity to investigate this dependence. This high-resolution detector
and 4096-channel analyzer combination performed satisfactorily. However,
when the reactor was at power the cacophony of gamma rays from short-lived
fission products in components such as the primary heat exchanger or reac-
tor off-gas line generally proved too much even for this system; peaks de-
generated into broad conglomerates from several gamma rays. In those cases,
either a fair amount of shielding had to be used between the detector and
the component observed to cut down the radiation level, or such a spectrum

could not be analyzed.

3.5 Collimator Assembly

Figure 3.3 shows the collimator assembly, which consisted of a col-
limator body and a collimator insert. The collimator body, a lead-filled
cylinder 32 1/2 in., high and 19 in. in diameter, had a central hole into
which a collimator insert, also a lead-filled cylinder, could be placed,
Three different, interchangeable collimator inserts having beam holes 1/16,
1/8, and 3/16 in. in diameter, respectively, were available to cope with
the different radiation intensities. Ultimately only the inserts with
1/16~ and 1/8-in. beam holes were used to collect data, The inserts were
12 in. long and 3 7/8 in. in diameter. The straightness of the beam holes
in the inserts proved to be a manufacturing problem since drilling of such
small holes over a 12-in. length is almost impossible. Reasonably straight
holes were obtained by placing thick-walled stainless steel precision tubes
of 1/16, 1/8, and 3/16 in. inside diameter in the collimator inserts and
then filling the inserts with lead.

The detector and Dewar bottle were mounted on a platform attached to
the collimator body. They could be moved on this platform with an adjusting
screw to ensure that the detector was always at the same location over the

collimator beam hole. The laser was fixed to the collimator insert with a
19

ORNL-DWG 69-8050

J

 

 

 

 

 

 

 

 

LASER——
.
i
LASER ALIGNMENT BEAM — i
Ge(Li) DETECTOR \H\
COLLIMATOR INSERT # 9
7
COLLIMATOR BODY é
?
/

 

 

 

 

 

-1

 

 

 

 

 

Fig. 3.3, Schematic of collimator assembly, detector, and laser.
20

small jig in such a way that the laser beam could be adjusted to shine
through the middle of the collimator beam hole. For the laser to shine
through the collimator hole, it was necessary to get the detector out of

the way by moving the detector and Dewar backward on the platform with the
adjusting screw, It would have been easy to guide the laser beam around

the detector into the collimator beam hole with some sort of a mirror sys-
tem, but it was thought that the anticipated rough handling of the detector,
laser, and collimator assembly in moving it from place to place would not
permit this extra complication. The collimator insert and laser (in fact,

one subassembly) could easily be taken out and replaced or adjusted.

3.6 Setup of Experimental Equipment

Several days were required after reactor shutdown to remove the upper
shield blocks and the containment membrane, set up the PMS and the spec-
trometry equipment, and start a survey. For this reason, it was decided
to drill holes through the upper and lower shield plugs at three different
locations in the shielding so that the spectrometer could be used while the
reactor shield plugs were still in place. The three locations chosen were:
one over the reactor system off-gas line, one over the primary heat ex-
changer, and one over a drain tank. The holes were drilled during the
June—August 1969 reactor shutdown period., The alignment of the holes in
the upper and lower shield plugs was done with surveyor's transits. Of
course, the containment membrane between the layers of shield blocks in
both the reactor cell and drain tank cells remained intact. This setup,
with the collimator assembly and detector placed on top of the upper shield
blocks, was very useful for taking gamma-ray spectra during and immediately
after reactor shutdown; it also gave us an opportunity to study the shorter-
lived isotopes, at least at these three locationms.

We used the PMS to survey the fission product activities along the

13

heat exchanger axis, the fuel lines, and the off-gas line, By placing

the collimator assembly and detector over one of the holes in the rotating

 

'?R. Blumberg and E. C, Hise, MSRE Design and Operations Report.
Part X. Maintenance Equipment and Procedure, ORNL-TM=-910.
21

work plug of the shield, we could survey 35 in. of the heat exchanger with
one setup. Figure 3.4 shows the general arrangement. By placing the PMS
over any group of two adjacent lower shield plugs, we could survey large
areas in the reactor cell, The collimator assembly and detector were placed
on a large axial bearing so that the assembly could be turned freely on the

PMS. This proved to be very useful for locational purposes.

3.7 Locational Equipment

It was important to know precisely the locations on the different com-
ponents from which gamma-ray spectra were taken. A low-power laser and two
surveyor's transits were used for this purpose. The laser was mounted in
the previously mentioned adjustable jig which could be attached to any of
the collimator inserts. The laser beam coming through the center of the
collimator hole and falling on the reactor component surveyed would then
accurately indicate the center of the spot from which a gamma-ray spectrum
was taken. The red laser '"dot" could be observed very easily through a
lead-glass shielding window in the PMS (even on a 1/2-in. line some 15 ft
away). Through previous calibration with a plumb line and an adjustable
bubble level on the collimator body, we could ensure that the laser beam
was perfectly vertical. With the surveyor's transits, placed in the re-
actor high-bay area, it was then possible to locate the vertical laser beam
and hence the spot examined on the reactor component., This rather simple

system proved to be accurate within a fraction of 1 in.

3.8 Shielding

As already mentioned, the radiation intensity from several components
was rather high. With the reactor at power, the radiation levels from the
holes in the concrete shield plugs over the reactor off-gas line and heat
exchanger were more than 1000 R/hr. An important dose due to fast and
thermal neutrons was also detected. (Except when spectrometry data were
being taken, the holes were plugged and covered with lead bricks.)

With the reactor at full power, it proved to be impossible to take

meaningful spectra through the hole over the heat exchanger. So many
 

   
   

PHOTO 97902A

oy,

RS B

 

 

 

 

 

 

 

 

 

B

CRY

 

Fig. 3.4. General arrangement of detector-collimator assembly and
portable maintenance shield placed over removed lower shield plugs.

) _)

Zc

 
23

fission product nuclides appeared in the circulating fuel salt that among
the great multitude of photopeaks present in these spectra, many coincided
to form meaningless, large conglomerate peaks., These spectra could not be
analyzed by our computer program for spectrum analysis. The off-gas con-
tained fewer different nuclides than did the fuel salt, but even so, spectra
taken with the reactor at full power through the hole over the off-gas line
were on the borderline. A liberal use of shielding material, such as 2 in,
of lithium-impregnated paraffin and 1 in. of copper, was necessary to render
those off-gas-line spectra analyzable. (Lithium-impregnated paraffin was
used since it absorbs neutrons without emitting capture gamma rays.)

With the reactor at low power or right after shutdown from high power,
we were able to analyze spectra from both the heat exchanger and the off-
gas line. Still a heavy amount of shielding was necessary to keep the
count rate from the detector within reasonable limits, 3000 to 10,000
counts/sec.

A problem encountered was that many good shielding materials, such as
lead, attenuate low-energy gamma rays (energies of a few hundred keV or
lower) too strongly when enough shielding is used to obtain the desired
degree of attenuation for the higher energy photons. Thus low-energy peaks
could be entirely wiped out under such conditions. Because materials of
low atomic number do not show such a strong shielding effect at low ener-
gies, this problem was largely overcome by using shields made of aluminum
and/or copper. Thus the count rate could usually be kept at a desired
level while allowing measurement of the lower energy photons. Chapter 5,
concerning calibration of the equipment, will show in more detail the prob-
lems encountered.

In general, we tried to operate the system at a dead time of 25% or
less using a minimum of shielding material and the 1/16-in., collimator

insert,
24

3.9 Calibration Setup

 

The detector system efficiency was determined empirically, since the
geometry, especially of the primary heat exchanger, was thought to be too
complex to ensure adequate precision from theoretical calculations. The
complete calibration procedure is given in Chapter 5. A subsequent check
of this calibration using simplified calculational models is described in
Appendix D.

Figure 3.5 shows the physical setup for the calibration experiments.
The detector, with a collimator, was installed in one of the underground
cells (the decontamination cell) in the reactor high-bay area. A mockup
of a complete section of the primary heat exchanger with a radiation source
tube was located in the adjiacent remote-maintenance-practice cell. Radi-
ation from the source tube positioned in the heat exchanger mockup passed
through a specially drilled hole in the wall between the two cells and the
collimator and then interacted with the detector. The wall between the
cells was thick enough to eliminate any radiation at the detector except
that coming through the collimator. The remote-maintenance cell was closed
off with the regular roof plugs, except for the area directly above the
heat exchanger mockup; the PMS was used here. Through holes in the PMS,
we could then manipulate the source and the dummy heat exchanger tubes in
the mockup with a simple tool.

A 24-Ci ''°"Ag source was used for the calibration experiment., This
source, activated by irradiation in the Oak Ridge Research Reactor, was
6.3 in. long and 1/2 in. in diameter. (It should be remembered that the
real heat exchanger tubes also have a diameter of 1/2 in.)

By placing this source tube in each of the heat exchanger tube po-
sitions, with dummy tubes in all the other heat exchanger tube positions,
we measured the influence of radiation from every one of these tubes on
the detector. The addition of the readings from all the heat exchanger
tube positions gave the detector response that would have been produced by
a heat exchanger completely loaded with activated '*°"Ag tubes. The dis-

tance between the heat exchanger mockup and the detector was made the same
25

ORNL-DWG 70-9563

 

 

 

 

HEAT EXCHANGER
MOCKUP

 

 

CRYOSTAT

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

N\

LEAD SOURCE
SHIELD

Fig. 3.5.

 

. . » ‘e
. o

‘e

LI - .o

' .
"o

' . ..

' . [

4

[ Con

e L .

SOt Ty

S

Plan view of calibration arrangement.

 

 

 

 

 

\LASER

DETECTOR

COLLIMATOR

 

 
26

as the average distance between the detector placed on the PMS and the
actual primary heat exchanger in the reactor cell.

The heat exchanger calibration experiments were done with a dummy
section of the heat exchanger outer shell (1/2 in. llastelloy N) placed be-
tween the tube mockup and the detector. The alignment of the collimator
hole between the source and the detector was again done with the laser rig,
which was mounted on the now horizontally placed collimator insert. The
red laser beam, passing through the collimator insert, was easy to locate
and was aimed at the center of the heat exchanger mockup.

The calibration required for the measurements on the reactor main off-
gas line was less complicated because the portion of the line that was ob-
served was simply a corrugated l-in. tube. It was estimated that the effect
of this corrugated tube in comparison with the source tube would be approxi-~
mately the same on the detector.

The ceramic part of a heat exchanger heater, as well as the shielding
plates we used during the actual survey of the heat exchanger and off-gas
line in the reactor cell, were calibrated with the same source to compare
their shielding capabilities with those calculated from literature data.
For this purpose, the ceramic heater was placed close to the heat ex-
changer mockup with the source tube in an unshielded center position. The
shield plates were placed as close as possible to the wall in the remote-
maintenance practice cell in front of the hole. The shielding capabilities
of the different plates as found by these tests agreed well with the values

calculated from literature data.
27

4, ANALYSIS OF SPECTRA
4,1 Purpose and General Procedure

The purpose of analyzing a gamma-ray spectrum is to identify radio-
active nuclides by their emitted photons and, taking into account the ab-
solute counting efficiency of the system, to determine the amounts of the
various nuclides present at a given location in a particular reactor com-
ponent.

The counts in each one of the 4096 channels of the analyzer system are
stored in the analyzer memory and, after the end of the counting time,
written on magnetic tape if so requested. Hence, the basic data are counts
assembled in channels. Each channel represents a particular energy interval
whose size can be set more or less at will. We generally used an energy
interval (so-called "gain') in the range of 0.3 to 1.0 keV per channel., 1In
other words, we could take a gamma spectrum comprising a range of 1200
(v0.3 x 4096) to 4000 keV (1.0 x 4096). A gamma-ray peak (usually called
a photopeak or full-energy peak) would identify itself in such a spectrum
as an increase in accumulated counts in several adjacent channels. Experi-
ments have shown that such a photopeak has essentially the form of a modi-
fied gaussian curve. The intensity of a photopeak is determined from the
height of, or more precisely from the area under, such a gaussian curve,
The energy of the gamma ray is determined from the location of the gaussian
curve in the spectrum, that is, the channel in which the top of the gaus-
sian curve falls,

The analysis of a recorded gamma-ray spectrum containing several photo-
peaks proceeds through the following basic steps:

1. Locate the peaks in the spectrum,
2. Determine for each peak if it is just a statistical fluctuation or

a real photopeak and determine a beginning and end of each peak.
3. Determine the background from adjacent channels for each peak.

4. Fit a gaussian curve to the net counts in those channels that form

a peak.
28

5. Calculate the area under the peak as a measure of the gamma-ray
intensity, subtract the background (above) from this area, and de-
termine the location of the top of the gaussian curve to assign an
energy to that peak.

6. Examine the spectrum peak by peak to determine the nuclides responsi-
ble for the observed spectrum; a search is made through an energy 1li-
brary to find those isotopes that might emit gamma rays within the
range of each energy peak.

7. Select the nuclide most likely to be the one responsible for the peak
in question. Each peak is subjected to several tests before it can
really be associated with a certain isotope. These tests include
half-life, presence of associated gamma rays and their relative in-
tensities, and presence of precursor and daughter nuclides.

8. Once a peak is thus identified, calculate from the area under the peak
how much of the particular isotope is present,

A requirement for such a spectrum evaluation is that the efficiency of
the detector system as a function of photon energy, that is, the ratio be-
tween the counts detected and the gamma rays emitted, must be well known
for the energy range considered. Also, the relationship between gamma
energy and channel number must be determined to evaluate the energy of the

peaks accurately. A typical spectrum is given in Fig. 4.1,

4,2 Problems

Because most of our spectra were taken relatively soon after reactor
shutdown, many short-lived isotopes were still present. This resulted in
a large array of photopeaks, many of which overlapped each other. This
means that several single photopeaks (singlets) would conglomerate (multi-
plets); the analysis of these multiplets poses an entirely new dimension
to the analysis of a spectrum. For example, the beginning and end of any
one peak in a multiplet is very difficult to determine, as is the subtrac-
tion of the background from this multiplet. Basically, a few gaussian

curves have to be shifted back and forth so as to fit a multiplet.
ORNL-DWG 72-7444

29

 

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30

The identification of isotopes by their emitted gamma rays is rela-
tively easy if one has an extensive, accurate library of gamma-ray energies
and branching ratios. The gamma rays of longer-lived isotopes are well
known, but the shorter-lived ones are not and their descriptive data are
scarce,

As implied above, it is a difficult task to analyze one spectrum com-
pletely, especially if it has to be done manually. If, as in our case,
approximately 1400 spectra have to be analyzed, it becomes obvious that a
computerized analysis is necessary. However, a computer program to handle
our rather complicated spectra was not readily available, and it would have

been too time consuming to have one written.

4.3 Table of Isotopes

 

Considerable effort was expended in setting up a library of isotopic
data. This library is probably fairly unique in that it contains both
short- and long-lived fission product data and is taken from the latest
references collected by the ORNL Nuclear Data Project. Gamma-ray energies
and branching ratios are listed per nuclide as well as in order of in-
creasing energy. Absolute gamma-ray abundances, as well as precursors and
decay products, are also given. Use of this table proved to be a major

improvement in the analysis of our spectra. It should be emphasized that

 

the entire analysis of our data is based on this table of isotopes. The
basic isotope table is given in Appendix A.
As an example of possible errors, we might mention the problems en-

countered with the analysis of the 129M7e data. First, this nuclide decays

129 129

partly to the ground state (" °“"Te) and then to I or directly from the

12971, In both cases, the abundance of emitted gamma rays

isomeric state to

is very small.
This means any statistical error in the original photopeak analysis is

amplified by a large factor for the activity calculation. Even more dis-

turbing is the fact that the whole ***"Te-'2?

Te decay scheme is not com-
pletely agreed upon; a literature survey yielded several different decay

schemes. Because the second largest photopeak (696.0 keV) observed in the
31

decay of '?®"Te coincides with photopeaks of other nuclides, we had to rely
only on the results yielded by the 459.6-keV peak.
Based on the latest available data, we estimated that the decay of

129Te and subsequently to *?®I would produce a

129MTe to the ground state
gamma ray at 459.6 keV with an abundance of 4.97%. Should better data be-

come available, the results presented later on should be corrected propor-
tionately. The abundance of the 459.6~keV photopeak for the decay of *?°Te

129

itself to I is estimated to be 7.7%.

4.4 Computer Programs

 

As already mentioned, computerized data handling was imperative for
the proper analysis of this experiment. The principal problem was where
to find a computer program that would analyze our spectra with their multi-
tudes of singlet and multiplet peaks. As an indication of the complexity,
it should be emphasized that we commonly had 150 to 230 peaks per spectrum.

The ORNL Mathematics Division had developed an effective gamma-ray
spectrum analysis program, GAMSPEC-3, which was operational but did not yet
have provisions for analyzing multiple peaks. The program was also limited
to the analysis of 75 single peaks per spectrum and would not identify nu-
clides., These limitations eliminated this program for the analysis of most
of our spectra, especially those still containing many photopeaks from
short-lived isotopes. Gunnink et al,,'* among others, had developed a
versatile gamma-ray spectrum analysis program, GAMANAL, which seemed to
suit our purposes. This program had been written in a Lawrence Radiation
Laboratory (LRL) version of FORTRAN for a CDC 6600 computer and had been
converted to FORTRAN-IV and made compatible with the IBM 360 series of com-
puters by N. D. Dudey et al., of Argonne National Laboratory. Through a
concerted effort of the ORNL Mathematics Division, the program was made

compatible with the ORNL computers. As an indication of the complexity of

 

'“R. Gunnink et al., Identification and Determination of Gamma
Emitters by Computer Analysis of Ge(Li) Spectra, LRL-UCID-15140,
32

the program, it might be mentioned that it took LRL close to five man-years
to write this program, ANL close to two years to make it compatible with
the IBM 360 computers, and ORNL some four months to get it running here.

This program will locate peaks (singlets, doublets, and triplets) and
identify and calculate the amounts of nuclides responsible for the peaks.
It can handle 400 singlet peaks, 100 doublets, and 100 triplets. Although
the basic program was originated by Gunnink et al., we made some modifi-
cations to it and supplied our own library of isotope data. For example,
the original program read in the efficiency of the whole detector system
as a function of energy as coefficients of a polynomial equation. This was
modified to have the program interpolate efficiency values from a set of
data that represented the efficiency curve itself. Normally we supplied
approximately 60 points on this efficiency curve. It was also changed to
permit a larger choice in output of different tables in order to reduce the
paper output,

Although this program worked satisfactorily for most of our spectra,
some problems remained, as follows.

1. In some specific energy ranges, many peaks appeared in multiplets,
Since the program would only analyze up to triplets, it was obvious that a
conglomerate of five or six peaks could not be properly analyzed.

2, The storage possibility of 100 doublets and 100 triplets some-
times was inadequate for spectra that contained many short-lived isotopes.
Such a spectrum would then be rejected.

3. It was only exceptional that all the nuclides of a particular de-
cay chain were deposited. When the program then went through the different
identification tests, it would sometimes fail to find the precursor or
daughter products and hence reject an identification. The only way out was
then to uncouple the parent-daughter relation in the energy library. For
example, in chain 95, ®°Zr remained with the fuel salt while ®3Nb tended to
plate out. The program would not identify ?°Nb as long as ®°Zr was coupled
to it as its precursor. Of course, this had nothing to do with the per-

formance of the program itself.
33

5. CALIBRATION
5.1 General

The equipment was calibrated to determine the absolute efficiency of
the whole gamma-ray detection system, taking into account the collimator
assembly and the physical arrangement of the equipment in the reactor cell.

The counting efficiency as used in our empirical method is merely a
proportionality constant relating photopeak count rate and source inten-

sity, viz.,

CR = EF +« S , (1)

where (R is the count rate, FF is the efficiency, and S is the source in-
tensity expressed as photons per unit of source dimension and per unit
time. It should be notedvthat even though CR is determined by photons from
the total radiocactive object subtended by the collimator, the source
strength can be expressed as any convenient fraction or multiple of the
total object. For example, in the case of the heat exchanger (see Sect.
5.5), count rates that corresponded to the collimator subtending a conical
section of the heat exchanger full of radioactive tubes were obtained. The
source strength, however, was taken as those photons emitted per unit time
per square centimeter of a single tube. Thus the counting efficiency is
expressed as

e CR(counts/min)
9 @)
Y/cm 5( Y )

cm? — min

FF(

 

where y denotes an emitted photon and ¢ a count registered in the photopeak.
When a count rate measured over the heat exchanger was divided by this ef-
ficiency, the result was the photon emission rate per square centimeter of
tubing in the heat exchanger. 1In the case of the reactor main off-gas line,
the source strength was expressed in units of photons emitted per inch of

tubing per minute (y/in.-min), causing the efficiency to have the units
34

c
v/in.
energy of the incoming gamma rays. Basically, there are two ways to de-

 

The efficiency of any system of this kind is dependent on the

termine the relationship between efficiency and energy. One could estab-
lish this efficiency curve by measuring first the detector efficiency in

a simple geometry with several different known sources and then discount
the effects of collimation, distance, shielding, and geometry for an actual
component to be surveyed. The other way would be to establish empirically
the efficiency of the system in the actual physical setup. We chose the
second path because it appeared to be somewhat simpler and more reliable.
However, we also performed our analysis using the more basic approach to
check the validity of the calibration (see Appendix D).

All gamma spectra taken from July to September 1969 were obtained with
the ORTEC-440A amplifier. During our calibration experiments, we changed to
the ORTEC-450 amplifier. Because it was not known at that particular time
what the influence would be on the resolution of the system, it was decided
to do a fair amount of the calibration work in duplicate using both ampli-

fiers.

5.2 Source

In order to simulate the actual conditions as much as possible, we
chose a 1/2-in.-0D silver source tube that was equivalent in diameter to
the primary heat exchanger tubes.

Silver was chosen because '*°"Ag has several gamma rays between 446.8
and 1562.2 keV and thus yields data over a sizable part of the energy range
for the efficiency curve. Furthermore, it was rather inexpensive to prepare
such a source by using the hydraulic-tube irradiation facilities of the ORR.
Because of the large effect of the silver tubing on the reactivity of the
ORR, only 2.3 in. of tubing could be irradiated in a single hydraulic tube;
so three pieces of silver tubing of this length were irradiated, each in
separate hydraulic tubes. Flux monitors (Co-Al foils) were first irradiated
to determine the flux gradients and total flux in the hydraulic tubes and
provide a basis for the silver irradiations. The assembly of our source

tube was done in the hot cells of the ORNL Isotopes Division and consisted
35

in putting the three silver tubes together in one thin-walled stainless-
steel container and seal welding it (Fig. 5.1). Appendix B reports in more
detail the calculations done concerning the source strength as well as the

activity gradient along the length of the source.

5.3 Slit Experiment — Source Strength

Because of the neutron flux distribution in the ORR, it was inevitable
that there was some activity gradient along the source tube., It was neces-
sary to know this gradient to properly calibrate the detector system.
Originally we planned to deduce the level of '*°"Ag radiocactivity at any
point on the tube from the measured activity values and the activation
gradients obtained from the flux monitors (Table B.2 and B.4 of App. B).
However, this procedure has many uncertainties. A much better procedure,
and one that was followed, was to scan the source with a collimator ar-
rangement using the Ge(Li) detector and 4096-channel analyzer. A slit
shield, set up in the remote-maintenance practice cell at the MSRE, con-
sisted of stacked lead bricks to a height of about 2 ft. An air-gap slit
1/8 by 1 in. wide was formed in the brick stack by using 1/8-in. lead
spacers; a glass tube large enough to accept the source was attached verti-
cally to the brick stack over the air gap. The Ge(Li) detector was placed
in the adjoining decontamination cell in front of the hole drilled through
the concrete wall. When the source was placed in the glass tube, gamma
rays from a 1/8-in. section of the silver tube could pass through the slit,
the hole between the cells, and the collimator to strike the detector.
Scanning was begun by lowering the source below the slit and then gradually
raising it until the detector signaled that the top of the source was in
front of the slit.

The scan was made by moving the source upward and obtaining gamma
spectra at 1/8-in. increments until the entire source had been moved past
the slit, A profile of the source intensity was obtained by resolving the
gamma spectra and plotting net counts of photopeak areas versus distance
along the source (see Fig. 5.2). From these results we could then calculate
the absolute activity at any point along the entire length of the source

tube (App. B).
36

ORNL-DOWG 70-9564R

 

 

T

 

 

 

 

  

 

 

 

 

 

 

 

 

 

 

 

. §.3 jp, - —w{ =Y in
g 72 J 7 L L L ol a Z £ L 7/11/{}//4////// [[J[I/J]_er[
VN . N . N N NN N S S N N S N8N NS N NN
0.5in. )
{[_,217/—7_LL11[LL ////LL/
y /
PLUG SOURCE TUBE PLUG BAIL

END CAP

Fig. 5.1. Source capsule assemblyv.
_ . ORNL-DWG 70- 9565
35,000 :

I |
i

 

MEASURED ACTIVITY 4,196 Ci/in.

,
- | |

LA e b s,
30,000 = | 1 MEASURED ACTIVITY 3.861 C|/|n

~T
25, 0Q0 : ! .

 

4!_

     
 

 

 

 

 

 

 

 

 

 

 

 

 
   

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

20,000 l ! ! e b e
£ ‘ .
.
£ .‘
' |
2 ‘ o | ’ | 1.
g 15000 r——— +—4 t— +—-—7;%_:_?4 ,
‘ | Hom i | ‘
A 657.7 k A
10, 000 ey b . &6 ev g o 1
o 0 884.7 kev ''0™aq J | | | | {
{ . ; | i
| ! © 937.5 kev '¥™Maq | 1 | i
5000 - f - V( | | — ‘ , ( f
| 1 | j ! - & : wi
0 — ) : ‘ | . i l . ; | | 2

 

TOTAL SOURCE LENGTH: 6.3 in.

 

Fig. 5.2. Net counts of 3 major '!o™

Ag photopeaks versus distance
along the source tube.

LE
38

Because the activity profile of the silver source tube was not com-
pletely uniform and the collimator subtended a significant portion of the
tube, an average value of the source strength as viewed by the system was

needed. The average source strength, S can be expressed as

 

L
f S(x) RE(x) dx
5§ = 2 , (3)
L
[ R(x) dr
0

where I is the length of the tube, S(x) is the source strength at point x,
and R(x) is a weighting function equivalent to the radial counting effici-
ency (sensitivity response to a point source as a function of distance from
the source-to-collimator axis). Since the source strength S(x) and the
count rate CR(x) found in the slit experiment are linearly related, we

can write

 

L
k [ CR(x) R(x) dx
5 - — , )
L
f R(x) dx
0

where k¥ is a proportionality constant relating S(x) and CR(x). Note that
k is the reciprocal of the counting efficiency for the source in the slit

experiment configuration and has the units of (y/in.)/c. We can now write

 

L
_ '([). CR(x) R(x) dx
— S _
Ck = z’ - . ’ (5)
[ R@) dx
0

where CR is the average count rate for each of the gamma rays for the total

source tube.
39

In determining the weighting function R(x), we assumed that the col-
limator-detector assembly had rotational symmetry; that is, only radial
displacement from the collimator axis was considered. (This assumption
required that the collimator hole be perfectly straight, which seemed rea-
sonable, at least for the 1/8-in. collimator.)

From the calibration work in the heat exchanger mockup (see Sect. 5.5),
we were able to measure the contribution to count rate at the detector from
sources to the right and left of the collimator axis. Figure 5.3 is a nor-
malized plot of the detector sensitivity to the silver source tube at var-
ious distances from the axis. This figure was made up as follows: For
three different '*°7Ag photon energies, the net count rates were added for
all the source tube positions in each column of tubes in the heat exchanger
mockup (parallel to the collimator axis, Fig. 3.5). The reason for summing
the readings of an entire column was to obtain a more uniform source dis-
tribution as well as better counting statistics. This count-rate distri-
bution across the mockup was then normalized to a central value of unity
for each of the three photopeak energies and corrected for a slight mis-
alignment of the collimator. Figure 5.3 is the average of the three nor-
malized curves; their deviation from this average was encouraging small,
The figure also represents the required weighting factor R(x). By multi-
plying the count rates along the source tube, as measured by the slit ex-—
periment (Fig. 5.2), in the appropriate way with the normalized count-rate
distribution as seen by the detector (Fig. 5.3), the result is the effective
activity of the entire source tube as seen by the actual detecting system
(Fig. 5.4). The average weighted count rate of the source tube for the
three energies considered is the area under each of the curves on Fig. 5.4
divided by the area under the normalized curve (Fig. 5.3). For the three

energies we find the following average weighted count rates:

 

Ener (keV Count rate (counts/min)
657.7 27,500
884.7 24,300
937.5 13,600

Since we now know the average count rates for the unshielded total

tube for three of the gamma rays of the source, we need only calculate the
RELATIVE COUNT RATE

0.9

0.8

0.6

0.5

0.3

ORNL-DWG 70- 9566

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

|“ TOTAL SOURCE LENGTH 6.3:in. : .e!

Fig. 5.3. Normalized detector sensitivity of gamma rays coming from
a source 15 ft away through a 1/8-in. collimator to the detector.

0%
COUNT RATE {counts/min)

ORNL-DWG 70-9567

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

    

 

 

 

 

 

 

 

 

7

 

30,000 ! ‘ w ! ‘
‘ | | :
; | | l
10m
A .
25,000 - + L E*_L _‘_‘ ]
\ ! o J
20,000 | - : | 1
| | - !
| ’ o
15,000 - : . + ' |
{
| | | |
! - | [ |
o
10,000 — - - | | .
| | | |
! S R
- . l ‘ |
5000 b —— S 937.5 kev”omug — . h l
|
0 ' ] ‘ ‘l r ‘
L* -~ ———————— TOTAL SOURCE LENGTH: 6.3 in. e — -
Fig. 5.4. Measured count rate from slit experiment multiplied by

normalized count rate distribution as seen by detector.
42

value of k to determine the average source strength. The value of kX can
be determined from the expression

__S(x)
K& = Tag

where S(x) and CR(x) denote the source strength in curies per inch and
counts per minute, respectively, at point x on the tube as determined by
the slit experiment. The subscript 7 denotes the 7th gamma and accounts
for the variation of % (a slit-experiment counting efficiency) with energy
and the fact that the three gamma rays measured each have different branch-
ing ratios (photons emitted per 100 disintegrations). Since we know the
source strength at two positions (see Fig. 5.2 and App. B), we can use the
results of the slit experiment to obtain the value of Xk at two positions

for each of the three gamma rays as shown in Table 5.1.

Table 5.1. Count rates measured at two positions along the
silver source for three different photopeaks

 

 

Count rate from Count rate from
Photon slit experiment slit experiment _
energy for 3.861 Ci/in. k(x 10%) for 4,196 Ci/in. k(x 10%) K
(keV) (counts/min) _(Cifin. (counts/min) _(Cifin.
counts/min) counts/min)
657.7 25,900 1.491 28,300 1.481 1.486 x 10™°
884.7 23,100 1.675 25,100 1.671 1.673 x 10™°
937.5 12,700 3.044 14,700 2.861 2.953 x 107°

 

We have listed the count rates and values computed for k for each of
the gamma rays for both "known'" positions of the source tube. In addition,
we have listed for each of the gamma rays the average value of k for the
two positions of the tube. Three values of the average source strength
were then calculated by multiplying these averaged k values times the cor-
responding averaged count rates for the three gamma rays given above. The

results of these calculations are:
43

 

Ener keV Source strength (Ci/in.)
657.7 4,09
884.7 4,07
937.5 4,00

From these results an overall average source strength was computed to be
4,05 Ci per inch of tube on June 1, 1969,

This source evaluation method leaves something to be desired and is to
a certain degree misleading because we used an essentially plane source
(source placed in all heat exchanger mockup positions) to evaluate the
weighting factor of a line source., However, since the normalized weighting
curve (Fig. 5.3) 1s quite steep and the source activity gradient in the
central portion of the source rather flat, the method should be acceptable.
Taking into account the errors of the radiochemical source-strength evalu-
ation and the weighting-factor evaluation, we estimate an error of 57 in

the source-strength computation.
5.4 Single Source Experiments

A single source experiment was performed to determine the efficiency
of the detector system for gamma rays coming from the main reactor off-gas
line. We .assumed that the detector system would have the same counting
efficiency for the **®"Ag source tube as for the corrugated 1-in.-ID main
off-gas line. The source tube was set up in an unshielded central position
of the heat exchanger mockup, and the collimator was aligned with the laser.
Minor collimator adjustments were made to cope with any misalignment by
searching for the maximum count rate for this setup. The distance between
the source tube and the detector was approximately equal to the distance
between the actual detector-collimator assembly on the PMS and the main
reactor off-gas line (line 522) in the reactor cell. Four spectra were
taken for this setup with the 440A amplifier and six with the 450. The
difference in response was small and judged to be due to statistical error.
A 1/8-in. collimator insert was used.

By taking into account the data of this experiment and the absolute
abundance of the different gamma peaks (15 photopeaks of **°"Ag), we were

able to calculate the detector-system efficiency at all these different
4

photon energies. This way we established an absolute efficiency curve from
the lowest ''°"Ag peak energy at 446.8 keV to the highest at 1562.2 keV.
The problem was to extend this curve beyond these limits, because many fis-
sion products have important photopeaks outside this energy range. It was
decided to use the actual fission product spectra taken from the off-gas

line for the extension of the efficiency curve. For example, with *°Mo,

131 132
I I

, , and *“°La, it would be possible to establish a relative effici-
ency curve all the way from 140 up to 2522 keV. By linking this wide-
energy—~range relative efficiency curve to the already established absolute
curve from ''°MAg, an absolute efficiency curve was obtained. It was very
encouraging to see that the relative curve from the fission products and
the absolute curve were almost identical in shape in the 450- to 1560-keV
range. Figure 5.5 is the efficiency curve adopted for the main reactor
off-gas line (line 522).

As a check on the above approach, we used also a 22¢

Ra source, which
emits photons over a wide energy range. The relative efficiency curve for
that source was essentially identical in shape with the established abso-
lute curve. The discrepancies, especially in the lower energy range, were
probably due to a different shielding of the actual photon emitter (for the
l1-in.-ID off-gas line, this would be the wall thickness of the pipe).

5.5 Heat Exchanger Calibration

The heat exchanger calibration was done with a full-size mockup of a
complete section of the heat exchanger. The 8 1/2-in.-~long dummy tube sec-
tions had the same outside diameter and wall thickness as the actual heat
exchanger tubes. In the mockup the tubes were held vertical and kept at
the specified triangular pitch of 0.775 in. by two horizontal grid plates
approximately 6 in. apart., With a simple tool it was easy to handle these
tubes from above through a work hole in the portable maintenance shield.
The mockup contained 20 rows of tubes (transverse to the detector-collimator
axis). Because of the triangular pitch, each row contained either nine or
eight tubes. A curved 1/2-in. stainless steel plate was placed in the
proper geometry in front of the mockup to simulate the heat exchanger

shell,
DETECTOR SYSTEM EFFICIENCY

300 500 700

Fig. 5.5.

DETECTOR SYSTEM EFFICIENCY =

!

ORNL-DWG 70-9568

COUNTS REGISTERED

 

PHOTONS EMITTED PER LINEAR INCH OF TUBE

 

900 {100 1300 1500 1700 1900 2100 2300 2500 2700
PHOTON ENERGY ( kev)

Absclute efficiency of detector system for photons emitted
from the 522 line.

2900

ah
46

Spectra were taken with the source tube placed in each of the tube
positions and dummy tubes in all the other positions. This procedure was
done twice, once with the 440A amplifier and once with the 450; the 1/8-in.
collimator insert was used. Some spectra were taken with the 1/16-in. col-
limator, but a full calibration was not performed.

By adding the count rates for each of the different *'°"Ag photopeaks
from the spectra of all the source positions in the mockup, a count rate
was obtained that represented the primary heat exchanger with an activity
of 4,06 Ci of **°MAg per inch on each tube, but with the internal shielding
of all the other heat exchanger tubes. By taking into account the data of
the experiment and the absolute abundances of the various 1197MAg gamma rays,
we could establish an absolute efficiency curve applicable to the heat ex-
changer for the energy range from 446.8 to 1562.2 keV. Again, there was
very little difference between the data obtained with the two different
amplifiers; we used the average of the two efficiency curves as our
standard.

Actual fission product data from the primary heat exchanger were used
to extend the energy range of the efficiency curve; we used the photopeaks
of *°Mo, '*'I, '32I, and, to a lesser degree, '“°La (because very little
'“°La was present in the empty heat exchanger). The shielding effect of
a heater element was measured and also calculated separately. The final
standard efficiency curve for the heat exchanger, including the heat ex-
changer shell and the heater box effects, is given in Fig. 5.6. It should
be noted that this curve, contrary to that for the off-gas line, bends down
at lower energies. The efficiency decrease at low energy was caused by the
internal shielding of the heat exchanger; hence the strong increase in the
photon attenuation coefficient with decreasing photon energy. This down-
ward trend with decreasing energy proved to be a problem in the initial
use of the computer program (see Chap. 4.4).

The above calibration procedure does not take into account the effect
of fission products deposited on the inside of the heat exchanger shell.
However, this amount, if comparable with the amounts deposited on the tubes,
would have only a minor influence on the total reading. Another limitation

was that the calibration would be reliable only if the deposition of fission
DETECTOR SYSTEM EFFICIENCY

10

{0

 

-7

300

500

DETECTOR SYSTEM EFFICIENCY =

700

Fig. 5.6€.

ORNL-DWG 70-9569

COUNTS REGISTERED

 

PHOTONS EMITTED PER cm? OF HEAT EXCHANGER TUBE

b

900 100 {300 {500 4700 {1900 2100 2300 2500 2700 2900
PHOTON ENERGY (kev)

Absclute efficiency of detector system for phctons emitted
from the primary heat exchanger.

LY
48

products were approximately uniform on all the tubes. Although this is
not necessarily true, we had no data to justify any other approximation.
As a check on our empirical efficiency calibrations, we have computed
from first principles counting efficiencies for both a line source corre-
sponding to the silver tube and the reactor off-gas line and a volume
source (truncated cone) corresponding to the cone subtended by the colli-
mator in the heat exchanger. These calculations are described in Appendix
D. We believe that the good agreement found between experimental and com-
puted counting efficiencies serves to confirm our measured values and in-

creases our confidence in the validity of our measurements.

5.6 Calibration of Shielding Materials

 

Throughout the experiments we tried to keep the count rate of the de-
tection system more or less constant in the range 3000 to 10,000 counts/sec.
Since the source strength varied with time and position, we varied the
detection-system efficiency accordingly by interposing different amounts of
shielding material between the source and the detector. In most cases,
disks, 1/2 in. thick and 7 in. in diameter, were placed over the work hole
in the PMS just under the collimator assembly. These disks could be easily
removed to permit use of the laser for location purposes. We used disks of
aluminum, copper, and lithium-impregnated paraffin. With some extremely
high activity levels, we also used some 1l/4-in.~thick plates of lead be-
tween the collimator insert and the detector. Aside from the very incon-
venient tendency for lead to attenuate especially well the lower energy
photons, the proximity of this shielding material to the detector also
caused the detection of lead x rays. In evaluating the influence of the
different shielding materials, we relied not only on the published attenu-
ation coefficients but we also calibrated the disks. We used both the un-

226pa source for this purpose. Our

shielded silver source as well as the
experimental values agreed well with the literature data for Pb, Cu, Al,
Cd, and steel; we had to rely on experimental values only for the lithium-
impregnated paraffin and the ceramic heat exchanger heater element. So-

called "shielding curves" were made for each of these different shielding
49

materials; these curves were plots of the shielding factors (reciprocals
of the more usual attenuation factors) as functions of energy from 140 to
2500 keV,

Absolute efficiency curves as functions of energy were synthesized
for all the different combinations of shielding material used in the ex-
periments. These were obtained by dividing, point by point, the unshielded
efficiency values by shielding factors appropriate for that shielding com-
bination. When two or more shielding disks were involved, composite

shielding factors were used. That is,

EF (E)
BE(F) = —H——
nia, 1"
where
EFS(E) = energy-dependent efficiency with shielding present,
EFM(E) = energy—-dependent efficiency without shielding,
Ai(E) = shielding factor for the 2th component of the shield,
Ni = number of pieces of component % used.

Altogether we used 12 different combinations of shielding for the main re-
actor off-gas line and 8 for the heat exchanger (see Tables 5.2 and 5.3).
The values of the efficiency curves generated for these shielding configu-
rations were used as input data to the computer analysis program to evalu-
ate the absolute amounts of nuclides present. The different efficiency
curves are given in Appendix C,

With the reactor system shut down and drained, we used primarily the
simpler cases of shielding with only aluminum or copper. With the reactor
at different power levels, we had to rely on much more shielding because
of the high gamma and neutron radiation levels. Gamma spectra taken at
the same place but at different times after reactor shutdown and recorded
with different shielding configurations yielded generally the same results
for the amounts of nuclides deposited at shutdown time; this gave us a

certain degree of confidence in the absolute efficiency curves used.
50

Table 5.2, Shielding configurations employed during
surveys of reactor off-gas line

 

 

Collimator
Shielding insert
case diameter (in.) Shielding material
1 1/16 1/8 in. steel
1/8 1/8 in. steel, 1 in. Cu
3 1/16 1/8 in, steel, 1 in., Cu, 2 in. paraffin
(Li), 1/8 in. Cd
4 1/8 1 in., Al
5 1/8 None
6 1/16 None
7 1/8 1/8 in. Cd
8 1/8 1/8 in. Cd, 1/8 in. steel
9 1/16 1/8 in., Cd, 1/8 in. steel
10 1/16 1/8 in. Cd, 1/8 in. steel, 2 in. paraffin
(Li), 1/2 in. Pb
11 1/16 1/8 in. Cd, 1/8 in. steel, 2 in. paraffin

(Li), 1/4 in. Pb
12 1/16 1/8 in. Cd, 1/8 in. steel, 2 in. paraffin (Li)

 
51

Table 5.3. Shielding configurations employed during
survey of primary heat exchanger

 

Collimator
Shielding case insert Shielding material
diameter (in.)

 

1 1/8 None

2 1/8 1 in, Al

3 1/8 1/2 in. Al, 1/2 in. Cu

4 1/8 1l in. Al, 1/2 in. Cu

5 1/16 None

6 1/16 1/8 in. steel

7 1/8 1/8 in. Cd

8 1/8 1/8 in. Cd, 1/8 in. steel

 

Extensive experiments were done to evaluate the effect of the 1/16-in.
collimator insert in relation to the 1/8-in, insert., During the calibra-
tion experiments, we found values appreciably higher than 4 for the decrease
in count rate when the 1/16-in. insert was used; also, there was consid-
erable scatter between individual measurements. It might have been that
this spread was largely due to minor misalignments between the collimator
axis and the center of the silver source tube. Therefore, another experi-
ment was done, this time by placing a lightly activated foil alternately
against the far ends of the 1/8- and 1/16-in. collimator inserts and the
detector placed at the front end. We found a shielding factor of 9.0 for
the 1/16-in. collimator insert in relation to the 1/8-in., insert. This test
was repeated several times, and the experimental values were within *10%.

Apparently the 1/16-in. collimator was not all that straight!
52

5.7 Calibration for Fission Gases in the MSRE

Because our calibration was done in terms of radioactive sources de-
posited on the coolant salt tubes in the heat exchanger, all results ob-
tained by the computer program, including those for gases, were expressed
in this manner. Although this "equivalent surface deposition' for gases
is a valid* mode of expressing their concentrations, it may for certain
purposes be more useful to express their concentrations in terms of free
cross-sectional volume of the heat exchanger. We have estimated the av-
erage ratio of free volume to surface area to be 0.55 cm®/cm®. Thus the
results for gases can be converted to disintegrations per cubic centimeter
per minute by dividing their values in disintegrations per square centi-

meter per minute by 0,55,

 

%
The validity of this mode of expressing gas concentration is easily
seen from the arguments in App. D.
53

6. MEASUREMENTS

Gamma-ray spectra were taken during the period of July 1969 through
December 1969. Altogether some 1400 spectra were recorded from several
different locations (Table 6.1). The purposes of the various sets of

measurements were as follows.

Group A — The reactor system had been drained and shut down since
June 1, 1969. The purpose of these spectra was to test the equipment and,
if possible, to study the deposition of longer-lived fission products,

Group B — During this period the reactor was usually at some power
level ranging from a few kilowatts to full power. By changing the reactor
conditions, such as fuel pump speed, helium purge rate in the fuel pump
bowl, and changing to argon as a purge gas, the concentration of the noble
fission gases was studied; of special interest was, of course, '*®°Xe,.
These spectra were all taken over a hole in the shield plug with the con-
tainment membrane in place and were considered in a separate report.

Group C — Calibration spectra were mainly taken with the *'°"Ag source
tube; some were taken with a *?®°Ra sourxce. Details of the calibration work
are described in Chapter 5.

Group D — A few spectra were taken through the shield plug over the
main reactor off-gas line during a beryllium addition to find out if the
noble fission gas concentration changed during and after an addition.
These results were also considered in Ref. 15.

Group E — Several spectra were recorded of the main reactor off-gas
line several hundred feet downstream of the fuel pump bowl. These spectra
were taken at a location where sample bombs of fission gases can be iso-
lated. Physically this location is in the vent house. The connection of
the off-gas line into the main charcoal beds is a few feet downstream from
this point. The purpose was to study the relative noble fission gas con-

centrations at this point.

 

157, R. Engel and R. C. Steffy, Xenon Behavior in the Molten Salt
Reactor Experiment, ORNL-TM-3464 (October 1971).
54

Table 6.1, Gamma-ray spectra recorded

July to December 1969

 

 

523 line roughing filters

Number
Group Place Time of
spectra
MSRE heat exchanger July 114
Main reactor off-gas line (522 jumper line) 20
Main fuel line (line 102) 5
Through shield plug hole over off-gas line Aug.—Sept. 235
Through shield plug hole over heat 15
exchanger
Calibrations (Chap. 5) Oct. 425
Through shield plug hole over off~gas line Oct. 9
(during Be addition)
Main off-gas line in vent house Oct. 6
Through shield plug hole over heat Nov. 18
exchanger
Through shield plug hole over off-~gas line 24
Through shield plug hole over drain tank 15
MSRE heat exchanger Nov. 241
Main reactor off-gas line (522 line) 91
Fuel pump bowl, fuel lines (lines 101 36
and 102)
Through shield plug over heat exchanger; Dec. 43
final shutdown
Gas samples, reactor cell Dec. 9
MSRE coolant salt radiator Dec. 5
Main coolant line Dec. 14
Samples of fuel salt and graphite Dec. 45
Miscellaneous: lube oil system, July—Dec. 10

 
55

Group F — From previous experience it was known that it takes at least
a few days from the moment of reactor shutdown and drain until the portable
maintenance shield is set up and operable over the reactor cell. In order
to get some data during this time, spectra were taken over the three holes
(off-gas line, heat exchanger, and drain tank) in the shield plugs. Reactor
shutdown and drain on November 2, 1969, at 1441 hr was initiated by thawing
the drain line freeze valve with the reactor still at full power. It was
judged that this type of drain would leave the maximum fission product de-
position in the system because the time span elapsed between the moment of
fuel drainage of the entire fuel system and full-power operation would be
shortest.

Group G — These are the main spectra taken with the detection equipment
mounted on top of the portable maintenance shield after reactor shutdown.

Group H — Several spectra were taken from the heat exchanger over the
hole in the shield plug. Spectra were recorded during full-power operation
before final shutdown, at shutdown, during draining of the system, and
thereafter.

Group I — Spectra were recorded from samples of the reactor cell air
after final reactor shutdown. These spectra were used in the analysis of
the piping leak that occurred after the shutdown.®®

Group J — Spectra were recorded of the coolant-salt radiator after the
system was drained; the detector was placed in front of the radiator tubes.
The purpose was to find any activated corrosion products that might have
settled in the radiator.,

Group K = Spectra were taken by C. H. Gabbard at two places along the
main coolant salt line to study the coolant flow rate by using the decay of
2°F and '°N in the coolant salt; *°F and *°®N are formed in the heat ex-
changer by delayed neutrons from the fuel salt.

Group L — Samples of fuel salt and graphite soaked in fuel salt were
taken from the fuel pump bowl and placed in the sampler enricher. Spectra

were taken by C. H. Gabbard with a special collimator and detector setup.

 

'®R. H. Guymon and P. N. Haubenreich, MSR Program Semiannu. Progr,
Rep. Feb. 28, 1970, ORNL-4548, p. 14,
56

Group M — Several miscellaneous spectra were recorded, such as from
the lube-oil system, the roughing filters from the containment ventilation
air exhaust, and from the off-gas line of the pump bowl overfiow tank (line
523).

The results of the analysis of the different spectra will be reported

by group in Chapter 7.
57

7. RESULTS

Presented in this chapter are all results which pertain to the distri-
bution of fission products in the MSRE. Brief mention is made of other
spectra, recorded for different purposes, but their analyses will be dis-
cussed elsewhere.

Results are reported in the order given in Chapter 6. Because of the
large number of spectra, results have been presented, where possible, in
graphic form. 1In general, all results have been included in the figures,
irrespective of possible experimental errors. Results that are somewhat
uncertain for a particular reason are specifically indicated. Only in very

few cases, because of obvious error, were data omitted.

7.1 Group A Spectra

These spectra were recorded during the July shutdown period and con-
cern the primary heat exchanger, the main off-gas line, and a fuel salt
line.

The shakedown of the equipment proved to be very useful, in that
problems with the analyzer and detector system were resolved; the alignment
and locational equipment, including the laser, proved to be sturdy and
reliable,

Most spectra were taken six or more weeks after reactor shutdown
(June 1, 1969); this implies that only a few longer-lived isotopes might
be expected to be present. The spectra were relatively simple and contained
no multiplets. The system had not been flushed with flush salt,

For nuclides emitting photons at different energies, we used the av-
erage of the computed activities yielded by the most prominent photopeaks
emitted by that nuclide.

The counting time was mostly 200 sec per spectrum. Although '*7Cs
and *“°Ba-La were identified a few times in the heat exchanger, their ac-
tivity level was so small that one might tentatively conclude that they
were deposited as a result of the decay of xenon present in the system

after shutdown and drain. No ?°Zr could be detected. It should be borne
58

in mind that, except where specifically noted, all results in this section
have been extrapolated back to reactor shutdown time by simple exponential
extrapolation. The reason for this extrapolation was to compare results
yielded by spectra taken at different times. This extrapolation might give,
however, an activity value at shutdown time that is too high because of the
decay of a precursor nuclide. For example, the extrapolated value of *®'I
might be too high by a few percent because of the decay of '*'Te., This

will be further discussed in Chapter 8.

7.1.1 Heat exchanger

About 65 spectra were taken along the longitudinal axis of the heat
exchanger, Another 49 were recorded by moving the detector transversely
across the heat exchanger; the transverse scans were taken at four differ-
ent places. Most spectra could be successfully analyzed. The quantitative
interpretation of the transverse-scanning results proved to be rather dif-
ficult, due to the varying shielding condition because of the changing num-
ber of heat exchanger tubes seen by the detector as well as the changing
photon attenuation through the heat exchanger shell. Qualitatively, the
fission product deposition was symmetrical on both sides of the longi-
tudinal axis of the heat exchanger.

The spectra recorded above the heater connector boxes (HTR plug)
proved to be of no use for quantitative interpretation because of the un-
known amount of shielding involved. All these spectra gave results that
were far too low in comparison with the other spectra. Since the "ac-
tivity depressions' near these boxes all had about the same shape, it was
decided that results from spectra recorded above the heater boxes were
biased because of the boxes and not because of different deposition of fis-
sion products in the heat exchanger. These spectra were discarded. Let
us now turn to the different nuclides identified.

Niobium-95 (Fig, 7.1). The disintegration rate along the longitud-
inal axis ranges approximately between 0.10 and 0.30 x 10'® dis/min per
square centimeter of heat exchanger tube. There appears to be an increase
of activity close to the baffle plates in the range of 0.25 to 0.30 x 10'*

dis min~™* e¢m™?, Between the baffle plates, the average activity level is

near 0.10 to 0.15 x 10'? dis min~> cm™®., Since the activity increase is
ORNL-DWG 70- 9570

 

 

 

3

| T T N T
LIL D 0.30 Ef2 n ol S | | |
}‘— '

o x | *] . .rfil |
~_ w 0.20 Ef2 . o * ., " i o i ’
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- g 0.10 Ef2 1 o l
E —
~ 5 | ]

T

|

HTR. PLUG SPACER

HTR. PLUG HTR. PLUG
|

 

 

  

FUEL SALT FLOW —=
|
L ||

FPig. T.1. Activity of ?°Nb at reactor shutdown on June 1, 1969, in
the MSRE heat excharger.

 

PEe—
L N

6§
60

more pronounced for those baffle plates against the upper side of the shell,
one may conclude that this activity increase is rather localized around the
low flow rate areas near the shell and the baffle. There does not appear

to be any general change in activity along the length of the heat exchanger.

Ruthenium-103 (Fig. 7 2). The disintegration rate (dis min=* cm™?

 

along the longitudinal axis follows nearly the same pattern as that for

°SNb: along longitudinal axis, 0.17 to 0.65 x 10*'; near baffle plates,

0,50 to 0.65 x 10°'; between baffle plates, 0.17 to 0.24 x 10'*,
Ruthenium—Rhodium-106 (Fig, 7.3). Ruthenium-106 is basically identi-

 

fied by its short half-life decay product, *°°Rh. The activity distribution
(dis min~’ cm™®) is essentially the same as for *°’Ru: along longitudinal
axis, 0.04 to 0.14 x 10°*; near baffle plates, 0.11 to 0.14 x 10*'; between
baffle plates, 0.04 to 0.06 x 10°*.

Antimony~125. This nuclide was detected in small quantities all along
the heat exchanger. The activity level, probably because of the very low
fission yield, was small; the computer program only sporadically found
peaks that could be assigned to '?°Sb, Variation of the activity level
along the longitudinal axis was 0.053 to 0.14 x 10%° dis min~* cm™?2.
There were not enough data to make other observations concerning the dis-
tribution of the deposition.

Antimony-126 and Antimony-127. These nuclides were both identified a

 

few times in the different spectra. However, it does not appear sound to
assign a quantitative value to those peaks in view of the very low activ-
ity left after such a long decay time.

Tellurium-129m (Fig, 7.4). The fact that both the fission yield of

 

"29"MTa and the abundance of its photopeaks are small makes it difficult to

detect this isctope with good precision (see also Sect. 4.3). There ap-

pears to be no strong tendency of an activity increase near the baffle

plates; however, a slight increase in activity near the cold end of the

heat exchanger seems apparent. Variation of activity along the longitudi-
2

nal axis is 0.090 to 0.26 x 10'* dis min=' cm™2.

Iodine~131 and Tellurium-Iodine-132. Most of the '3®'I and '?2*Te-I

 

had decayed, although these nuclides could be positively identified in

several spectra. The average activity of ***I at shutdown was approxi-

mately 0.2 x 10'* dis min~* cm~*. The average activity of *®2?Te~I at
-cm? OF HEAT
EXCHANGER TUBE

dis / min

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

ORNL~ DWG 70-9571

| i T
060 El b t e e I e - e 1
. - S | e | |
. ! I I
0.40 E .o e J—.——f * ; * |
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0.20 EN [— — — | . i - : . |
| | \ |

| | i | | |

| | | | | |

HTR. PLUG HTR. PLUG : SPACER HTR. PLUG l SPACER |

1 |

 

    

|
FUEL SALT E-I_LOW —

Fig. T7.2. Activity of 1033y at reactor shutdown on June 1, 1969, in
the MSRE heat exchanger.

 

N N, ,

19
     
  
 

 

    
    

EE ORNL-DWG 70-9572
Ll % [ : | B T C
T 5 140 E10 ¢ ! i ol ‘ |
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' | i ; | ' Nl
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s ' - ‘ |
- | | | 1]

Fig. T7.3. Activity of 106Ru~Rh at reactor shutdown on June 1, 1969,
in the MSRE heat, exchanger,

29
dis /min * cm2 OF HEAT

ORNL- DWG 70-9573

 

   
   
   
 

 

 

 

 

 

 

 

 

EXCHANGER TUBE

 

 

    
 

' T T l
0.30 EM : i , T o T
- Ll *
® T I
0.20 Ef i S - A |
P L | b
0.10 E# "o . .- L |
t - r { f
| | :
| | |
HTR. PLUG |~|+ HTR. PLUG E E
|

 

   
 

 

 

|
FUEL SALT FlLOW—-

Fig. T.L4. Activity cf '*°MTe at reactor shutdown on June 1, 1969, in
the MSRE heat exchsanger.

 

 

 

£9
64

shutdown was approximately 0.4 x 10'? dis min~' em~?. Taking into account
the relatively short half-life of these isotopes in comparison to the decay

time, not much precision can be expected of these values,
7.1.2 Main reactor off-gas line

The flexible 1-in.-ID corrugated tubing, called the jumper line, was
scanned, This 2-ft-long jumper line connects the main off-gas outlet at
the fuel pump bowl with the 4-in.-diam section of the main off-gas system.
This is about the only part of the off-gas system that could be "seen" from
the portable maintenance shield and hence was the only piece surveyed.

The 20 spectra were taken about six weeks after reactor shutdown (shut-
down of June 1, 1969) at approximately l-in. intervals along the jumper

137
Cs

line, Although there was some scatter in the data, especially for
and *“°Ba-La, we could not detect a definite trend of decreasing activity
in the downstream direction along the line. Although cesium, barium, and
lanthanum were detected, these apparently were deposited in the off-gas
line by the decaying xenon stripped from the salt rather than from salt
deposits. Actually, no photopeaks could be positively identified from a
nuclide that would indicate the presence of fuel salt; for example, °°Zr
could not be positively identified\* Table 7.1 lists the nuclides that
were identified; the activity is the average of the 20 different spectra.
These data are extrapolated to the moment of shutdown (June 1, 1969) and
represent disintegrations per minute per inch of off-gas line. It should
be noted, however, that the main off-gas line was partially plugged during
the latter part of the power operation prior to reactor shutdown. Off-

gases were then routed through the fuel-pump overflow-tank off-gas line.

7.1.3 Main fuel lines

Only a few spectra were taken from the main fuel line (line 102).
Given the different counting geometry, due to the distance to the detector
and the shape of the source, these results could only be qualitative.

However, based on experience of the calibration experiments, we can expect

 

%
The minimum activity at which a peak could be identified in this case
was «0.10 x 10°° dis/min per inch of line.
65

Table 7.1. Nuclide activities found in the jumper line of
the main reactor off-gas system at shutdown
on June 1, 1969

 

Activity Standard deviation
Nuclide (dis min-? in.-1') of the average value
(dis min~—* in.” %)

 

*3Nb 0.22 x 10*?® 1.0 x 10**
1032Ru 0.45 x 10*3 2.8 x 10*?
1¢€Ru-Rh 0.19 x 10*? 0.8 x 10**?
123gp 0.36 x 10*? 5.0 x 10°
128MTe 0.55 x 10%? 5.4 x 10*°
1317 0.95 x 10** 6.2 x 10*°
137Cs 0.53 x 10*2 11.0 x 10*°
14%Ba~La 0.44 x 10*3 5.7 x 10**t

 
66

that the detection system has a fairly constant efficiency in the energy
range 500 to 1000 keV. This would enable us to compare the activities of
those nuclides that emit photons in that energy range. All nuclide activi-
ties have been made relative to ?°Nb, because this isotope had the most
prominent photopeak in the spectra at 765.8 keV.

The nuclides identified, along with their relative activities, are
given in Table 7.2. Restraint should be applied in using these numbers!
It is intended to use these ratios only for comparison with the nuclide

activities found in the heat exchanger and fuel line spectra taken in

November 1969,

Table 7.2 Relative nuclide activity found in the main fuel
line (102) at shutdown time on June 1, 1969

 

 

Photopeak energy used in Activity relative
Nuclide activity calculation (keV) to °°Nb
>3Nb 765.8 1.0
193Ru 496.9 0.7
*°€Ru~Rh a 0.1
1238h 427.9 0.02
127gh b
12°MTe 459.6 0.4
13171 364.5
1927e-T e 7.0

 

aAverage of 511.8 and 621.8.
Activity too small for proper calculation.

“Average of 667.7, 772.6, and 954.5.
67

7.2 Group B Spectra

These spectra were recorded through shield plug holes on the main off-
gas line and heat exchanger. As mentioned earlier, holes were drilled in
the lower and upper shield plugs to permit the recording of gamma spectra
from the main reactor off-gas line, the heat exchanger, and one of the
drain tanks. There was one through-~hole over each of these components;
the steel containment membrane between the lower and upper shield plugs
was, of course, left in place. Because of this membrane, the aiming of
the collimator had to be done by trial and error rather than with the laser
jig. We tried to obtain the maximum reading on a particular location by

moving the collimator and detector back and forth.
7.2.1 Main reactor off-gas line

Most of these spectra were recorded with the reactor at some power
level and under different reactor conditions. The variable reactor param-
eters were: use of argon instead of helium as purge gas, variations in
pump speed, and different reactor power levels. The purpose of these spec-
tra was to try to find a relation between the different noble fission gas
concentrations (kryptons and xenons) and the changing reactor parameters.
Especially at the higher power levels, the radiation intensity from the
off-gas line was very high, in the order of 1000 R/hr or more. Fair doses
of fast and thermal neutrons were also detected. This necessitated the ex-
tensive use of shielding materials, and even then some spectra contained
too many overlapping peaks for proper analysis but were still useful for
comparative purposes.

The different noble fission gases could be identified in most spectra
and their concentration calculated in a major part of the spectra. These
data were examined in connection with a study of '>°Xe behavior in the re-

actor.'’ However, the values were too scattered to permit detailed analysis.

 

'77. R. Engel and R. C. Steffy, Xenon Behavior in the Molten-Salt
Reactor Experiment, ORNL-TM-3464 (October 1971).
68

7.2.2 Heat exchanger

These spectra were taken to determine the change in fission product
deposition on the heat exchanger after flush salt was circulated through
the primary system, Spectra were taken before and after the flush-salt
operation. Although the analyzed spectra did show the presence of noble
metals, the activity was so low that it must be assumed that the collimator
was improperly aimed and that we analyzed the general background in the
high-bay area. This was confirmed by the analyzed data; there was no
change in the activity of the considered nuclides before, during, and after

the flush-salt circulation.

7.3. Group D Spectra

 

These spectra were recorded through the shield-plug hole above the
main reactor off-gas line during a beryllium addition. The purpose was to
evaluate the possible change in noble fission gas concentration during and
after a beryllium addition, but, as with the group B spectra, the scatter

in the results precluded the determination of small changes,

7.4 Group E Spectra

 

Just upstream of where the main off-gas line ties into the charcoal
beds, there is a facility for isolating samples of reactor off-gas in any
of three sample bombs. A hole was drilled in the shielding over the center
sample bomb, and the off-gas flow was temporarily diverted and forced to
flow through this bomb. Although no quantitative results can be expected,
it is of interest to determine the ratio of the different fission gas con-
centrations. Heavy shielding was necessary to cope with the high radiation
level, As might be expected, some of the decay products of those identi-
fied noble gases were also found. Because of the temporary diversion of
the main off-gas flow, the quantitative activity value of these decay prod-

ucts does not bear much importance,
69

Table 7.3 gives the ratio of activities of the identified noble fis-
sion gases in relation to the ®°°Kr activity. It should be noted that the
values of ***"Xe and '®®Xe are to be considered with less confidence.
There was considerable scatter in the value of *®*Xe: +50%; the °°®Xe

might be off as much as *60%.

Table 7.3. Ratio of activities of identified
noble fission gases in relation
to the °®Kr activity

 

Activity relative

 

Nuclide to ®°Kr
87Kr 0.8
®8Kr 1.0
®°Kr Identified only
135Mya 0.05
135%e 0.6
13%%e 0.07

 

7.5 Group F Spectra

These spectra were recorded through the shield plug holes of the heat
exchanger, the main off-gas line, and the drain tank. From previous ex-
perience we knew that it would take at least two days to remove the upper
shield plugs from the reactor cell, cut the containment membrane, and set
up the portable maintenance shield (PMS). The holes in the shield plug
would give us a convenient opportunity to record spectra at those locations
during the time-span between reactor shutdown and installation of the
equipment on top of the PMS. Drain tank data were taken through the hole

only.
70

There was some disagreement between the data recorded through the
holes from the heat exchanger and off-gas line compared with the data taken
at the same locations with the PMS installed. It seemed that the disagree-
ments were due to misalignment of the equipment, and the results were ad-
justed accordingly by normalizing to a known nuclide (°°Nb).

Several nuclides that otherwise would have been totally decayed could
be identified and their activities estimated. The detector and collimator
were moved several times during this two-day period from one location to
the other.

Reactor shutdown and drain from full power occurred on November 2,
1969, at 1441 hr. The procedure was as follows: with the reactor still
at full power, the thawing of the system drain valve was requested. Once
this freeze valve was thawed, the drain of the fuel salt caused a drop of
the fuel=-salt level in the fuel pump bowl which automatically stopped the

fuel pump and then scrammed the reactor.

7.5.1 Main reactor off-gas line

 

The collimator and detector were set up over the off-gas line hole
two days before the actual shutdown and drain of the system. Several spec~-
tra were thus recorded with the reactor at full power; heavy shielding was
necessary for the recording of these spectra.

Because of the very large number of photopeaks in the spectra during
and shortly after reactor operation, it is obvious that many minor peaks
would go undetected in the analysis and sometimes would be added by the
computer program to the areas under the larger peaks. This leads to a cer-
tain overestimation of some of the nuclides present. TFor the evaluation
we have tried to select those photopeaks that seem to be relatively iso-
lated. Another check on the validity of a selected photopeak was the iso-
tope half-life time as deduced from the presented graphs.

Because of nuclide activity variations due to sometimes short-half-
lived precursors, all these off~gas line data are presented as activities
measured at the moment of counting time rather than extrapolated back to

reactor shutdown time.
71

First, the spectra were recorded with the reactor still at full power,
and from these the average activity was calculated for the identified noble
fission gases. Although many isotopes with longer half-lives were also
identified, their activity appeared to be highly influenced by the many
large photopeaks of the noble fission gases. An exception was °°Nb, with

1

a calculated average activity of 0.99 x 10'? dis min™* in.”", which is close

to the "best estimate" of the ®°Nb activity of 0.90 x 10*? dis min~' in.”*
given in Section 7.6.2; all activity results are normalized to this last
°5Nb activity. Table 7.4 shows the activities of noble fission gases de-
tected in the off-gas line with the reactor at full power; we estimate the
uncertainty of these results to be *40%.

The activities reported in Table 7.4 appear to be much too high. If
one calculates the maximum nominal disintegration rate of these fission
gases in this section of the off-gas line, the values in the table appear
to be a factor of 10 to 50 too high. The nominal calculation takes into
account the yield of the fission gases and the reactor power and allows
no holdup time in the primary loop. We found no way to explain this dis-
crepancy except to postulate a longer than normal residence time for these
fission gases in the off-gas line §such as might be produced by adsorption
on the wall or on other deposits that were known to be present).

Let us now consider the activities of different nuclides after reactor
shutdown. These spectra, being recorded through the shield plug hole, have
also been normalized to the best estimate of the °°Nb activity.

It appears that most photopeaks from spectra taken at 1.04 and 1.49
days after reactor shutdown yield results that are too low. Although the
cause 1is unknown, little confidence should be placed in the results calcu-
lated from these two spectra; these points are indicated in Figs. 7.6 to
7.20 as black points.

Krypton-87 (Fig. 7.5). There is a rapid activity decrease at the
start of the fuel-salt drain, subsequent fuel pump stop, and reactor shut-
down. The raising of gas bubbles from the salt after the fuel pump stop,
the release of fission gases from the graphite, and the possible back surge
of decaying fission gases back into the system because of drain might all

be contributory to the first rapid and then slower decrease of activity.
72

Table 7.4, Noble fission gas nuclide chains identified
in the main reactor off-gas line with the reactor
at full power

 

 

Nuclide Average activity (dis min=* in.”?
87kr 0.13 x 10*®

88Ky 0.12 x 10*°

®®Rb 0.18 x 10**

®9Kr 0.22 x 10*%%

#%Ru 0.82 x 10**

°Okr 0.73 x 10**

>°Nb 0.90 x 10'°

133%e 0.13 x 10**

1357y e 0.50 x 10**

138%e Identified only, order of activity: 10%*
13%%e Identified only, order of activity: 10%3
13%¢Cs Identified only, order of activity: 103

149%e Identified only, order of activity: 10%?

 
73

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|
; j : ;
1010 ; | i !

 

 

 

 

0.4 0.6 0.8 1.0 1.2 1.4 1.6 1.8 2.0
DECAY TIME AFTER REACTOR SHUTDOWN {doys)

o
O
n

Fig., 7.5. Activity of 87Kr end '35Xe after reactor shutdown in the
jumper line of MSRE main off-gas line (November shutdown).
74

Krypton-88—Rubidium-88 (Fig. 7.6). There is a similar rapid activity

 

decrease with °®Kr; the °°Rb activity decreases almost at the same rate as
88,
I\r-

Strontium-91 (Fig. 7.7). This nuclide is a decay product of the °'Kr

 

chain. The decrease in activity fits the decay half-life well, which would

confirm its identification., Extrapolated back to zero time, one can esti-

mate its maximum activity to be 0.20 x 10'® dis min™'! in.”'. This extrap-

olation seems reasonable if one considers the short decay half-1life of its
precursors.
Niobium-97. This nuclide was identified only a few times directly

after reactor shutdown. Its activity was in the order of 10'* dis min™*

. -1
ln. -

Molybdenum-99 (Fig. 7.6). Although a few of the data appear somewhat

 

too low, most of the activity results decrease according to the °°Mo decay
half-l1ife. From the actual survey of the jumper line, we estimated the
°°Mo at the moment of shutdown to be 0.33 x 10'®*, The activities plotted
in the figure agree well with this estimate.

Ruthenium-105—Rhodium-105 (Fig. 7.8). The calculated activity of
195Ry is deduced from two of its photopeaks; these agree well. The decay
of the activity conforms very well to the literature data for its decay
half-life. Extrapolated back to reactor shutdown time, the estimated maxi-
mum '°’Ru activity is 0.34 x 10%? dis min~' in.”%.

The daughter nuclide, *°°Rh, is rather difficult to identify because

its photopeaks are not very prominent and are close to peaks from other

nuclides; one might expect erroneous results. The maximum '°5Rh activity
is approximately 0.8 x 10*° dis min~* in.”'., The amount of '°°Rh present
in the off-gas line could be due to the buildup from the decaying *°°Ru as

'°°Rh separation from the salt.

well as from the separate mechanism of
Antimony-129—Tellurium-129m (Fig. 7.9). In a few spectra we could

identify '2°Sb; the lack of dependable nuclear data (decay scheme and

abundance of its gamma ravs) made it impossible, however, to assign an ac-

tivity to this nuclide.
O—h
o N

—

1)

~o

NUCL IDE [ (dis/min)/in. of of¢-gas line]
~

o

ORNL-DWG 70-9582

 

ACTIVITY OF
|

Kr WITH REL\CTOR AT 8 MW

 

 

 

 

 

 

 

 

 

I
g e ; :
+ e C————
—t o 4
- e :,
e : Lo

 

. —
— - +
—_— -~ + 4

 

 

 

 

T

~ACTIVITY OF 88Rb WiTH
REACTOR AT 8 MW

 

-
>

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

o 1529.8 kev ——  —— s

G/

 

 

 

 

 

 

 

 

b 23920 kev

 

 

 

 

 

¢ 898.0 kev
¢ 1836.1 kev
A 739.7 kev ! ; _
A 778.2 kev R Do —
FE e e
b L

 

e 7T77 e - § e .,.—%—7,._~.a.._..._._g b e
; .
| .

 

b i
3
| i
!
0 0.2 0.4

Fig. T7.6.

0.6 0.8 1.0

DECAY TIME AFTER REACTOR SHUTDOWN

Activity of 88Kr,
MSRE main off-gas line (November shutdown).

2.0 2.2 2.4 2.6 2.8 3.0
(days)

and ?°Mo in the Jumper line of the

 
ORNL-DWG 70-9583

LT T T | - i et { | “W'W‘:;"”W' ]

] ‘ . R S | ',,44
| } | i Mgr. o 1024.2 kev |

: ‘ 0 B847.0key |- At ¢
13411 ¢  884.1kev
4 1072.5 kev :

 

 

 

 

 

 

 

[ e e R - S
+ . o - S e - S i e o -
|

b e e e - e . < o
* g b boee e e
b
.

go\q e e i “{, - . e e e e e v [ EEL - M
k.
1,
f
§

 

 

 

 

{

o
|
| |
|

 

 

 

 

 

 

 

 

g o s,

 

 

 

 

 

 

NUCLIDE [(dis/min)/in. of off-gas line

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

f
b
| | | ; |
1010 l ) i 1 L I z |
0 0.2 0.4 0.6 0.8 1.0 .2 1.4 1.6 1.8 2.0 2.2 2.4 2.6 2.8 3.0
DECAY TIME AFTER REACTOR SHUTDOWN (days)

 

Fig. 7.7. Activity of 2'Sr and '**I in the jumper line of the MSRE
main off-gas line (November shutdown).

9.
NUCLIDE [ (dis/min) /in. of off-gas line |

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

43 o ORNL-DWG 70-9584
1O g ——a - 4 T T
_?)_g . .w : 1 - ifi*“— — : ] _ _
qo-0 L e ; _77;__»,,__,, _
5 .- & & ] | 4 |
| { _ o ! AH 5
e ' } —— e S
— . 5
2 -Q- f t @
N+ ‘ |
1012 & ‘ : | ; ‘
B e , ‘ % e
| - 7H¢ ?\L‘i : 105Rh 0 306.3 kev ‘ . 7_77 :
5 ¢ - & 319.2 kev
; | ; | (04694 kev
- e * P e 109-y 14 676.3 kev - - e
i | | | 2 4 724.2 kev
21 v T T | |
| | 1 } o
: | i
ol o .
0.2 04 Q6 08 1.0 1.2 1.4 1.6 1.8 2.0 2.2 2.9 2.6 2.8 30

Fig. 7.8.

DECAY TIME AFTER REACTOR SHUTDOWN (days)

Activity of 195Rh ana !'°°Ru in the Jumper line of the MSRE
main off-gas line (November shutdown).

LL
of off-gas line]

NUCLIDE [(dis/min}/in.

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

s ORNL -DWG 70-9585
10 T T T 1 : T ,
. SR ok . Lo - '.,*,, el ,,,,,,%,.,, i
e . {_ - I . e e z 429Te; 0 459.6 kev R ! b
- e -4 i . e S — e 'f .- T —
5 - e e e - e e e = — - . --4‘—----— R
- o I 0 793.8 kev ‘
| ' T 13mredd 852.3 kev ’j
T Ty 4 1206.6 kev - -
O i E ‘ i
\O " '
! .o :
0% L ” o ; i ol
e e e e e P e i - ]
s| 0 ; ¥ I
| e e : ; —
. . v | | .
S | 0 s
ol b I R
|
i s
‘ ! ! !
1ot j e } | ;
0 0.2 0.4 0.6 0.8 1.0 1.2 1.4 1.6 1.8 2.0 2.2 2.4 2.6 2.8
DECAY TIME AFTER REACTOR SHUTDOWN (days)
Fi .o 129 1317y, .
ig. 7.9. Activity of Te and e after reactor shutdown in the

jumper line of the MSRE main off-gas line (November shutdown).

3.0

8L
79

Tellurium-129m was detected by its decay from the isomeric state to
the ground state and the subsequent disintegration of the ground state.

We identified '?°"Te primarily by the most abundant photopeak of the '#°Te
daughter at 459.6 keV.

In the decay of *?°8b, the branching ratio is 85% to the ground state,
12°Te, and 15% to the metastable state, **°"Te, Since the half-life of
129Ta jitself is short, the decay of its 459.6-keV photopeak should, after
a short time, follow a composite of the decay half-lives of 12%5b and '*°MTe,
Figure 7.9 shows the activity related to '*°Te.

The low yield of only 7.7% for the 459.6-keV photopeak in the decay
of '?°Te implies also that any experimental scatter is amplified by more
than a factor of 13 for the activity calculation. This would, of course,
explain the state of the results.

The '2?°MTe activity can be deduced from Fig. 7.9; this activity ex=-
trapolated to zero time agrees reasonably with the estimate of Section
7.6.1. The initial faster decrease of activity of **°Te, with a half-life

129

of about 5 hr, could possibly be attributed to the Sb decay.

Tellurium-131m (Fig. 7.9). Based on the results from three *?'MTe

 

photopeaks, the activity of this nuclide could be well established. Its
identification was confirmed by the decay half-life as deduced from Fig.
7.9. Extrapolated back to reactor shutdown time, the estimated maximum
activity of *?*Te is 0.75 x 10*? dis min~' in.~'. The maximum activity
calculated by extrapolation to reactor shutdown time is acceptable since
one can discount the effects of an eventual ®?Sb decay. Only a few photo-
peaks in the different spectra could be identified as '317e; these results
were too sporadic for any interpretation.

Todine-131 (Fig. 7.10). With the exception of two sets of data (1.04
and 1.49 days after shutdown), the results are in good agreement. It ap-
pears that the maximum activity occurs shortly after reactor shutdown time.
This maximum activity is 0.5 x 10*® dis min™' in.”', somewhat lower than
the best estimate (Sect. 7.6.1) of the jumper line survey. It might be ex-

13MMTa, Another argument is that

plained by the influence of the decaying
1317 pight “evaporate" or transfer from the hot, empty fuel system and set-

tle in the off-gas line (see Chap. 8).
NUCLIDE [(dis/min)/in. of off—qgas line]

1012

ORNL—DWG 70— 9586

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

. 1 { 1 z I [ T
| : . : ‘ |
J \ 5 ; ; 131, [© 364.5 kev
" o * L j ‘ } ®637.0 kev —— ! )
o QO g o 8 | 1 . . i ' % g e e o ‘
. | | | ] ¢ W
| | | % ‘
| B K |
| | T ‘
i | , | ! }
! ‘ . [ : ' |
i | | | | |
0 0.2 0.4 0.6 0.8 1.0 1.2 1.4 1.6 1.8 2.0 2.2 2.4 2.5 2.8 3.0

DECAY TIME AFTER REACTOR SHUTDOWN {days)

Fig. T7.10. Activity of 1317 gfter reactor shutdown in the Jumper line
of the MSRE main off-gas line (November shutdown).

08
81

Tellurium-132—TIodine-132 (Fig. 7.11). There is considerable scatter

 

of both the !32*Te and '°2?I results in the first hours after reactor shut-

down. During the first 10 hr, disintegration rates vary between 0.45 x 10*°

132

and 1.3 x 10*® dis min~' in.—'., Since three I photopeaks seem to yield

consistent activities among each other, one has to accept apparently an

132
f

extra increase o I activity after shutdown. This cannot be explained

132 132

from the decay of Te and I alone. This surge in activity could be

132

conveniently explained if it is assumed that some of the I formed by
decay of '32Te in the hot, empty fuel system transfers to the off-gas line
along with the circulating purge-gas flow. An extrapolation to reactor
shutdown time, based on the data taken after 2.3 days and taking into ac-
count an equilibrium condition between '*2Te and '®2%I, yields 0.67 x 10*®
dis min=* in.~', in agreement with the estimate in Sect., 7.6.1 (0.65 x
10*®), However, if one admits a transfer of iodine to the off-gas line,

a maximum *?*I activity of 0.12 x 10*“ dis min~' in.”' can be expected.

Todine-133 (Fig. 7.12). The results are in reasonable agreement with

 

the *?°I decay half-life. The activities calculated during the first 6 hr

f 133

seem to be low, which might again indicate that some transfer o I occurs

from the reactor system to the off-gas line after reactor shutdown time.

Taking into account an equilibrium condition of Te-I, and the possible

133

iodine transfer, it is estimated that the maximum I activity after re-

actor shutdown is 0.17 x 10*® dis min-! in.™?.

Todine-134 (Fig. 7.7). The later spectra yield activities that are

 

in very good agreement with the expected decay half-life value. It is esti-

1347 activity after reactor shutdown is in the order

mated that the maximum
of 0.30 to 0.40 x 10'3® dis min~?! in.”*.

Xenon-135 (Fig. 7.5). Again there is a rapid decrease in activity di-

 

rectly after reactor shutdown which might be explained by the different
circumstances related to the stoppage of the fuel pump and the drain. Even
several hours after reactor shutdown and drain, the '®°Xe activity is still

appreciable,
NUCLIDE [(dis/min)/in.of off-gas line]

ORNL-DWG 70-9587

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

DECAY TIME AFTER REACTOR SHUTDOWN (days)

Fig. T.11.

J | o 667.7 kev i i
_ | 321 J4 9546 kev 1 J
;{> $ gg ! 4 . #4398.6 kev . 3 R S
8-- —£~~~r : 4 . 3271 52282 kev | T ?
g___ 00 o O + , .o :; i+ e 4{. e ' |
. - - . o ; . . . | ;?1?
; t 'I' b e o
¢ , ; .
‘ | |
i :
i .
| x | A
j 1 oo : :
? f J b - :
+ : - - + + - l -—
! |
1 | | i
i | | |
! ] ' ' 1 L [ | i
0.2 0.4 0.6 0.8 1.0 1.2 1.4 1.6 1.8 2.0 2.2 2.4 2.6 2.8 3.0

Activity of 1327e and '3%I measured after reactor shutdown

at the jumper line of the MSRE off-gas line (November shutdown).

c8
NUCLIDE [(dis/min )/in. of off-gas line]

0.2

 

0.9 0.6

Fig. T.12.

ORNL-DWG 70-~9588

 

—————— L +

1331 6 529.9 kev
1
|

 

 

 

 

 

 

 

0.8 .0 .2 1.4 1.6 1.8 2.0 2.2 2.4 2.6 2.8 3.0
DECAY TIME AFTER REACTOR SHUTDOWN ({days)

Activity of 1331 after reactor shutdown in the Jumper line

of the MSRE main off-gas line (November shutdown).

£8
84

7.5.2 Heat exchanger

With the reactor at full power, the radiation coming through the
shield plug hole over the heat exchanger was too high to record any spec-
tra. The main problem was that the setup of the collimator with detector
caused such a high dose rate of scattered radiation during the alignment
procedure that it was judged unwise to continue as long as the reactor was
at power. Had the collimator and detector been set up and aligned before
the reactor was at power, we could have recorded spectra; however, we found
in later experiments that spectra taken over the heat exchanger with the
reactor at full power could not be analyzed anvhow because of too many peaks.

This situation also prevented the recording of heat exchanger spectra
until the collimator was properly aligned after the drain. Thus we lost
almost three wvaluable hours of recording.

With our location equipment we determined the position of the shield
plug hole and hence could relate the calculated activities found from the
spectra through the shield plug hole with those deduced from the actual sur-
vey afterward. All calculated activities were normalized by comparing the
®31b calculated from a spectrum taken through the shield plug hole and that
actually found during the survey (Sect. 7.6.2). Ve used 0.93 x 10** dis
min~' em™® as the normalizing activity for °°Nb at this location. All ac-
tivities given below are calculated at counting time and are expressed in
disintegrations per minute per square centimeter of heat exchanger tube
surface.

In view of our calibration methods, it should be noted that the ac-
tivities of gaseous fission products, which are obviously in the space in
the shell side of the heat exchanger, are expressed in equivalent disinte-
grations per minute per square centimeter of heat exchanger tube.

Krypton-88—Kkubidium-88 (Fig. 7.13). Even 3 hr after reactor shutdown,

the ®°%Kr and °°

Rb activities are still appreciable. Comparing the krypton
and rubidium activities, it appears that these nuclides are close to an
equilibrium condition,

Strontium-91. A few spectra indicate the presence of °*Sr, with its
photopeak at 1024.,3 keV. Since we found ®®Kr, one might, by the same token,
expect decay products of °*Kr. The data were too sporadic to really indi-

cate an average activity; 10 hr after reactor shutdown, a few spectra
10”

ORNL-DWG 70-9589

 

T e e S T

et

 

 

10’0 b e

N

o
©

 

NUCLIDE

2 - O S

 

|
|

(dis/min)/:m2 of heot\exchanger 1ube]
v
a-o%0-

e e e . .
88 0 898.0 kev
§ 1836.1 kev
— A : '- I S P ]
+ § . e . e e e o - 1_ o - b ]
t | . o . - .
+ + . i . y . . - ‘ - e 7,vt_,777 .
‘ ‘ . ‘ ' . e e — . { ‘ b
' ' : ! . + e i 7;, C e ,.4:_, bl ; :
| ' Ly T ? r , T ;
b = | | i | ; |
P ' . - e ——«‘——w»_ 4 e
. e | 1
, ; | e ]
ey i + ¥ 1L _ i — — -

 

 

' : |

 

 

 

,
‘
1

 

 

 

 

 

 

O 8347 kev ! ‘
— 88K 16 1529.8 kev - - ‘ _ -
4 2392.0 kev |

 

 

e e 4 e b
;

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

4 e i e s
+

 

 

 

 

 

 

108 |

0.4 0.6 08 1.0 1.2 t.4 1.6 1.8 2.0 2.2 2.4 2.6 2.8 3.0
DECAY TIME AFTER REACTOR SHUTDOWN (days)

Fig. 7.13. Activity of ?®Kr and ®®Rb after reactor shutdown in the

MSRE primary heat exchanger (November shutdown).

¢8
86

indicated a ®'Sr activity of approximately 0.15 x 10*° dis min~' em™? {not
extrapolated back to zero time).

Niobium=-97 (Fig. 7.14). This nuclide has only one strong photopeak

 

(658 keV), which makes it hard to assign much confidence to its identity,
especially since many other peaks in these spectra are located very close
to it. The fact that the observed photopeak decayed with a half-life of
108 min indicates there was interference in measuring the peak. We believe,
however, that this decay rate is sufficiently close to the accepted half-
life (78 min) to justify the peak's assignment to ®’Nb.

Extrapolation to reactor shutdown time is acceptable, since its pre-
cursor (°7Zr) was not identified in the heat exchanger. The estimated maxi-
mum activity at zero time is 0.6 x 10" dis min™' em™ 2.

Molydenum=-99 (Fig., 7.14). Judging from the plots of different spectra,

 

we have more confidence in the activities yielded by the 739.7-keV photo-
peak than in those from the 778.2-keV peak. The latter is located very
close to a large **°I and °°Nb peak; this would make the activity calcu-
lation somewhat doubtful, especially since other short-lived isotopes also
have peaks in that energy range.

The activities yielded by the 739.7-keV peak are rather consistent and
in good agreement with the results of the actual survey afterward. The

estimated maximum activity at reactor shutdown time is 0.18 x 10*? dis

-

o 1 -—

min cm T .
Ruthenium-105—Rhodium-105 (Fig. 7.15). It appears difficult to evalu-
ate exactly what the activities of these two nuclides are. One would expect

that *°°

Ru and *°?Rh behave approximately the same.

The three '°°Ru photopeaks (469 4, 676.3, and 724.2 keV) decay with a
half-life close to that of '°°’Ru. The first two peaks, however, yield a
nuclide activity almost half that from the third peak. The activity de-
duced from the first two peaks at reactor shutdown time would be 0.3 x 10%*
dis min~! cm~?; that from the 724.2-keV photopeak would be 0.46 x 10"
dis min~* cm™®. It is very well possible that our nuclear data are not
correct for this isotope.

'°5Rh activity results is even worse. Especially

The scatter of the
during the first hours after shutdown, the 319.2-keV peak indicates a high
activity; these possible aberrations might be due to photopeaks from other

nuclides, which could cause this overestimate.
NUCLIDE

(dis/min)/cm2 of heat exchonger tube]

[
S,
o

 

ORNL—DWG 70-9590

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

&
|
| —
| ' 1 : ‘ ‘ ‘ :
ol L ee, ot . |
* ‘ | \ | Vo778 2kev ; |
. | | | | °"Nb 0 658.2 kev !
ey e
% e e = T L + - —_— ;{ e R r— e
,’__;_-‘—— — ~L — + - } — - + t —————————fr———— <|+-—~ ————— —_ i —— —
I U SN f - — & . G e 3 _ —_—— TN
ek + - , - ‘ f ;
et e
0 i ‘ i : ,
| T — ] | "*
- et e w-qt-m e e ] - + _—t —_ — —o l ‘ L e ——
‘ - |
t

 

 

 

 

 

0.2 04 0.6 Q.8 1.0 1.2 1.4 1.6 1.8 2.0 2.2 2.4 2.6 2.8 30
DECAY TIME AFTER REACTOR SHUTDOWN (days)

Fig. 7.1k, Activity of ’Nb and °°Mo after reactor shutdown in the
MSRE primary heat exchanger (November shutdown).

L8
NUCLIDE [(dis/min}/cm? of heat exchanger tube]

 

 

ORNL—DWG 70—259f

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

o' — T 1 T T T | 1
' . . T ’ i _ ‘ -405 IO 3063! keViii:‘ihuj‘..:'.—_“ ]
[ | | ~ ‘ * - +— 6 319.2 kev —+
¢ z | - - 0 469.4 kev
¢ o | | | I | +{O5Ru{¢ 676.3 kev
¢+ 1 | , ; l 3 © B ¥ 724.2 kev I
_Q_ : | ‘ ¢ , :
y 8 4 i o | Q o
100 o 7 | | : N . 1
L | | r l [ oy S T __Q_f__ -
1 8 | | |o | 1 _1 T ~
5 o . 4 . i 4 + T ]
- R —
M+ e L T”, 1 L ]
ol o % S N S R
00 | | | | |
02 04 06 08 0 42 14 16 48 20 22 24 26 28 30

Fig. 7.15.

DECAY TIME AFTER REACTOR SHUTDOWN (days)

Activity of '95Rh and '°SRu after reactor shutdown in the
MSRE primary heat exchanger (November shutdown).

88
89

Based on activities deduced from later spectra, we might conclude the

105

maximum Rh activity would be in the range of 0.4 x 10! dis/cm?®.

Antimony-129—Tellurium-129—Tellurium-129m (Fig. 7.16). Since several

 

other antimony isotopes are known to be present on the metal surfaces of
the heat exchanger, there is no apparent reason why ‘2°Sb should not be
present too. The problem is that the decay scheme of this nuclide is not
known with certainty; some of its gamma rays are indicated in the litera-
ture, but their reported abundances are doubtful.

We believe we have identified '2°Sb by one of its photopeaks (1028 keV)
and by its decay half-life; but, due to lack of data on the branching ratio
for this gamma ray, we did not calculate the disintegration rate of *2°Sb.
The photon emission rate of this gamma ray at reactor shutdown time was
0.90 x 10*° photons min~* cm™2,

As reported in Section 7.5.1, we identified '?°™Te primarily by the
459.6-keV peak from the decay of *2°Te. The intensity of the 459.6-keV
(and 1084.0-keV) peak would reflect the decay of **°Sb, *?°Te, and '**"Te,

Figure 7.16 shows the '?°Te activity based on the 459.6- and 1084,0-keV

129Te

photopeaks. The absolute abundances of these peaks per decay of
(ground state) are taken as 7.7 and 0.6% respectively. In the same figure
is shown the photon emission rate of *2°Sb for its 1028-keV peak.

Tellurium-131m (Fig. 7.17). Even if '?'Sb would deposit on metal sur-

 

faces, the maximum *>'"Te activity would occur, because of the short *°'Sb
half-1life, very soon after reactor shutdown time; hence an extrapolation
to zero time for the maximum activity seems justified. The 1206.6-keV
photopeak appears the least influenced by other peaks and should be con-
sidered the most trustworthy., Our estimate of the maximum *°'"Te activity
is 0.73 x 10** dis min~' cm™?,

Todine-131 (Fig. 7.18). Based on the chemical properties of iodine,

 

there is ample reason to believe that iodine remains with the fuel salt.
This was confirmed by gamma-ray spectra taken in a later stage of the ex-
periment. This would mean that any iodine detected came from the decay of
either antimony or tellurium,

The half-lives of '®'Sb and '*'Te are so short that the buildup of

1317 from these nuclides could not be observed. The formation of iodine
¢ [photon emission rc:'re/cm2 of heat exchanger tube ]

¢ NUCLIDE [(dis/min)/cm? of heat exchanger tube]

 

ORNL-DWG 70—-9592R

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

| | - | | ‘
10" -0 ) - [‘ - *fif"f___;fl*:_ - ‘Jf“ I““fiL‘ i:j
ey o = o | j_m =
5 | el . o ' ! L —[ ‘Jr_ IL ; i
—_ Oy T _— + - e — { 4" e 4 g e - ! w-————m——+—i —A—]
|
I L - _ S
3 | b | o
2 |— _ [ o N . i -5 e q | .[ ]
l ~ o |
10'% | ¢ P : o °
) LT - B T
e T T 'fi;gfig““”ii:;m:”fif” "ff  - ~ N L T
5 __._o__", i — ! e L 129T 0 459 .6 kev 1
- ' 16 1084.0 kev oy
e b "%b ot028kev  — i .
b e
| 0 | . i
10° N | j -
T T T 8T _ )
— e 4 e VV—A—-T————Af - - ,‘ —_——
5 : | 1 . ‘
0 0.2 0.4 0.6 0.8 1.0 1.2 1.4 1.6 {.8 2.0 2.2 2.4 2.6 2.8 3.0

DECAY TIME AFTER REACTOR SHUTDOWN (days)

Fig. 7.16. Activity of '?°Te and photon emission rate of '??Sb mea-
sured after reactor shutdown in the MSRE primary heat exchanger (November
shutdown).

06
91

ORNL—DWG 70-9593

10“

- 1

‘3t © 793.8 kev
Te 4 § 852.3 kev
1206.6 kev

847.0 kev

134 $884.| kev
4~ 1072.5 kev

NUCLIDE [(dis;/min)/Cm2 of heat exchanger tube]

 

0 0.2 04 0.6 0.8 1.0 t.2 1.4 1.6 i8 2.0 2.2 24 2.6 2.8 3.0
DECAY TiME AFTER REACTOR SHUTDOWN {days)

Fig. T.17. Activity of 131mpe and !3*I after reactor shutdown in the
MSRE primary heat exchanger (November shutdown).
]

N

-
ol
=

N

(dis/min)/em?2 of heat exchanger tube
o

NUCLIDE [
3,
o

ORNL-DWG 70-9594

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

T T T T
! o
| I
! 0 364.5 kev :
L . , . R 134 _
b . e {¢ 637.0 kev ¥ ,
s r s - mfmn . ; 1
e e “ » ; e
- . } | :
; ¢ 1 g . L o ) j¢ ¢ -
N | ¢ . . 0 . R
0? \ ’ ! ? 0
¢ ! 0 o ! 3
09 ° | | 1 ; i i
0 0.2 0.4 0.6 0.8 {.0 1.2 1.4 {.6 1.8 2.0 2.2 2.4 2.6 2.8

Fig. T7.18.
mary heat exchanger (November shutdown).

DECAY TIME AFTER REACTOR SHUTDOWN (days)

Activity of 1317 after reactor shutdown in the MSRE pri-

3.0

4
93

from the decay of '®'"Te, in the case that primarily *°'Sb deposits on the
metal surfaces, presents only a few percent of the total iodine activity
and is difficult to observe. A simple extrapolation to reactor shutdown

time to obtain the maximum *3*

I activity would then mean a very small over-
estimation because of this decay of *3*'Te,

In case tellurium rather than antimony deposits, the overestimation
would be much higher, in the order of 60%, because of the slow decay of
131/MTg,

There is a fair degree of scatter in the calculated activities. The
estimated maximum activity of *®'I would be approximately 0.24 x 10'' dis
min=' cm~? and is based on the average of the 364.5- and 636.9-keV photo-
peaks. This conforms with the results of the actual survey by extrapo-
lating those back to reactor shutdown time. The evolution of the iodine
activity is considered in more detail in Chapter 8.

Tellurium-132—Iodine-132 (Fig. 7.19). For the calculation of the

'32Te activity, a simple extrapolation to reactor shutdown is justi-

maximum
fied. Even if it is the tellurium precursor, **2Sb, that plates out on the
metal surfaces, the latter decay half-life is so short that this would jus-

132

tify the assumption. The maximum Te activity is approximately 0.40 x

10'? dis min™' cm—2.

The iodine activity is expected to grow in from the tellurium decay;
that is, its maximum activity should occur about 9 to 10 hr after reactor
shutdown. The observed *32?I activities do not agree with this. Its ac-
tivity went down again after 5 hr and then after 20 hr or more finally came
into equilibrium with its precursor. The excess of **?I in the main off-
gas line at about the same time period might indeed point to the transfer
of iodine from the reactor system to the off-gas line. Because of this
temporary deficiency in the heat exchanger, the maximum **?I activity occurs

-2

some 24 hr after shutdown time and amounts to 0.34 x 10'? dis min—! em™2.

Iodine-133 (Fig. 7.20). The *3?I activity seems to decrease faster

 

than can be explained from decay only (transfer!). Taking into account

133

the decay of its precursor, the maximum I might be expected to be below

0.20 x 10'! dis min=* cm™® (before its apparent transfer).
U

M

1042

N

NUCLIDE [(dis/r'nin)%m2 of heot exchanger tube:’
o

—
—_
—

'ORNL-DWG 70-9595

 

 

 

 

Fig. T7.19.

 

 

 

 

 

 

 

 

 

 

 

DECAY TIME AFTER REACTOR SHUTDOWN (days)

MSRE primary heat exchanger (November shutdown).

‘ ;' | ;
J[ R l L : ‘, | o - — ;‘]
; ; . l ’ ; L I e .
> 3 1321¢ { o 228.2 kev f
b ; | o ____’%_ . I
| g 0 667.7 kev r ; |
| 321 1§ 954.6 kev | |
I | 4 1398.6 kev 1’
. 3; : : ’ ) S T T
| . l . § , } e . +‘ e e |
. % . . . S 4 e g
fi L o . i . o4 .l L T ]
0 o o % : i i
+§ oy | © o O $ 4 |
L T_ ,g‘¢, | “1} RS e
v | | ° © |
: l i ! ‘ |
| | ; ‘ |
| R o ; 1 | |
0 0.2 04 06 0.8 1.0 1.2 1.4 1.6 1.8 2.0 2.2 2.4 2.6 2.8 3.0

Activity of 1327e and '32I after reactor shutdown in the

%6
NUCLIDE [(dis/min)/crn2 of heat exchanger tube]

ORNL-DWG 70-9596

35%e:

© 249.5 kev _ ___

1331, 0 529.9 kev

 

 

0.2 04 0.6

Fig. T.20.

¢

0.8 1.0 1.2 1.4 1.6 1.8 2.0 2.2 2.4 2.6 2.8 3.0
DECAY TIME AFTER REACTOR SHUTDOWN (days)

Activity of 1337 and !3°Xe after reactor shutdown in the

MSRE primary heat exchanger (November shutdown).

)
96

Todine-134 (Fig., 7.17). The maximum '°*I activity might be expected

 

toc be around 0.30 to 0.60 x 10'! dis min=! cm~? if one takes into account

134

that I 1s on the metal walls only as a decay product of tellurium,

Xenon-135 (Fig, 7.20). The presence of xenon in the reactor system is

 

ample and in the first hours after reactor shutdown forms an important ac=-
tivity source. Its maximum activity might, at that time, run as high as
0.30 to 0.40 x 10'* dis min=" cm™?, The decrease in activity is obviously

due to decay and purging of the system.
7.5.3 Drain tank

Several spectra were taken from one drain tank after reactor shutdown,
These spectra were recorded through the hole in the shield plugs over the
drain tank. There was no possibility for calibration of the results or even
for proper alignment. By moving the detector in small amounts back and
forth, we tried to obtain a maximum activity reading.

All data, of course, had to be normalized to have a basis for compari-

son. We chose the 724.2-keV photopeak of ®°Zr, extrapolated back to reactor

 

shutdown time, This appeared to be the most convenient normalization point,

 

since zirconium is supposed to remain with the salt and will be at its maxi~-
mum activity very shortly after reactor shutdown time because of its short-
half-life precursors.

The spectra were taken primarily to check the evolution of the niobium
activity, However, there were some other interesting observations too.

Since we were not entirely sure about the location of the place on the
drain tank from which spectra were taken or the amount of shielding ma~-
terials (flanges, drain-tank vessel, steam drum, etc.), we will report only
on relative results of nuclides having photopeaks close to each other in
the energy range of 600 to 800 keV.

Table 7.5 indicates the evolution of the ratio of the °°Nb to °°Zr
and the ?’Nb to ®’Zr activity. The °°Zr activity is based on the average
of the photopeaks at 724.2 and 756.9 keV. The °°Nb, °’Nb, and °7Zr activi-
ties are based on the 765.8-, 658.2-, and 743.4-keV photopeaks respectively.
The °’Nb/°’Zr ratio is not of great value for analysis, since any °’Nb iden-

tified is formed in the drain tank in our case. (Any °’Nb in the salt that
97

Table 7.5. Ratio of ®°Nb to °°Zr and ?'Nb to °7Zr
activity in relation to decay time after
reactor shutdown in November

Activities at counting time were used; no
extrapolation to reactor shutdown time

 

Decay time after
reactor shutdown

 

(days) ®SNb/®%zZr *’Nb/®7Zr
0.872 0.78 0.97
0.971 0.79 0.99
1.982 0.61 0.97
2.797 0.62 1.02
2.873 0.65
3.920 0.97
4,900 1.55
7.780 2.07
8.893 1.98
11.904 2.30
11.963 2.30
15.885 2.65
16.907 2.89

 
98

drained from the reactor would have decayed before these spectra were
taken.) It is, however, a good check on the credibility of the °°Nb/’°Zr
data.

The °°Nb/°®Zr ratios do not appear to indicate the real activity ratio

°5Nb in the reactor

in the salt itself. If we consider all of the °°Zr and
system (regardless of location), the maximum value for this activity ratio
is 1,0, immediately after shutdown. In view of the ?°Nb deposits in the
primary loop, the activity ratio in the drain tank would be expected to be
appreciably smaller, unless the detection efficiency for ?3Nb were somehow
enhanced. One explanation might be that the °°Nb concentration near the
salt surface is much higher than the average in the salt.

The first spectra taken after reactor shutdown (after about 0.800 day
of decay) revealed the strongest activities to be due to °°Kr(Rb) and '*°Xe.
Besides the nuclides expected to be with the salt, such as yttrium, zir-
conium, cesium, cerium, and barium-lanthanum, there were also some unex-
pected observations. For example, we did not detect any molybdenum, anti-
mony, or tellurium; the amount of ruthenium present (both *93Ru and *°°Ru)
was very small, Molybdenum-99 has several photopeaks by which it could be
identified; none of these peaks was present. This would indicate that the
®°Mo activity has to be at least a factor of 100 smaller than the °°Zr ac-
tivity. The same holds true for all the tellurium isotopes; neither the
relatively short-lived ***"Te nor the longer-lived '**"Te could be identi-
fied. A further confirmation of the absence of tellurium was the fact that
we could not identify **?I. Since '°*I has a 2.3-hr half-life, this would

132
T

then indicate the absence of the longer half-lived e. LEspecially the

absence of *°?1 is an important proof because this nuclide can be identi-

fied more easily than any other nuclide in a gamma-ray spectrum. (Ve did

131 133
I I

identify in the drain tank all longer half-life iodines such as

and '%°1.)

3 ’

As in the case of molybdenum, the activity of antimony, tellurium, or
1327 had to be at least a factor of 100 smaller than the ®°Zr activity to
go unidentified in these spectra. Table 7.6 shows the major nuclides identi-
fied at various times after reactor shutdown. There were several peaks in

232

the spectra that could not be identified, probably due to the U chain

decay products.
99

Table 7.6. Major nuclides identified in fuel salt
in the drain after November shutdown

 

®8Kr (Rb) 13371

ler 1351

9szr 135Xe
?5Nb 137¢Cs
®77r t4%Ba(La)
*7Nb 41Ce
'93Ru (very small activity) 143¢Ce
1°Ru(Rh) (very small activity) 14%Ce(Pr)
1311

 

7.6 Group G Spectra

 

These spectra were recorded during the November shutdown period on the
heat exchanger, the main off-gas line, and the fuel lines. The portable
maintenance shield was used for the survey.

The results presented in this section are those we were really looking
for. The data recorded in July were useful for the analysis of the long-
half-life nuclides, and, above all, it served as a very profitable shakedown
of the equipment. It should be borne in mind that all results are extrapo-
lated to reactor shutdown time by simple exponential techniques. The rea-
son for this extrapolation was to compare results from spectra taken at
different times. This extrapolation might yield, however, a maximum ac-
tivity value for some isotopes that is too high because of the decay of a
precursor nuclide. For example, the extrapolated value of *>'I might be
too high by a few percent because of the decay of *®Te. We will discuss
this in detail in Chapter 8.

Although the analyzer broke down about 24 hr before this planned re-

actor shutdown and drain, everything was made operational in time. The
100

spectra recorded through the shield plug holes in the first two days after
the reactor shutdown while the portable maintenance shield (PMS) was being
set up are reported in Section 7.5. The reactor shutdown and drain occurred
on November 2, 1969, at 1441 hr.

Never before was the PMS installed so soon after a reactor drain,
However, this did pose some problems: the radiation coming from the open
reactor cell was considerable, and extra precautions had to be taken in
installing the PMS at the location where the two lower shield plugs had
been removed. However, no one received a radiation dose above the per-
mitted level. During every day shift, the PMS was installed over a dif-
ferent part of the heat exchanger (or from time to time over the jumper
line of the main off-gas line); spectra were then taken during the evening
and night shifts. As reported earlier, we could scan with one PMS setup
approximately 35 in. along the longitudinal axis of the heat exchanger or
the whole length of off-gas jumper line.

Spectra were taken at intervals of 1 to 2 in. along the heat exchanger
axis and 1 in. along the length of the jumper line. Based on our previous
experience, we judged it worthwhile to increase the real counting time to
800 sec per spectrum. This proved to be advantageous for the counting
statistics and allowed us to identify smaller peaks better. Once a 35-in.
scan was finished, we set the detector over a specific spot and recorded
several long counts of 10,000 sec real counting time for an even better
analysis of the nuclides present. These long counts were continued until
the day shift was ready to move the PMS to the next position.

The entire survey along the length of the heat exchanger was done
twice. Several areas were scanned a third time; the off-gas jumper line
was scanned three times. No transverse scans were made over the heat ex-
changer. It took approximately three weeks to finish the complete survey.

It is worthwhile to note that spectra taken of the same spot but at dif-
ferent times after reactor shutdown did, in general, yield values that were
very similar; this gives some confidence in the results obtained.

There were two series of spectra that gave higher nuclide activity
results. These were taken with the 1/16-in. hole collimator insert at the
very beginning of the survey. We estimate that these slightly higher re-

sults were due to two problems:
101

1. The 1/16~in. hole in the collimator was far from straight and
proved to be difficult to calibrate in relation to the 1/8-in. hole col-
limator; its calibration value, hence the detector system efficiency curve
for this case, might have been somewhat in error.

2., The cacophony of nuclide photopeaks shortly after reactor shutdown
was such that it was obvious that photopeaks from several shorter- and
longer-half-life nuclides fell on top of each other, thus causing the com-
puter analysis program to overestimate some nuclide activities. As much
care as possible was taken to ensure that the activity calculation con-
sidered only those photopeaks that had little or no interference from other
peaks. This was, however, not always possible for the spectra taken in the

early stages of the survey.

7.6.1 Heat exchanger
About 241 spectra were taken along the longitudinal axis of the heat

exchanger. With a few exceptions, all could be analyzed successfully. The
spectra taken over the heater connector boxes were discarded; the extra
amount of shielding material that was obviously present attenuated the gam-
ma rays too much to allow intelligible conclusions from those spectra. The
results are presented per nuclide and in disintegrations per minute per
square centimeter of heat exchanger tube. Data in which we have less con-
fidence because of the above reasons are indicated in the figures as open
circles.

Niobium~95 (Fig. 7.21). As was also found from the spectra taken in
July 1969, there appears to be an ipcrease of activity near the baffle
plates. There was a good agreement between the data taken in July 1969 and
those in November 1969. All results were based on the single °°Nb photo-
peak of 765.8 keV. The range of activity is 0.090 to 0.27 x 10'? dis
min~' cm”?

Molybdenum-99 (Fig. 7.22). Because of the relatively short half-life

 

of °°Mo, a large portion had decayed before we could survey the whole heat
exchanger. There is, however, ample reason to believe that molybdenum

would deposit all over the heat exchanger.
dis/min +cm2 OF HEAT

EXCHANGER TUBE

ORNL-DWG 70-7366R

 

 

T 1 ! ] T
| | | | e e
0.30 Ef2 | Po |
! 1 P ° ‘... O.L
0.20 E12 - .. o . e { . o e ..
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HTR. PLUG HTR. PLUG ‘ SPACER HTR. PLUG | SPACE!R ‘
R | - ‘ -

  

 

 

POSITION HOLE IN ‘
SHIELD PLUGS

Fig. 7.21. Activity of 95Nb at reactor shutdown on November 2, 1969,
in the MSRE heat exchanger.

¢0T
ORNL -DWG 70-9574

 

 

 

< . ! T | 1 r
w w 0.50 B2 ¢ i | e | i [
T M | .
= | 1 I ol ° o '
&' 0.40 EfR i | x P Lo o |
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NE & ; ; o c f ' )
030.30512' | | ¢ .| el e e
= | . 1 i
g % 0.20 Ef2 - i : S i . i ’ N
E % . ' t o - . » » l i
fEomee | S Tt |
S | ‘ ] i o ;
| ‘fi | | ?fi |
HTR. PLUG | SPACER HTR. PLUG t ER |
1 .

HTR. PLUG

    

 

 

 

  

|
FUEL SALT FLOW —
|

o POSITION HOLE IN I
SHIELD PLUGS

 
 

Fig. 7.22. Activity of ?’Mo at reactor shutdown on November 2, 1969,

in the MSRE heat exchanger.

€01
104

Although there is considerable scatter of the results (due to poor
counting statistics). it does appear that there is again a concentration
of activity near the baffle plates.

Results were primarily based on the average activity yielded by the
739.7- and 778.2-keV photopeaks. The rather steep drop in the detector
system efficiency below 250 keV made the 140.5-keV photopeak less desirable
for analysis (see Chap. 4). The range of activity is 0.15 to 0.40 x 10?2
dis min~* cm~?.

Ruthenium-103 (Fig. 7.23). Essentially the same distribution of the

 

ruthenium deposition occurs as with the previous nuclides. The increase in
activity near the heat exchanger tube sheet or near the baffle plates is
somewhat more pronounced than with °°Nb or °°*Mo.

The results were primarily based on the 496.9-keV peak and to a small

extent on the 610.2-keV peak (for confirmation only). The range of ac-

5 ~
-

tivity is 0.10 to 0.70 x 10 dis min™* em™*

RutheniumRhodium-106 (Fig. 7.24). 1In addition to the increase of
activity near the baffle plates, there is a good deal of scatter in the
data; this might be expected because of the longer half-life of '°°Ru and
hence its smaller disintegration rate. For the same reason, one should not
expect this nuclide to be even close to the saturated deposition concen-
tration.

In general, the resultis were based on the average of the 511.8- and
621.8-keV photopeaks. The range of activity is 0.050 to 0.14 x 10*! dis
min~* cm™?

Antimony-125 (Fig. 7.25)  Antimony-125 has a long half=life, a small

 

fission yield, and photopeaks that could be easily confused with those of
other nuclides, all circumstances which are not particularly advantageous
for a good gamma-spectrometric analysis.

d 125

Contrary to the molybdenum results, we detecte Sb mostly near the

end of the survey when most of the shorter-half-life nuclides had disappeared.

i258
Sb was

Figure 7.25 presents the results for those locations where
identified; it is, however, believed that this nuclide was distributed in
comparable amounts all over the heat exchanger. The scatter in the data

is largely due to the bad counting statistics.,
dis /min - cm2 OF HEAT

ORNL-DWG 70-7364A

 

 

 

 

 

 

 

 

 

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FUEL SALT FLOW —»
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POSITION HOLE IN ‘
SHIELD PLUGS

Fig. T7.23. Activity of 1038y at reactor shutdown on November 2, 1969,
in the MSRE heat exchanger.

0T
ORNL- DWG 70- 95?5

 

 

 

 

 

'<_I ) 1 T T ST T 1
W& 0.20 EM ! J | | e 1 | |
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. = ] -
FUEL SALT FLOW —~ |

- - POSITION HOLE 1NT
SHIELD PLUGS

Fig. 7.24. Activity of '9Ru-Rh at reactor shutdown on November 2,
1969, in the MSRE heat exchanger.

  

 

 

 

901
 

 

 

 

 

 

E"[ ORNL-DWG 70-9576

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Fig. T7.25.

 

i
FUEL SALT FLOW —=

POSITION HOLEIFfiJ
SHIELD PLUGS

125

Activity of ob at reactor shutdown on November 2,

1969, in the MSRE heat exchanger.

    
 
  

L0T
108

Results were based mostly on the 427.9-keV photopeak and, to a lesser
degree, on the 600.8-keV peak. The range of activity is 0.080 to 0.30 x
10*° dis min~' cm™?.

Antimony-126. This nuclide proved to be difficult to identify. Its
fission yield is relatively small, and its two major photopeaks (666.2 and
695.1 keV) fall together with those of '°?I and ***"'Te. The next important
peaks were also close to photopeaks of other nuclides.

Although we identified '*°Sb several times, the scatter in the com-
puted activity is such that the accuracy is expected to be low. A nuclide
distribution pattern along the heat exchanger could not be established.

The range of activity is 0.023 to 0.14 x 10*° dis min~' cm™?; average ac-

?; and standard deviation of the average

tivity is 0.89 x 10° dis min~' cm™
activity is 0.11 x 10°,

Antimony~127 (Fig. 7.26). The increase in activity near the baffle

 

plates, although apparent, is by far not as strong as with °5Nb or *°°Ru.
Although '?7Sb could be detected in almost every spectrum, we had to rely
almost entirely on one photopeak (684.9 keV) for the activity analysis;
naturally this causes some scatter in the data. The range of activity is
0.20 to 0.55 x 10'* dis min™" cm™?,

Tellurium-129n (Fig. 7.27). The amount of *?*Te deposited on the

 

metal surfaces might be of crucial importance in explaining the deposition
mechanism of several of the decay chains. There are, however, several
problems that put a strain on an accurate analysis of the *?*°Te (see
Chap. 4.3).

The scatter of the results is considerable, although most within a
certain range. There does not appear to be a strong increase of activity
near the baffle plates; one might observe, however, a slight increase of
activity near the tube sheet of the heat exchanger., The range of activity
is 0.10 to 0.30 x 10*' dis min™' ecm™*%.

Iodine-131 (Fig. 7.28). The increase of activity near the baffle

 

plates is definitely there, although much less pronounced than with the

103
R

metals °*Nb and u. The photopeaks were easily identifiable, and the

confidence in the data is good. For the analyses we used the 364.5- and
dis/min-cm? OF HEAT

S 3 IR I

ORNL-DWG 70-9577

 

 

 

 

 

 

 
 

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FUEL SALT FLOW —»

 

 

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T
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POSITION HOLE IN’I
SHIELD PLUGS

Fig. 7.26. Activity of '27Sb at reactor shutdown on November 2,

1969, in the MSRE heat exchanger.

60T
 

 

 

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\w w S SO i 1. _
2 ‘ | | |
o E | ! I |
! | |
l : | | |
HTR. PLUG HTR. PLUG SPACE!T& HTR. PLUG | SPACER
LT ¢ | I

    

 

r

1 L.

 

 

FUEL SALT

POSITION HOLE H‘LT
SHIELD PLUGS

Fig. 7.27. Activity of 129Mpa gt reactor shutdown on November 2,
1969, in the MSRE heat exchanger.

011
dis/min-cm® OF HEAT
EXCHANGER TUBE

ORNL-DWG 70-9579

 

 

 

 

 

 

- v T T —' ! T
060 Ef | o i |
{ 1 +. ’ } : . P
040 EN ; P el s | “ e e
} . ! s " *% U‘%H 25 ° O. :_ : *..'b‘ '...' | .o .
0.20 EM1 L..on R * Cetst .t . T %
o 3 . ' ' ° |
0 f T
| | |
f : ! |
HTR. PLUG HTR. PLUG ‘ SPACER HTR. PLUG SPACER

      

fryrrrETT ]

 

 

 

FUEL SALT F;LOW -

o : : 1
. ; !
/ -

|

.

! ’ !

- ’ i !

L Do

1 . ‘

L

[ - Lo e e - - S e — - . — e - -

POSITION HOLE IN_I
SHIELD PLUGS

 

 

Fig. 7.28. Activity of '3*'T at reactor shutdown on November 2, 1969,
in the MSRE heat exchanger.

TITT
112

636.9~keV photopeaks; the 284.3-keV photopeak was used less because this is
in the energy range where the detector efficiency begins to drop consider-
ably.

The '®*"Te had already decayed to a large extent when we started the
actual survey with the PMS. As with antimony and tellurium, there appears

to be some increase of activity near the tube sheet of the heat exchanger.

11 -2
.

The range of activity is 0.15 to 0.40 x 10"' dis min™' cm

Tellurium—TIodine-132 (Fig., 7.29). Because of the short half-life

 

of '®21, one observes basically the decay of '*?Te. Iodine-132 conveniently
decays with a multitude of well-known gamma rays, many of which are easy to
identify. For our tellurium and iodine analysis, we mostly used the 667.7-,
772.6-~, 954.5-, and 1398.6~keV photopeaks of '®2I to compute an average
disintegration rate., Since the most abundant *32Te gamma ray (228.2 keV)
was so low in the observed energy range, we did not rely much on this peak.
The nuclide distribution along the heat exchanger axis was about the

131
I

same as for ; again the nuclide activity increase near the baffle plates

was not as strong as with °°Nb and *°?Ru. The activity range is 0.30 to
0.75 x 10*? dis min~' em™?

Iodine-133, The '°°I was detected but had already decayed appreciably;
the results have been reported together with the gamma-~ray spectra taken
through the shield plug holes immediately after reactor shutdown (Chap. 7.5).

Barium—Lanthanum-140 (Fig. 7.30). Some of the spectra revealed the
presence of *“°Ba-La. Its most abundant photopeak at 1596.6 keV is very

easy to distinguish and was generally relied upon for the activity analysis.

The range of activity was 0.050 to 0.20 x 10'° dis min™* em™2.

7.6.2 Main reactor off-gas line

The only part of the reactor main off-gas line that is clearly visible
from above without obstruction and hence available for a gamma-spectrometric
survey is the jumper line. This jumper line is a corrugated, l-in.-ID
flexible tube connecting the fuel pump purge-gas exhaust at the pump bowl
with the 4-in. pipe section of the off-gas line. This jumper line is ap-
proximately 2 ft long.

Three surveys were made in the November scanning period, each com-

prising about 20 spectra taken approximately 1 in. apart along the off-gas
 

 

 

 

 

~ ORNL-DWG 70-7365R
L<lIJ ! T o T T T T
T 5 080 Ef2 | | o | ° | e
= | ! . ‘e ° e e
o @ l o| . o ° «of ."l . o .
o W 0.60 Ei2 | | wV.e o .. o ... .I. '.. 1 o ’ . ! :
g% .: .J% . ['o ; . 1; - oo 0. .o r :. S o|+. l :
g g 00 B » o™ T el | et |3 o L
EX % | 1 | l 6] B
o W 4 — ! - i ! } !
S | | | | |
i | | | | |
HTR. PLUG I—‘iTR. PLUG : SPACER HTR. PLUG : SPACElR |

    

 

 

 

|
FUEL SALT FLOW —»

) L]

 

 

POSITION HOLE IN-_T
SHIELD PLUGS

Fig. 7.29. Activity of 1327e_1 at reactor shutdown on November 2,

1969, in the MSRE heat exchanger.

ETT
dis /min-cm2 OF HEAT

EXCHANGER TUBE

ORNL-DWG 70-9580

 

 

 

 

 

 

 

 

 

  

T I l | l l
¢.30 E10 ! i 1 ey | f'
; I I i e ]
| l ! |
0.20 E10 | : i s 4 : i
} { I:. | - l ——— |
0.40 E10 » ’ . : 1 . . .i .Qo: g e - _.{__4
e i } P R . g R I . S
| } ' ' .
0 | ! T I + il
| | | | | |
| i I | I |
! | I
HTR. PLUG HTR. PLUG i SPACER HTR. PLUG | SPACE|R :
! e o | | |

 

 

 

 

 

 

 

 

POSITION HOLE IM
SHIELD PLUGS

Fig. T7.30. Activity of 1%0Ba_La at reactor shutdown on November 2,
1969, in the MSRE heat exchanger.

711
115

line. Since there was no noticeable decrease in activity in the downstream
direction, the results of the recorded spectra from each survey have been
combined. With the exception of possibly *2°Sb, *2*7Te, *?7Cs, and '“°Ba-La,
the scatter of the calculated activity results was quite small.

There was a difference in the results of the three different surveys.
Because the surveys were taken approximately one week after each other, the
decline in the general activity level necessitated different shielding con-
figurations between the detector and the jumper line. For example, the
first survey was done using the 1/16-in. collimator insert with no shielding
material; the second and third survey were executed with a 1/8-in. colli-
mator insert and 1 in. aluminum shielding. We estimate that the results
of the first survey are about 357 too high because of discrepancies in the
1/16-in, collimator calibration (we observed this also with the heat ex-
changer data). Tables 7.7 to 7.9 represent the average activities calcu-
lated from each of the three surveys. We believe that Table 7.9 represents
the best estimate of the activities detected in the jumper line.

During the first survey, '°°Ru~Rh and '°°I were detected several times.
Their calculated activities have been incorporated in the results given in
Section 7.5.

We were unable to detect positively any component suggesting the pres-
ence of salt in the jumper line. In some spectra, a visual inspection of a
spectrum plot revealed very, very weak peaks that might be coming from ®3zr.
These peaks were so small that the computer analysis in most cases did not
detect them., In a practical sense this implied that the *°Zr/®°°Nb activity

ratio was much less than 0.0005.
7.6.3 Fuel pump bowl and fuel lines

The spectra recorded from gamma rays coming from the main fuel lines
(lines 101 and 102) as well as from the fuel pump are useful only as quali-
tative indications. Our calibration work certainly did not cover the com-
plex geometry of the fuel pump bowl nor the fuel lines. From experience
we know that the efficiency curves for these geometries might be expected
to be relatively flat in the area of 500 to 1000 keV, so we did compute
relative activities for the identified isotopes. Those relative activities
are only indicative figures, and restraint should be used in employing

these wvalues.
116

Table 7.7. Average activity measured along the jumper
line of the MSRE main off-gas line of nuclides
detected during the first survey after the
November shutdown

 

 

Nuclide Average activitya Standard deviation
®SNb 0.12 x 10** 0.24 x 10*?

* Mo 0.48 x 10** 1.11 x 10*2
193Ru 0.25 x 10**® 0.65 x 10*?
10€Ru~Rh 0.43 x 10*?3 3.5 x 10**
123gh 0.16 x 10*? Not sufficient data
t285h 0.86 x 10'* Not sufficient data
1275 0.79 x 102 4.7 x 10*°
129MT¢ 0.22 x 10?3 5.0 x 10*?
1311 0.82 x 10*° 2,9 x 10*?
1327e.1 0.83 x 10'® 1.7 x 10*?
*37Cs 0.77 x 10*? 7.4 x 10*°
14%Ra-La 0.40 x 10*° 1.7 x 10*?

 

aAverage activity is given in disintegrations per minute
per linear inch of off-gas line (jumper line).
Table 7.8.

117

Average activity measured along the jumper
line of the MSRE main off-gas line of nuclides

detected during the second survey after the
November shutdown

 

 

Nuclide Average activitya Standard deviation
®SNb 0.84 x 10'? 1.6 x 10**
?%Mo 0.33 x 10** 0.91 x 10*?
193Ru 0.18 x 10** 0.26 x 10'?
19¢Ru-Rh 0.24 x 10'3 0.61 x 10**
1253p 0.73 x 10! 5.3 x 10°
12835 0.47 x 10*? 2.5 x 10°
1273 0.47 x 10*? 1.9 x 10*°
129Me 0.16 x 10*? 0.42 x 10*?!
13171 0.51 x 10*? 0.93 x 10**
1227ew1 0.58 x 10*? 1.1 x 10**
137¢s 0.57 x 10'? 7.5 x 10*°
14°Ba~La 0.21 x 10*3 0.86 x 10**

 

aAverage activity
linear inch of off-gas

in disintegrations per minute per

line (jumper line).
118

Table 7.9. Average activity measured along the jumper
line of the MSRE main off-gas line of nuclides
detected during the third survey after the

November shutdown

Best estimate of activities along jumper line

 

 

Nuclide Average activitya Standard deviation
#5Nb 0.90 x 10*® 1.5 x 10*?
®*Mo 0.35 x 10** 1.2 x 102
1°3Ru 0.20 x 10** 0.25 x 10*?
106Ru-Rh 0.27 x 10*® 0.80 x 10*?
125gp 0.74 x 10*?! 6.0 x 10°
1285b 0.56 x 10** 4,9 x 10°
1273b 0.67 x 10*? 4.9 x 10*°
129MTa 0.18 x 10'3 0.46 x 10'*
1317 0.58 x 10'3 0.98 x 10*?
1327e-1 0.65 x 103 1.4 x 10'*
'37Cs 0.57 x 10*® 8.2 x 10*°
14%Ba-La 0.22 x 10*° 1.0 x 10**

 

aAverage activity in disintegrations per minute per
linear inch of off-gas line (jumper line),
119

Fuel Pump Bowl. Besides the isotopes identified throughout the sys-
tem, we identified ®°Zr and other salt components. Although not surprising,
this would indicate the presence of small amounts of fuel salt, apparently
left over from the drain. Table 7.10 shows the isotopes identified to-
gether with their relative activities in relation to the °°Nb activity.

101 Fuel Line Connecting the Fuel Pump with the Heat Exchanger. Besides
the usual deposited fission products, again there was evidence of small re-
mains of fuel salt. Table 7.11 shows the identified isotopes as well as
their activity in relation to the ®°Nb activity.

102 Fuel Line Connecting the Heat Exchanger with the Reactor. The
usual nuclides were detected but no evidence of fuel salt could be posi-
tively identified. Table 7.12 shows the identified isotopes as well as

their relative activity in relation to the ®°Nb activity.

7.7 Group H Spectra

From November 26 to December 12, 1969, the reactor was operated at
full power for approximately 16 days. The final shutdown of the MSRE oc-
curred on December 12, 1969, at 1046 hr. The spectra in this group were
taken from the heat exchanger shortly after the final power shutdown through
the shield plug hole.

During this period of operation, three beryllium exposures were made
which produced a total of 4.4 gram equivalents of reduction in the fuel
salt. Since it is believed that this has a large influence on the behavior
of the noble metals, in particular niobium, it was difficult to estimate
how much niobium was left from the previous runs, how much of the niobium
formed during this run deposited, and how much had gone from the metal walls
back into the salt.

The aiming of the collimator with the detector through the shield plug
hole was difficult. During the November shutdown we were able to normalize
all the results of the spectra taken through the hole (Sect. 7.5.2) to the
®5Nb activity found at the same spot during the later survey. However,
this was not possible this time because we did not do a survey afterward.
All results were normalized to a °°Nb activity of 0.465 x 10'* dis min~*

cm~?, this value being an arbitrary number and based on the average of the
120

Table 7.10. Longer-half-life nuclides identified in the
MSRE fuel pump bowl together with their activity
relative to the ®°Nb activity (November shutdown)

 

 

Nuclide Relative activitya
950, Identified only
95011 1

®2Mo 5

103, 1
105Ru~Rh Identified only
106p.,~-Rh 0.1

1256y Identified only
12661 Identified only
1275y 0.2

1297 o 0.2
1317, 0.7

1317 Identified only
132qa71 3

1331 0.3

1370g Identified only
140p. 1 Identified only

 

a . . . . e .

No numerical value was given when the identified nuclide
was not present in most spectra or if its major photopeak did
not fall in the mentioned energy range.,
121

Table 7.11. Longer-half-life nuclides identified in the
main fuel line (10l) together with their activity
relative to the °°Nb activity (November shutdown)

 

 

Nuclide Relative activitya
954 . Identified only
95N1 1

995 4

103p. 0.8
106p.._Rh 0.08
1256y 0.01
126y, Identified only
127 0.2
129Mqg 0.2

1317 Identified only
132101 3

137 Identified only
14%pa-La Identified only
1410, Identified only

 

a
No numerical value was given when the identified nuclide
was not present in most spectra or if its major photopeak would
not fall in the mentioned energy range.
122

Table 7.12, Longer-half-life nuclides identified in the
main fuel line (102) together with their activity
relative to the *°Nb activity (November shutdown)

 

 

Nuclide | Relative activitya
®°Nb 1

**Mo 6

193Ru 1
'°®Ru-Rh 0.09
t25gb 0.02
t2€sh 0.02
*273b 0.6
129M7e 0.4

31 Identified only
122Te-1 7

137¢Cs Identified only
14%Ba-La Identified only

 

a . . . s .

No numerical value was given when the identified nuclide
was not present in most spectra or if its major photopeak would
not fall in the mentioned energy range.
123

measured ®°Nb activities through this recording period. The actual value
might be somewhat different.

Another problem that complicated the analysis of the data was that the
final reactor shutdown was not done in the same way as the November 2 shut-
down. The reactor power was gradually decreased, starting about at 1018 hr;
28 min later the reactor scram occurred, and the drain of the fuel started
at 1105 hr. Under these conditions a proper evaluation of the activities
of the different nuclides is extremely difficult for the shorter-half-life
isotopes.

Spectra were taken for a period of approximately three days over the
heat exchanger only. Again, the activities are reported as measured at the

time of counting.

7.7.1 Heat exchanger

Because of the mentioned uncertainties, it does not seem reasonable
to analyze these data extensively; they might indicate qualitative results
only. The activity results of the most abundant activities are presented

in Figs. 7.31 to 7.40, All figures are normalized to the arbitrary °°Nb

 

 

activity of 0.465 x 10'! dis min™' g@fz. We will only make some general

 

 

remarks concerning the nuclide activity results presented in the figures,
since the absolute values are uncertain,

In view of our calibration methods, it should be observed that the
activities of gaseous fission products, which are obviously filling the
space in the shell side of the heat exchanger, are expressed in equivalent
disintegrations per minute per square centimeter of heat exchanger tube.

Figure 7.31. Based on only one photopeak, we observed a maximum ®7Kr
activity of 0.4 x 10*! dis min™' cm~2.

Figure 7.32, Rubidium-88 apparently was formed by the decay of 88Ky
after the reactor shutdown, since its activity increases to a maximum before
it comes into equilibrium with its precursor. The rate of activity decrease
of ®®Kr might be influenced by fuel pump starts at 0.03 and 0.3 day after
reactor shutdown. Concerning the ®°Mo activity, we have more confidence in

its 739.7-keV photopeak; this in accordance with previous observations.
NUCLIDE {(dis/min)/cm? of heat exchanger tube]

1

ORNL—DWG 70— 9597

 

 

 

 

 

5
I
|
§
!

t e
|

I i L

 

 

 

 

 

 

 

 

 

 

 

5 b e e s e R e e - .

8?Kr e 25‘57,0 kev

 

 

o —

 

 

 

 

e -#__._“ e e ]

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

0 0.2 04 0.6 0.8 1.0 1.2 1.4 1.6 1.8 2.0
DECAY TIME AFTER REACTOR SHUTDOWN (days)

2.2 2.4 2.6 2.8 3.0

Fig. T.31. Activity of 87Ky after final reactor shutdown in the
MSRE primary heat exchanger (through the shield plug hole).

weT
NUCLIDE l:(dis/min)/cm2 of heat exchanger iube]

ORNL-DWG 70-9598R

-
-

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

10 . L T _H—I—-——-—‘M
1 :
R oo e |
- T ] ]
5 - | -
A. A
S 8 a4 & oot os
‘ ‘M‘¢ A a N A A s ol
2 . e okl ,_# T
f | 4 A
10'© ; j 4 A Aob 4 ‘ ]
o 834.7 kev 1T L. i N ]
88, 14 1529.8 kev f ' B
' | . —
5 ‘ - 2392.0 kev W A 5 ks ‘
o ‘ | 88 o 898.0 kev
TO , i Rb {¢ 1836.1 kev
\
' g9 A 7397 kev | ‘
2t 3 }J“—M Mo {é- 778.2 kev + g :
o i ‘ | ‘ i
109 i ¢ l ‘ 1‘ . ] i ’ [ ‘
0 0.2 0.4 0.6 0.8 1.0 1.2 1.4 1.6 1.8 2.0 2.2 24 2.6 2.8 3.0

DECAY TIME AFTER REACTOR SHUTDOWN (days)

Fig. 7.32. Activity of %%Kr, ®%Rb, and °Mo after final reactor
shutdown in the MSRE primary heat exchanger (through the shield plug
hole). '

GC1
126

We observed the following maximum activities (dis min™! em™2): %%Kr,
0.42 x 10**; ®®Rb, 0.13 x 10*'; and ®®Mo, 0.41 x 10**.

Figure 7.33. 1Identified by its decay half-life as well as its only
photopeak at 658.2 keV, we feel reasonably sure to have identified °7Nb.
Its activity, in relation to the °°Nb activity, is quite large: maximum
?’Nb activity, 0.11 x 10'? dis min™ ! em™2.

Figure 7.34. Ruthenium-105 and '°5Rh were definitely detected; their
ratio of activities appeared to be different from the one found in the No-
vember shutdown (see Sect. 8.3.2).

1 —_—2
cm ) were observed:

The following maximum activities (dis min~
*°5pu, 0.30 x 10**, and '°°Rh, 0.12 x 10*%.
Figure 7.35. The observed activity of '?°Te appeared again to come

129
Sb appears

from the decay of **°Sb, *?°Te, and **®"Te; the presence of
to be confirmed by the photon emission rate of its 1028-keV photopeak.

The observed maximum activity of **°Te, based on the 459.6-keV photo-
peak, is 0.69 x 10'' dis min~' cm~?

Figure 7.36. The presence of '®'Te could not be identified with cer-
tainty, since only a few spectra showed photopeaks which might be assigned
to that nuclide; '*'"Te was definitely identified. A close look at the ac-
tivity shortly after shutdown at least does not contradict the idea that
part of the **Y"Te is formed after reactor shutdown by decay of '®!Sb; at
least two of its photopeaks (852.3 and 1206.6 keV) seem to indicate a maxi-
mum activity of 0.4 x 10'' dis min~! cm™® about 1 1/2 hr after shutdown.

Figure 7.37. We have more confidence, especially shortly after reac-

tor shutdown, in the results of the 364.5-keV *°'I photopeak. As expected,

132 13

I is formed by the decay of 2Te after the reactor drain; its maximum
activity would occur several hours after shutdown.

We observed the following maximum activities (dis min=! em—2): *3'I,
0.14 x 10**; *3®*2Te, 0.20 x 10%*%*; and *3°1, 0.16 x 10'%,

d '*°I could be positively

Figures 7.38 and 7.39. Tellurium-133 an
identified, although there is some discrepancy of the '°?1 decay half-life.
Iodine-134 typically grows in from the decay of '®“Te after reactor shut-
down; its maximum activity occurs, as expected, after reactor shutdown:
0.32 x 10! dis min™* cm™*. Other observed maximum activities were (dis

min~* em™2%): !®*Te, 0.33 x 10'*, and '*%1, 0.12 x 10'*.
NUCLIDE {(dis/min)/cm2 of heat exchanger fube]

ORNL-DWG 70—-9599

 

|
0 0.2 0.4 0.6 0.8 1.0 1.2 1.4 1.6 1.8 2.0 2.2 2.4 2.6 2.8 3.0
DECAY TIME AFTER REACTOR SHUTDOWN (days)

 

10

Fig. 7.33. Activity of *’Nb after final reactor shutdown in the MSRE
primary heat exchanger (through the shield plug hole).

LTT
NUCLIDE [(dis/min)/cm? of heat exchanger tube]

10

10

 

ORNL-DWG 70—-9600
i T ) | ,] T “*—T_fT”L*jnw f ) 3

? ) : C , T 0% 03192 key o

j ’ ’ ' ' ' {o 469.4 kev 7
t * + + . + -- I 105 . —
Ru :

 

 

 

$ 676.3 kev
4 724.2 kev "”“”T“" o

 

 

 

 

 

 

 

 

 

 

 

 

 

e L
I | . | l . © °. ; .o
é { ! 1 ‘ o
| l | | ; | | L

 

0.2 0.4 0.6 0.8 1.0 1.2 1.4 1.6 1.8 2.0 2.2 2.4 2.6 2.8 3.0
DECAY TIME AFTER REACTOR SHUTDOWN (days)

Fig. 7.34%. Activity of '%°Rh and '°°Ru after final reactor shutdown
in the MSRE primary heat exchanger (through the shield plug hole).

871
¢ [photon emission rate /cm? of heat exchanger tube ]

¢o NUCLIDE [(dis/min)/cm2 of heat exchanger tube]

ORNL-DWG 70-9601R

 

.‘011
5 $1084.0 kev
1299 4 1028.0 kev
2
1010
5
2
i
i
109
0 0.2 0.4 06 0.8 1.0 1.2 1.4 16 1.8 20 2.2 2.4 2.6 2.8 3.0

DECAY TIME AFTER REACTOR SHUTDOWN {days)

Fig. T7.35. Activity of 1297¢ gng photon emission rate of '2°Sb after
final reactor shutdown in the MSRE primary heat exchanger (through the
shield plug hole).

6CT
NUCLIDE [(dis/min}/cm? of heat exchanger fube]

 

 

ORNL-DWG 70 — 9602

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

10" | t . 1 T T e [
] | T —
@ ; ! \ ' i [ 1 + r- - ——
5 oo o r \ : l | l { L 4 I _ - |
T ' f i S - —
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- { . ‘r + oo
| f | © g % g °
| |
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0% ‘ | 5 l | ‘ | B
. , L | | : |
L , ; | l I . { I S
5 . 1‘ ! + . — | I
b o | _
‘ © 793.8 kev | 7 o -
'm_""_1ymTe{¢ 852.3 kev | i
2 ‘ $1206.6 kev b — = |
| | ' ‘ ‘ ‘
e
109 l__ ! . 1 ! . } ‘ L I
0 0.2 0.4 0.6 0.8 1.0 1.2 1.4 1.6 1.8 2.0 2.2 2.4 2.6 2.8 3.0

Fig. T.36.

DECAY TIME AFTER REACTOR SHUTDOWN (days)

Activity of 131Mpe after final reactor shutdown in the
MSRE primary heat exchanger (through the shield plug hole).

0€T
—_
-
-

o

N

S,
S

8]

MUCLIDE [(dis./min)/cm2 of heat exchanger fube]

ORNL-DWG 7C-9603

 

T I I
e b e e e e - - ——— . ek e .- + . - - . 1 R _—
. r ; . j ]
A ) ! '
AT a ‘

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

; |
L e I ] |
& S |
o | ? 4 b 4
© o ! . ¢ o¢ ¢: o ¢ |
e | et o .
L e ey
b i o 43, [0 364.5 kev ! - ‘ - f - e
| 1{4)637.0 kev ___,—#__-_}. - ; i °° % o 30 - °© o, ° |
e 0 667.7 kev T ' ‘ -
Sl 13219546 ke L - e
i 4-1398.6 kev | l | i ll
W _—"mm‘32Te{A:228.2kev l ! o '"”"W““"m“”{” T
. ‘ : !
! ‘
ot S
0 0.2 0.4 0.6 0.8 1.0 1.2 1.4 .6 1.8 2.0 2.2 2.4 2.6 2.8 3.0

DECAY TIME AFTER REACTOR SHUTDOWN (days)

Fig. T.37. Activity of 1317 1327 and !32Te after final reactor
shutdown in the MSRE primary heat exchanger (through the shield plug hole).

1¢T
NUCLIDE [ (dis/min) /ecm? of heat exchonger tubel

ORNL~ DWG 70-9604

 

 

 

 

 

 

 

 

 

~133m o, { 0 647.4 Kev o o |
¢ 863.9 kev N ; S S S

 

 

 

 

 

|
|
i

 

 

 

'[ _* S G — e P

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

e i + . et

 

 

 

¢

 

 

 

 

 

 

 

 

L -

 

 

 

 

 

 

 

 

0.2

0.4 0.6 0.8 1.0 1.2 t.4 1.6 1.8 2.0 2.2 2.4 2.6 2.8
DECAY TIME AFTER REACTOR SHUTDOWN (days)

Fig. T7.38. Activity of 133mTe after final reactor shutdown in the

MSRE primary heat exchanger (through the shield plug hole).

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DELAY TIME AFTER REACTOR SHUTDOWN (days)

Activity of '337 and '3%I after final reactor shutdown

in the MSRE primary heat exchanger (through the shield plug hole).

et
134

Figure 7.40. The activity of '2%%Xe was again appreciable, with a
maximum of 0.60 x 10'! dis min~* cm™2.

It might be of interest to note that during the very first spectra
taken after reactor shutdown, there was still some salt in the heat ex-
changer (in the course of draining away). These spectra contained several

1337 Once the salt

photopeaks that could be identified as coming from
was drained from the heat exchanger, those photopeaks disappeared entirely,
indicating that at least '®°I1 and, in all probability the other iodines,
remain with the salt,

It should also be noted that °'Sr was also identified in many of these

spectra,

7.8 Group I Spectra

 

These spectra were taken after the final reactor shutdown from samples
of the reactor cell air after there had been indications of an increase of

the cell-air activity.- Results are reported separately.®®

7.9 Group J Spectra

 

Spectra were taken from the coolant-salt radiator a few days after the
final shutdown and drain of the fuel- and coolant.salt systems. The ob=
jective was to determine if any radioactive corrosion products in the coolant
salt were depositing in the coolant radiator.

Since some of these corrosion products might have been activated by
delayed neutrons in the primary heat exchanger, those corrosion products
in the coolant-salt radiator could possibly be observed with the gamma-ray
spectrometer. The detector was set up in front of the radiator with the

radiator doors openj no collimator was used. Since the MSRE main off-gas

 

18p. H. Guymon et al., Preliminary Evaluation of the Leak in the MSRE
Primary System which Oceurred During the Final Shutdown, internal memo-
randum (April 197Q).
S

NUCLIDE [(dis/min}/cm2 of heat exchanger tube]
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DECAY TIME AFTER REACTOR SHUTDOWN (days)

Fig. 7.40. Activity of '?°Xe after final reactor shutdown in the

MSRE primary heat exchanger (through the shield plug hole).

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0.4 06 0.8 1.0 1.2 1.4 1.6 1.8 2.0 2.2 2.4 2.6 2.8 3.0

CET
136

line, although heavily shielded, passes not far away from this detector lo-
cation, it appeared that the majority of the recorded photopeaks came from
the off-gas line. Typical examples of identified nuclides that should not
be expected in the coolant radiator were: °°Nb, °°Mo, *°°Ru, *°°®Ru(Rh),
1317 132pa_1, '3%%Xe, '*7Cs, and '“°Ba~La. These nuclides formed the major
peaks in the recorded spectra.

Some other nuclides could be identified with some degree of confidence.
An approximate relative activity has been calculated for these in relation
to the activity of °°Co (Table 7 13).

Because the unshielded detector obviously picked up gamma rays from
different locations, the identified corrosion product radiations do not
necessarily need to have come from the coolant radiator. A collimator
would have been useful in identifying the sources of the activities; however,

the activity level was too low to permit use of the available collimation

equipment.

7.10 Group K Spectra

 

The purpose of this study was to measure the flow rate of coolant salt
in the loop. Certain components of the coolant salt will be slightly acti-
vated by delayed neutrons in the primary heat exchanger to produce *®N and
'?F. By measuring their activity at two places along the coolant-salt line,
it is possible in principle to determine the flow rate from these decaying

nuclides. The results of these measurements have been reported elsewhere.'’®

7.11 Group L Spectra

 

A graphite sample was lowered into the fuel pump bowl through the
sampler enricher; this sample was left for several hours in the fuel pump

bowl and then moved back into the sampler enricher. A special arrangement

 

*°C. H. Gabbard, Reactor Power Measurements and Heat Transfer Performance
in the MSRE, ORNL-TM-3002 (May 1970).
137

Table 7.13. Relative activities of corrosion product
nuclides found in the survey of the coolant radiator

These nuclides need not necessarily be in the radiator

 

 

Confidence in Disintegration rate
Nuclide nuclide identification relative to °°Co
®°Co Good 1
2%Na Reasonable 0.06
*icr Low %2.5
*°Fe Good 0.4

S5Ni Low z3,5

 
138

of a collimator with the detector was required to record gamma-ray spectra
from this sample as soon as it was back in the sampler enricher. The re-
actor was at full power during this experiment. The objective was to study
the deposition of short-lived fission products on graphite. A similar ex-
periment was performed with a sample capsule filled with fuel salt as well
as with a dummy sample capsule without salt., These experiments have yet to

be analyzed in detail.

7.12 Group M Spectra

 

Several miscellaneous spectra were taken from the lube o0il system, the
523 off-gas line, and the roughing filters.

Argon instead of helium was used as a reactor system purge gas for
some special experiments to study the influence of a different gas on the
stripping efficiency of noble fission gases in the fuel pump bowl as well
as the change in the bubble fraction in the fuel salt. Since part of this
argon could circulate with the fuel salt in the reactor system, some argon
activation was to be expected. As a result, we even found some “*Ar in the
fuel pump lube o0il system during this experiment. This was proved by placing
the detector near this lube o0il system. Apparently “*Ar was entrained by
the lube o0il from the fuel pump to the lube o0il system outside the reactor
cell,

Shortly before the reactor shutdown on June 1, 1969, a plug developed
in the main reactor off-gas line. Purge gases were then exhausted through
the pump bowl overflow tank and its off-gas line (line 523). A plug also
developed in this line after a period of time. Through other methods it
was established that the plug was in a particular section of the 523 line.
In an effort to locate the plug precisely, the gamma-ray spectrometer was
used to survey that pipe section. This method would have been viable if
the plug had consisted of radiocactive material. Although there was some
increase of activity near a valve in the line, we could not identify the
real location of the plug. Afterward, when this pipe was taken out and
inspected, it was learned that the plug consisted of carbonaceous material;

hence it is not surprising that we did not detect any activity increase,
139

The nuclides identified in this pipe section were similar to those found
in the main off-gas line such as noble metals and decay products from noble
fission gases.

The roughing filters, or prefilters, are used for the containment ven-
tilation system. The ventilation areas normally comprise the reactor high-
bay area but also include the reactor and drain tank cells during mainte-
nance operations when the reactor cell containment is opened. During peri-
ods of maintenance, one can expect some activity from the reactor cell (main-
ly activation products) to collect on these filters. During replacement of
the roughing filters, some gamma-ray spectra were taken to check on the
nature of the retained activity. Table 7.14 shows the nuclides identified
from these spectra together with their activity in relation to °°Co. The
spectra were recorded many weeks after use of the filters, so no shott-

lived activities should be expected.
140

Table 7.14, ©Nuclides identified from gamma-ray spectra
recorded from the containment ventilation
roughing filters

 

Confidence in Disintegration rate

 

Nuclide identification relative to ®°Co
®°Co Very good 1
5%Fe Very good 0.54
Sicr Very good 0.50
*“Mn Questionable 0.55
120Mp g Very good 0.17
193Ru Very good 1.32
198Ru-Rh Very good 1.89
?3Nb Very good 0.77
'37Cs Reasonable 0.18
1827y Reasonable 0.085
2%Na Reasonable 0.005
12%gh Questionable 0.069

 
141

8. CONCLUSIONS

The primary purpose of this study was to compile, in a readily usable
form, a block of data on certain aspects of the behavior of fission products
in the MSRE. We believe that the results given in the previous chapters are
a fair representation of the analysis of all the recorded spectra that per-
tain to the fission product deposition in the parts of the MSRE that could
be examined by this method.

In general, we have found remote gamma-ray spectrometry to be readily
applicable and productive of much useful information about the MSRE. Many
observations (e.g., possible discovery of new gamma rays, fundamental evalu-
ation of collimators, etc.) are believed to be in disciplines that are not
directly connected to the fission product behavior of this reactor and con-
sequently are omitted from this report,.

We realize that this work is only part of a very complicated matter
in which the coordination of several specializations might be required to
develop a more complete picture of fission product behavior in the MSRE,
Hopefully, the information presented in this report can be combined with
other data to achieve that goal.

This chapter presents some of our observations about the performance
of this study. In addition, we draw some conclusions concerning the be-

havior of certain fission products.
8.1 Data Collection and Analvysis

8.1.1 Gamma spectrometer system

One of the main reasons that we were able to analyze these complicated
spectra successfully was the excellent resolution capability of the de-
tector. Since we had to collimate and shield the incoming gamma rays con-
siderably, there was no need for a large, expensive, high-efficiency de-
tector. For these studies a medium-sized (26-cm®) and hence rather inex-

pensive high-resolution crystal sufficed.  The need for a multichannel

 

*
It should be noted that although the detector was only of medium size,
it had a rather high peak-to-Compton ratio (27:1), thus permitting detection
of many low energy gammas that would not otherwise have been observed.
142

analyzer (4096 channels) with magnetic tape recorder seems indispensable
with such a detector. Since our spectra were complicated, the spectrum
analysis could only be done efficiently by computer. The computer program
was complex and required a large memory, 610 k-bytes. It is doubtful that

a smaller computer coupled directly to the analyzer system would have served
Our purpose.

The collimator assembly served its purposes well; with the proper
choice of a collimator insert, we could limit the total amount of radiation
to the detector without excessive use of shielding material. Also, we
were able to observe fission product depositions in reactor components at
very definite locations. Since the assembly was mounted with three adjust-
able set screws on an axial thrust bearing, instead of directly on the
portable maintenance shield, we could easily make small adjustments to the
assembly orientation. However, it was difficult to make a straight col-
limator beam hole, The idea of a free-~floating thick-walled stainless
steel precision tube around which the shielding lead is poured can be
fruitful if some more thought is given to it.

The locational equipment, that is, laser and surveyor's transits, per-
formed very satisfactorily and proved to be precise, reliable, and inex-
pensive. Since we moved the detector frequently, it is felt that any so-
phistication of this equipment (e.g., such as mirrors to guide the laser

beam around the detector) would have been detrimental to its reliability.

8.1.2 Calibration

Very little is known about the efficiency of gamma-ray detectors in
relation to a collimated beam. This, together with the complexity of the
heat exchanger geometry, led us to believe that the effort given to the
empirical calibration of equipment was worthwhile. The fact that differ-
ent spectra recorded from the same spot on a component, but taken with en=-
tirely different shielding configurations, yielded virtually the same ac-
tivities lends confidence to the calibration results obtained.

Silver-110m proved to be a good calibration source. If one has re-
actor irradiation facilities available, such a source can be made rather
easily. The extension of the efficiency curves, as obtained with **°"Ag,

to both the lower and higher energy ranges was possible by using the actual
143

fission product spectra obtained during the experiment itself. The rela-
tive efficiency curves obtained from the fission product spectra were
consistent with the absolute efficiency curves found with the calibration

source.

8.1.3 Computer analysis program

 

Considerable time and effort were required to convert the computer
analysis program, which originated at Lawrence Radiation Laboratory, for
operation at the ORNL computing center. The program is quite powerful and,
generally speaking, satisfied our needs well. For application to spectra
like ours, where there are so many multiple photopeaks, however, we can en-
vision several improvements, particularly in the analysis of multiplets.

Our table of radionuclides, as used in the library of the computer
program, was a very useful todl for the analysis of the spectra. This
table might be also of interest to others working in the gamma-spectrometry
field, and so is presented as Appendix A. Although this table contains the

latest published data, we realize that it is far from complete.

8.2 Results — General

 

We will draw some general conclusions concerning the fission product
distribution in several reactor components, without, however, going here

into detail on the different decay chains.

8.2.1 Metal surfaces in direct contact with the fuel salt

Metallic fission products, such as isotopes of niobium, molybdenum,
ruthenium, rhodium, antimony, and tellurium, were present on the walls of
the reactor system that were in contact with the fuel salt. Their depo-
sition on the walls represented, in general, a large fraction of the total
amount of that nuclide present in the entire reactor system.

Because of additional metal surface exposed to the salt near the baf-
fle plates in the primary heat exchanger, one might expect a small increase
in activity there. It appears, however, that the observed large increases
in activity near these baffle plates (two to four times higher than in
intervening areas) are due to additional effects. It may be that the salt
flow pattern influences the deposition rate of these fission products.

For example, niobium, molybdenum, and ruthenium-rhodium exhibit a higher
144

activity increase near the baffle plates (2.5 to 4 times) than antimony,
tellurium, and iodine (1.5 to 2 times).

A study of the activity of the several iodine isotopes directly after
reactor shutdown and subsequent fuel drain during the final reactor shut-

131

down in December revealed that the iodine activities ( I and especially

'327) build up after reactor shutdown. Although we were never able to

detect **°I after the drain of the fuel, we did detect it during the drain.

Therefore we conclude that the observed large iodine activities in the

empty system were mostly due to the decay of their precursors that deposited

on the wall; iodine itself, however, remains with the fuel salt.

In comparing the relative activities of the deposited metal fission
products (relative to °°Nb) in the heat exchanger and in the fuel lines
(101 and 102), there is no major difference in the deposition rate on the
heat exchanger (between the baffle plates) and on the fuel lines. Molybdenum
and ruthenium are possible exceptions; their relative activity appears to be
somewhat higher in the fuel lines. It should be stated, however, that this
is a tentative conclusion since we did no calibration work on the entirely
different geometry of the fuel lines.

The amount of activity in the heat exchanger caused by the decay of
noble fission gases is appreciable. Although most gas activities disap-
peared after a few hours because of decay and some circulation of purge
gas, their decay represents a significant heat source during the first hours
after drain of the fuel salt. These noble fission gases probably came from
several sources:

1. A release of gases absorbed in the graphite bars in the reactor core,

2. A release of gas bubbles from the fuel salt while in the course
of draining to the drain tank.

3. A back-flow of gases from the holdup gas volume of the main reactor
off-gas line. In order to drain the fuel salt, a gas pressure
equalizer line is necessary between the drain tanks and the fuel sys~
tem; this line (521) ties into the main reactor off-gas line at the
4-in,-ID gas holdup section. A calculation of the volume of these
back-flowing fission gases during drain largely explains the unex-
pectedly high activity of noble fission gases in the reactor system

after shutdown.
145

It is reassuring to note that the magnitudes of the activities de-
tected in the heat exchanger during the July shutdown were very close to

those detected during the November survey.

8.2.2 Main reactor off-gas line

As might be expected, the major activities in the main reactor off-gas
line during reactor operation are due to the decay of noble fission gases
and their decay products. These nuclides still form a large source of ac-
tivity shortly after reactor shutdown, but they decay rapidly and are di-
luted by purging of the empty fuel system.

Metals such as niobium, molybdenum, ruthenium, rhodium, antimony, and
tellurium (iodine) could barely be identified during reactor operation but
formed the major activity source soon after reactor shutdown.

Since we were not able to detect '?°I in the off-gas line, although
other iodine isotopes with metallic precursors were identified,* we believe
that iodine proper does not separate from the salt and will neither disap-
pear in large quantities into the off-gas line nor remain with the metal
walls when the system is filled with circulating salt.

We were not able to positively identify any nuclide, such as °°Zr,
that is supposed to remain with salt; that is, we could not detect any ap-
preciable quantity of fuel salt in the off-gas line. Although on a few
occasions the computer program found a photopeak possibly due to *°Zr, we
scrutinized the actual plots of these spectra and were not convinced of the
presence of salt within the range of our nuclide detection sensitivity.
Concerning the off-gas line geometry, this meant that we could not detect
activities less than 0.5 x 10'*° dis min™" in.”! or an activity less than
5 x 10=* of the detected °°Nb activity.

Taking into account the purge-gas flow rate through the jumper line,

it is not surprising that we did not detect a decrease in nuclide activity

along the jumper line.

 

%
The precursors of **°I have half-lives too short for significant
amounts to escape from the salt.
146

Comparing the activities calculated from the July and November surveys,
the longer-lived nuclides, such as '°°Ru-Rh and '*’Cs, seem to be con-
sistent; °°Nb, ‘°?Ru, and '?°"'Te appear to be about four times higher
during the November survey. This could partly be explained because of the
power history before shutdown as well as by the fact that the main off-gas
line was partially plugged during the latter part of the power operation
before July. (At that time the off-gases were routed through the off-gas
line of the fuel pump overflow tank.) Shorter-lived nuclide activities

(e.g., **'I) would be even more influenced by this plugging problem.

8.3 Elements and Nuclide Chains

 

Although it is clear that the reported results are only a part of the
total picture of fission product behavior, it is possible to draw some ten-
tative conclusions concerning the behavior of certain elements and nuclide
chains. 1In particular, it is of interest to examine the behavior of iso-

topes that belong to the same element or to the same decay chain.
8.3.1 Niobium

Both ®°Nb and °’Nb were identified in the heat exchanger after the
November and the final shutdown. Because of the short half-life of °’Nb
as well as its precursor, the disintegration rate of this nuclide reflects
only the deposition of niobium shortly before shutdown. The °°Nb activity
would represent a much longer history where chemical conditions of the
salt were different.

Because absolute activities derived afterxr the final shutdown may be
uncertain, we have compared the ?’Nb results for the November and final
shutdowns by normalizing them to the °°Nb values. Table 8.1 shows these
results corrected for decay to reactor shutdown time,* Also shown are
half-lives and fission yields. These data pertain to the area in the heat

exchanger under the shield plug hole.

 

*
Since it is niobium that deposits on the metal walls or disappears
into the off-gas line, it appears legitimate to extrapolate back to reactor

shutdown time to obtain the maximum niobium activities.
147

Table 8.1. Comparative values of °°Nb and °7Nb

 

Activity relative to

 

 

°5Nb activity in heat exchanger Fission yielda
November shutdown Final shutdown (%) Half-life
?SNb 1 1 6.05 35.5 days
7Nb 0.63 3.0 5.62 72.0 min

 

91t is estimated that the total fission rate in the MSRE was due to
the contribution of the following isotopes: 233U, 94%; 2°°U, 2.25%;
23°pu, 3.75%. Actual fission yields were calculated from B. M, Rider,
A Survey and Evaluation of Thermal Fission Yields, GEAP-5356.
148

Concerning the activity results after the November shutdown, one might
expect the °’Nb activity to be in equilibrium with its environment whereas
the ?°Nb activity is not; the uninterrupted power run was less than two
months.

A possible explanation for the November shutdown ®’Nb/®°Nb activity
ratio might be to assume a time lag between formation and deposition. An
average time lag, or residence time in the salt before deposition, of ap-
proximately 100 to 150 min would explain the calculated ratio of 0.63.

The small ?°Nb buildup during the short power run after the long shut-
down period in November, as well as two beryllium additions prior to the
final shutdown (which are known to affect the niobium behavior), might ex-
plain the rather different activity ratio of 3.0 after the last shutdown.

Currently it is thought that the oxidation potential of the salt, which
is influenced by the beryllium additions, affects the niobium solubility in
the fuel salt, This, of course, forms another dimension to the deposition
problem. Therefore, a study of the °’Nb concentration in the salt in re-
lation to the oxidation potential might be quite useful in future reactor
systems. The reason for using the °’Nb concentration is that it is not in~
fluenced by a long reactor power history as is °°Nb. A simple gamma-spec-
trometry assay done at the reactor site would be ample.

We were not able to positively identify ’Nb in the main off~gas line;
there were indications of this nuclide, but too few to assign a numerical

value to it.
8.3.2 Ruthenium-rhodium

Three ruthenium nuclides were identified in both the heat exchanger
and the off-gas line: *°?Ru, *°°Ru, '°°Ru-Rh. Since it was assumed that
the rutheniums are the first nuclides in their respective fission decay
chains that deposit on the metal walls or escape into the off-gas line,
an extrapolation to reactor shutdown time would then be legitimate to ob=-
tain their maximum activities, This might not be the case for *°®Ru, since
its precursors, *°°Mo and ‘°°Tc, also will deposit on the metal walls; it
would, however, not introduce a large error in our evaluation of the maxi~-

105

mum Ru activity. Table 8.2 shows comparative values for these three

isotopes. The activities are given relative to the activity of *°°Ru-Rh.
149

Table 8.2. Comparative values of three ruthenium isotopes
identified in the reactor system

 

Activity relative to '°°Ru-Rh

Heat exchanger, Off-gas line, Heat exchanger,
Nov. shutdown Nov. shutdown Final shutdown Half-life Fission yielda

103p. 3.6 7.4 1.2 40 days 1.80
1°5Ru 6.0 1.3 6.2 4.43 hr 0.60
"°®Ru-Rh 1.0 1.0 1.0 367 days 0.41

 

aIt is estimated that the total fission rate in the MSRE was due to the contri-
bution of the following isotopes: 223U, 94%; ?3°U, 2.25%; *°°Pu, 3.75%. Actual
fission yields were calculated from B. M. Rider, 4 Survey and Evaluation of Thermal

Fission Yields, GEAP-5356.
150

Let us look first at the *°?Ru and *°®Ru-Rh. Taking into account the

103
h

whole power history of the reactor one can calculate how muc Ru and

*°®Ru are present in the entire reactor system: in the salt, on the graph-

ite the off-gas line, or on the metal walls. These calculated inventories

103 106

yield a ratio of the Ru activity relative to the Ru~Rh activity of

roughly 6.5 both for the November shutdown and the final shutdown. Because
of the difference in decay half-lives of these isotopes, this ratio should

decrease with increasing age of the mixture. Tentatively one would conclude

103 106
R_

from the November shutdown ratio of 3.6 that the u Ru~Rh mixture has
an average age of several weeks or that ruthenium does not tend to deposit
very readily on metal walls.

The same ratio calculated for the off-gas line is 7.4. This seems to
suggest that ruthenium separates from the salt and disappears into the
off-gas line (or possibly deposits on the graphite rather than on the metal
walls). 1In other words, since we could barely detect ruthenium in the fuel
salt in the drain tank, it appears that this element has only a slight ten-
dency to deposit on metal surfaces and rather will disappear into the off-
gas line (or deposit on the graphite) Data on the deposition on graphite

would be very important for a complete picture of the ruthenium behavior.

105 103

If one assumes that Ru deposits by the same mechanism as Ru and

'°®Ru, the '°°Ru data from Table 8.2 are rather hard to explain. One al-

ternative possibility would be to assume that it is the short-lived *9%Mo

105 105

or Te that really deposits, the observed Ru in the heat exchanger

being mainly the result of its precursors deposition. A check on the ac-
tivity of ‘°’Ru relative to “°Mo both in the off~gas line and the heat ex-
changer does not exclude this possibility.

Because many other photopeaks are adjacent to or even coincide with
its peaks, *°°Rh is difficult to identify. From a study of Figs. 7.8,

7.15 and 7.34, two conclusions may be drawn. The results concerning the

heat exchanger, Figs. 7.8 and 7.34, suggest that the *°°Ru activity is
1053

equal to or higher than the Rh activity; the opposite is true for the

off-gas line. Also, the '‘°°’Rh does not gradually build up from nothing

105

to a maximum in the heat exchanger. Apparently, Ru remains on the metal

walls after its formation from *°°Mo-Tc. The *°°Rh seems, to certain degree
151

at least, to dissolve from the heat exchanger and then partly transfer to
the off-gas line.

In view of the *°°Ru—*°®Ru-Rh behavior we envision the following pos-
sibility for the 105 decay chain: molybdenum deposits more readily on the
heat exchanger than ruthenium; in agreement with the °°Mo data, it also
escapes into the reactor off-gas line. The ruthenium does not desorb from
the metal walls; therefore the observed deposits of this element on the
heat exchanger and in the off-gas line are merely the decay products of
the deposited molybdenum, Rhodium, formed by the decay of ruthenium, not
only decays on the metal walls but also goes back into the salt and eventu-
ally disappears appreciably into the off-gas line. The fact that only very
small amounts of various rutheniums were found in the salt confirms the
contention that these nuclides transfer rather quickly from the salt into

the off-gas line (or onto the graphite).
8.3.3 Antimony—tellurium—iodine

The identification of *2°Sb, *2?°Sb, *?7Sb, and **°Sb in the heat ex-
changer and in the off-gas line leads one to believe that antimony is at
least partly instrumental in the presence of tellurium and, as a result,
also of iodine after drain of the reactor system. The emerging question
then is: What governs the tellurium deposition? There are several possi-
bilities, the extremes of which are (1) all tellurium, after formation from
decay of deposited antimony, returns to the salt and moves elsewhere; that
is, when the system contains fuel salt, no tellurium is present on the metal
walls; (2) tellurium deposits or remains on the wall,

Let us take an example of the fission decay scheme of the elements of

mass number 131:

827%

1
131y 1317
s

 
152

Antimony-131 decays for 6.8% to the isomeric state of tellurium, *®'Te,
and for 93.2% directly to the ground state, '*'Te; '*'Te decays for 82%

1211 and for the rest first to the ground state and then to

directly to
1317 Table 8.3 shows the cumulative fission yields as well as the decay
half-lives.

In the first case assumed above, tellurium would remain on the metal
walls only after the drain of the system., Because of the short half-life,
the equilibrium concentration of '®'Sb on the walls would be relatively
small, After shutdown, the iodine activity would build up rather quickly
since both *®'Sb and '®'Te have short half-lives. The buildup of *3*'I
through '?'Te would be small and provide roughly 5% of the maximum iodine
activity. Since we did not detect any tellurium in the salt, the bulk of

this element would then have to be either on the graphite or in the off-gas

line,

131

In the second case, Te and '*'"Te would be also in equilibrium on

the metal walls when the salt is still in the system; this means that their
decay rates are governed by the respective fission yields. Since 131Mre
has a much longer half-life than '®'Te, the equilibrium concentration of
'31mTae is much larger than that of *>'Te. The buildup of '°'I is expected
to come from three sources: “>'Sb, '?'Te, '?*YTe, 1In view of the fission
yields and decay half-lives of these iodine precursors, roughly 70% of the
maximum iodine activity would be due to **'"Te and the remainder to *°?Sb
and '>'Te. The maximum iodine activity would occur approximately three
days after reactor shutdown; however, this maximum would not be very pro-
nounced, since the iodine activity from '®'Sb and *°'Te builds up quickly
after reactor shutdown.

Let us now examine the different figures and try to decide what actu-
ally happened in the heat exchanger.

Figure 7.18. Ve believe that the data based on the 364.5-keV photo-
peak are the most trustworthy, although the values might be somewhat too
high shortly after reactor shutdown because of a prominent °°Kr photopeak
at 362.6 keV. Apart from the first data after shutdown, one would conclude
from this figure that the *>'T does build up to a maximum activity two to

three days after reactor shutdown. This observation would be in line with
153

Table 8.3. Information concerning the element-~131
fission decay chain

 

 

Nuclide Half-life Cumulative fission yield® (%)
t31gp 21 min 2.93
13177, 1.2 d 0.20
1317a 25 min 2.77
1317 8.07 d 2.93

 

Tt is estimated that the total fission rate in the MSRE
was due to the contribution of the following isotopes: ?*°°U
94%% 223y, 2,25%; 2°®°Pu, 3.75%. Actual fission yields were
calculated from B, M. Rider, 4 Survey and Bvaluation of Thermal
Fisston Yields, GEAP-5356,

3
154

31T activity

the above-mentioned second hypothesis. The relatively high
in relation to, for example, °°Xb or °°Mo would also favor the second case.

Figure 7,17 . If the first hypothesis were to hold, one would expect
the **'"Te activity to be appreciably lower than the **'I activity because
the amount of '°*'Te formed from *3'Sb would be quite small. This is defi-

nitely not the case.

Figures 7.18 and 7,19 . Antimony-132 has a decay half-life of approxi~

 

mately 2.1 min and *°'Sb approximately 25 min. (The decay half-lives and
fission yields of these two isotopes do not appear to be very well known.)
This means that their equilibrium concentrations on the metal walls, also
taking into account the fission yields, differ by a factor of at least 8
to 10, Consequently, if the first hypothesis were correct, the ***I should
be higher than the ***Te~I activity This is definitely not the case.
Although it is fully acknowledged that a more precise analysis should
be made of the antimony-tellurium-iodine data, it seems reasonable to con-
clude from the above that both the antimony and tellurium isotopes do de~-
posit on the metal walls while the fuel salt is in the system. Both of
these elements seem to contribute to the total iodine activities. ITodine
itself, however, will not remain on the metal walls as long as there is
salt in contact with these walls. A study of the antimony, tellurium, and

iodine activities of the other fission decay chains basically points to the

same conclusion.

8 3.4 Extrapolation back to reactor shutdown time

In order to plot and compare the recorded data during the actual sur-
vey of the heat exchanger and main reactor off-gas line, all activities
were extrapolated back to reactor shutdown time. This seems a reasonable
procedure to obtain the maximum activities of the deposited elements. There
is possibly one exception to this rule, *>'I. Because '’'I appears to at-
tain its maximum activity onlv after roughly three days, the extrapolation
would overestimate the maximum *°*I activity by about 20%. Thus, '°'I data
given in Fig. 7 .28 and Tables 7.7 to 7.9 could be considered 20% too high.
Since all data of the heat exchanger and main reactor off-gas line were
taken during the actual survey, three or more days after reactor shutdown,
131

these extrapolated I activities are consistent among each other,
155

9. EPILOGUE

We observed a rather large, unexpected activity due to the fission
gases krypton and xenon in the heat exchanger. This could be explained
by taking into account a flow of these gases from the main reactor off-gas
line back into the reactor system during the drain of the fuel salt. Actu-
ally, if one calculates the volume of the gases involved and their average
age after fission, the observed activity in the reactor system can be ap-
proximately accounted for. Where substantially larger activities are in-
volved with large molten-salt reactors, a different layout of this gas
equalizer line should be considered to avoid such an activity surge.

With the exception of the iodines, all identified fission products
migrated to the walls of the heat exchanger (and consequently to other
metal walls in the reactor system), while the fuel salt was in the systemn.
Unless the flush salt has entirely different chemical properties, it is
difficult to expect that this salt would dissolve the deposited elements.
With the exception of '*'I, *?3*I, and '°“I, the current flush salt is not
expected to reduce appreciably the deposited activities; it still might be
useful, however, to dissolve pockets of fuel salt or absorb radioactive
dust.

In the previous chapter we have tried to define a few causal relation-
ships for the observed activities of deposited fission products. In view
of larger molten-salt reactor systems, it might be useful — beyond academic
interest — to study the recorded data in much more detail. TFor example, a
detailed study of the niobium, molybdenum-~ruthenium-rhodium, and antimony-
tellurium~iodine decay chains should be of interest since these elements
seem to represent the major heat sources in the empty reactor system and
main reactor off-gas system.

After this satisfactory experiment, in which a sensitive gamma-ray
spectrometer was used in a very intense radiation field, we feel confident
that a similar technique could be very useful in larger molten-salt reactor
systems. For example, a remote gamma-ray spectrometer aimed at a bypass
line of the main fuel system could economically reveal very useful infor-
mation: the ®’Nb/°°Nb ratio in connection with the redox potential of

233

the salt and the Pa concentration.
156

Although the gamma spectrometry system used in this study was in most
respects highly satisfactory, two changes could be made that might signifi-
cantly extend its usefulness. One change pertains to the detection of weak
gamma lines (small photopeaks) in the presence of strong lines. Each gamma
ray produces, in addition to a photopeak, a Compton continuum distribution
with an energy range varying from O keV to just below the full energy of
the gamma ray. A count in the Compton distribution arises when a photon
loses only part of its energy in the detector. The remaining energy leaves
the detector as a Compton scattered photon.

Although the detection of low-energy gamma rays is favored because the
counting efficiency rapidly increases as the gamma energy decreases, this
effect is partly offset if higher-energy gammas are present. Because of the
Compton distribution of the higher-energy gammas, the lower-energy photo-
peaks will be masked to an extent that depends on the relative intensities
of the various gammas. If they are of sufficiently low intensity the low-
energy photopeaks will be completely masked.

In the past few vears, a technique has evolved to greatly increase the
peakfCompton ratio and thereby significantly extend the ability of the sys~-
tem to detect low-energy gammas in the presence of high-energy gammas.

This technique makes use of a Ge(Li) detector enclosed in a much larger de-
tector called sn anticoincidence mantle. The mantle is often made of NaI(Tl),
but can be a plastic scintillator. Ports are provided in the mantle to pro-
vide access for the Ge(Li) detector and a beam of gamma rays. The gamma
beam is allowed to hit the Ge(Li) but not the mantle. Gamma rays that lose
all their energy in the Ge(Li) produce a pulse (count) in the full energy
peak, Most of the gamma rays that scatter out of the Ge(Li) detector will
be detected by the mantle. For such events, pulses are detected simultane-
ously in both detectors. By proper electronic gating the spectrometer can
be made to reject these coincident pulses and store only those that occur
singly in the Ge(Li) detector. The resulting anticoincident spectrum has

a very large full energy peak but very low Compton distribution. Because

of the lower Compton distribution, the ability of the system to measure
low-energy photons in the presence of high-energy photons is greatly en-

hanced., An anticoincident spectrometry system has been used successfully
157

20

by Holm et al. in scanning reactor fuel elements and would have compar-
able merit in studies similar to our MSRE measurements.

The second suggested change in the spectrometry system pertains to an
improvement in the accumulation of spectra under transient conditions. Dur-
ing transient conditions, such as reactor startup, scram, or modifications
of fuel or salt chemistry, it is desirable to collect as much spectral data
as possible. To obtain suitable resolution of the large number of peaks
and permit computer processing of spectra, it is necessary to collect spec-
tra with a gain of about 0.7 keV or less per channel. However, with such a
gain a 4096-channel spectrum will only extend to about 3000 keV. Because
several of the short-lived fission products emit photons above 3000 keV,
their photopeaks will be missed. Under transient conditions there is not
enough time to obtain spectra at two different gain settings, and the higher
energy portion of the spectrum is lost. The total spectrum can be obtained
in two ways. First, a spectrometer system with about 7500 channels could
be used; this would permit measurement of the 5230-keV photon of °°Rb,
which is the highest energy photon among the fission products with half-
lives long enough to permit measurement in most experiments. Second, two
spectrometers, one with 4096 channels and another with 400 channels, could
be used. Because fission products emit only a few photons above 3000 keV,
it is unnecessary to collect spectra at low gains above this energy. The
two-spectrometer systems would be used by routing the signal from the de-
tector preamplifier to two amplifiers. One amplifier would be operated
at a gain of 0.75 keV per chamnel and its output processed by the 4096~
channel analyzer to produce a spectrum with an energy range of 0 to 3000
keV, The other amplifier fiould be operated at about 5 keV per channel and
its signal processed by the other analyzer. A zero-suppression amplifier
would be interposed between the main amplifier and the 400-channel analyzer
to suppress the lower 3000 keV. Pulses above 3000 keV would be processed
to yield a spectrum between 3000 and 5000 keV.

With either of the above schemes, spectra could be obtained over the
full energy range of fission product photons, and no information would be

lost during critical moments of transient reactor conditions.

 

20D, M. Holm et al., "Examination of Fast Reactor Fuel Elements with a
Ge(Li) Anticoincidence Gamma-Ray Spectrometer," pp. 22842 in Proceedings of
the 16th Conference on Remote Systems Technology, Idaho Falls, Idaho,

Mar. 11—13, 1969.
159

APPENDICES
161

Appendix A
TABLE OF RADIONUCLIDES

The successful analysis of the gamma-~ray spectra, in particular the
identification of the nuclides, depends very much on the nuclear data stored
in the nuclide library of the spectrum analysis program. Considerable
effort was expended in preparing a table of radionuclides for this library,
based on the most updated information from the Nuclear Data Project located
at ORNL. Since the absolute nuclide concentrations calculated by the com-
puter program depend on the supplied nuclear data, we believed it to be use-
ful to include this table. 1In case better nuclear data that are significantly

different become available, the results presented in Chapter 7 should be

changed accordingly.
LIST OF INPUT CATA

THE HEADING SYMBOLS A,B,C,D,E.F,G,AND H FOR THE INPUT TABLE HAVE THE

A=PRECURSOR S5YMBOL MAINLY USED FCR THERMAL NEUTRTN REACTICNS
E=PRECURSCR MASS NO.
C=PRECURSCR NATURAL ABUNCANCE
D=NEUTRON REACTICN CROSS SECTION
E= RACIGONUCLIDE TYPE,WHERE
C=N,GAMMA PRCDUCT
1=FISSICN PRCDUCT
2=B0TH N,GAMNMA AND FISSION PRCDUCT
3=NEUTRON DEFICIENT NUCLIDE
4=NATURAL RACIUNUCLIDE
S=(THER RADICNUCLIDES
= N0« COF GAMMA ENERGIES TABULATED
G = BRANCHIIG RATIC STATUS,WHERE
A=ABSGLUTE
R=RELATIVE
D=BASED ON DECAY SCHEME GIVEN IN REFERENCE

n

FOLLOWING MEANING

291
TELE. AT, CECAY

NCe ISCTCGPE NOs HALFLIFE UNIT MULOE A B C D E FISSION YIELD
1 019 8 26,500 S B- 0 18 C.2C4 0.00C O 0.0
ket 43 ADAMS AND DAMS J,RAGCICANAL CHEM =-ENERGIES BLUE BOOK FOR OTHER DATA
ENERGY INTENSITY G ENERGY INTENSITY G
167.4C0 CT.C00 1356.000 5¢.000
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ENERGY INTENSETY G
£z4.810 100.000
€ €O %6 27 77.300 D ECB+ C G.C 0.0 3 0.0
REF 0 NUCLEAR DATA TABLES SEC A 1970
ENERGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G ENERGY INTE
511.G06 484000 733,700 0.110 788.0C0 0.400 846,750
©77.5C0 1.400 1937.900 12,900 1141,20C C.170 1175.,130
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2451 .400 0.780 3548,200 0.180
7 CC &0 27 5260 Y B~ CO 5% 100.,GC0 37,5060 ¢ 2.0
REF 0 NUCLEAR DATA TABLES SEC A 1970

ENERGY INTENSITY G ENERGY INTENSITY G
1173.23¢0 92.880 1332.51¢ 1C0,000 °

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403.000 59,700 6744300 2.000 836.000 0. 760 845,800 8,200
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ENERGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G
£84G00 0.0 166.000 6.800 196.100 37.800 2404400 0.300
362.L00 3.000 390.400 0.600 4724300 0.600 789.000 0.400
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562.200 0.070 986,700 1.600 1017.60CC C.110 1039.600 0.400
1C49,500 G.100 1090,400 c.070 11414700 1700 = 1179.,500 C. 750
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H LYCKLAMA, ET AL CAN Jo PHYSICS 47,393(196%)

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REF 42 GUNNINK ETAL UCID 15439  14JAN 1969
ENERGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G ENERGY
2754450 1.33C 6204170 1,900 631,360 04610 6524 980
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REF 42 GUNNINK ETAL UCID 15439  14JAN 1969
ENERGY INTENSITY G
12864000 90,000
18 Yy °0M 39 3,100 H IT Y 86  100.000 0,001 O 0.0
REF 43 ADAMS ANC DAMS JL.RADIOANAL CHEM —ENERGIES BLUE BOOK FOR OTHER
ENERGY INTENSITY G ENERGY INTENSITY G
2024400 97.000 479,200 91.000
15 Y %0 38 64,000 H B~ Y 86  1€0.0C0 1.300 2 0.577F
REF 5 NUCLEAR DATA TABLES SEC A 1970
ENERGY INTENSITY G
1742.700 C.016
0 Y 91M 39 50.000 M IT 0 0.0 0.0 1. 0.340E
REF 42 GUNNINK ETAL UCIC 15439  14JAN 1969
ENERGY INTENSTTY G
£55,630 £3,00C
21 y o1 30 59,000 D B- ¢ 0.C 0.0 1  0.590€
KEF 42 GUNNINK ETAL UCID 15439  14JAN 1966
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REF 42 GUNNINK ETAL UCID 15429  14JAN 1569
ENERGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G ENERGY
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REF 42 GUNNINK ETAL  UCIC 15439  14JAN 1966
ENEKGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G ENERGY
267,059 64400 478,400 04011 68C. 250 0. 700 £95,700
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92

991
ca kR ©5 49 €5.000 D B- IR G4 17.40C0 0.C080 2 Qe £Z29E
R:F 7 LEGRAND ETAL,s NUCL4PHYSe A107,177 (1968}
ENERGY INTENSITY G ENERGY INTENSITY G

T244240 43,600 7564870 54.E80C
& IR 97 40 17.000 H @~ IR Q€ 2. £0C g.a0s8¢c 2 0.5G0E
REF “2 GUNNINK ETAL UCID 15439 14JAN 1966%

ENERGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G ENERGY
254.2C0 1.2C0 3%5.700C 2.200 602.5C7 14280 703.8C0
743,279 24.0C7 1148.000 282G 1276.100 G. 300 1262.7C0

17504609 1.080 1852.000 0260

e NB S5M 41 90,000 H IT IR 94 17.4CG 0.0860 2 N.128E

KEF 43 ADAMS AND DAMS J,RADIOANAL CHEM -ENERGIES BLUE BGOK FOR DOTHER

ENERGY INTENSITY G
¢35.7C0 107,.,0¢C0

27 NB 95 41 35.150 D B- 0 0.0 0.C 1 O.641E

ReF 1 LEGRAND ETAL, NUCL.PHYS. ALQT7,177 (1968}

ENERGY INTENSITY G
TéE.84C 100,000

8 NB S€ 41 23,000 H B- 0 C.0 0.0 1 D.600E-

REF 8 NONARC+BARRETTE,BOUTARD CANeJoPHYS. 46, 2375(1968)

ENERGY INTENSTITY G ENERGY INTENSITY G ENERGY INTENSITY G ENERGY
2184900 2.900 241.600 3.300C 350.200 1. €00 352,200
271.800 2,900 460,100 26.800 4804500 5.6G0 568,700
$91.70Q 1,000 719.700 6.700 721.700 1.200 778,200
£10.,4G0 9.000 812,500 5.50C 849,800 21.000 1091.,400

1128.,000 C,700 1200.100 2C. 509 1441,300 0.4C0 1497,00

1497.%00 0.400

29 NB 9TM 41 1.000 M IY IR 96 2800 0.050 2 0.59%E

REF 43 ADAMS AND DAMS JJRACIDANAL CHEM —ENERGIES BLUE BOOK FQOR (OTHER

ENERGY INTENSITY G
7424260 98,000

30 NE 97 41 72,000 M B- IR S¢ 2.800 C.050 2 0.620E

REF 42 GUNNINK ETAL UCID 15436 14JAN 1969

ENERGY INTENSITY G

€58.180 9¢,.200

01 2
01 10 A
INTENSITY G
1.000
1.080
Qo 1 A
DATA
01 1
02 21 R
INTENSITY G
1100
56000
100.000
EC. QGO
2«56G0C
01 1 A
DATA
01 1 A

NE-3E

NE €7

L9T
zl MG 99 42 66,690 H B~ MO ¢&¥¢ 23.780 0.150¢ 2 0.610F

REF 37 VAN ETJK ETsALs NUCL, PHYS, A121,440(1968)
ENERGY INTENSITY G ENERGY INTEMSITY G ENERGY INTENSITY G ENERGY
18.251 11.000 18+.367 0.0 40.584 €+ 3C0 140.511
142.630 0.900 181.060 7.600 249.000 C.005 3664400
280.700 0.010 409,000 0.001 411,500 G« 024 458,000
528.900 0.C5%3 620. 700 0.004 €20.7C0 G.024 739.700
T7€.,200 4.800 823.190 0.140 961.000 C.110 588.200
1€21.700 0.C06 1016.C00 0.001
32 MO1l01 42 14,600 M B- MOl100 2.630 c.2C0 2 0.500E
REF l4 CRETU ETAL REV.ROUM,PHYS.,TOME11l,N0O2,P143 (1966)
ENERGY INTENSITY G ENERGY INTEANSITY G ENERGY INTENSITY G ENERGY
BC.J00 2,000 190.000 T.000 191.0CC £9.000 300.000
40G.000 2.060 510.000 2.000 51C. 000 18.000 590,000
700.000 14.0C0 890.000 14.00C0 950. 000 1.0C0 1020.000
1180.000 9.000 1280.000 2.709 1380.000C G.000 14460.,000
156u. 000 12,000 1680.000 4,000 1800.,00C 1.0C0 1890.000
208C.000 9.009 2160.600 1.000

23 TC 9SM 43 6,000 H IV MO 98 23,780 0.510 C 0.616E

REF 0 NUCLEAR DATA TABLES SEC A 1970
ENERGY INTENSITY G ENERGY INTENSITY G
140.511 89.500 142. 630 Q.027
e TCl01 43 14,000 M B- MO100 S4620 0.2C0 2 0.502E
KEF 43 ADAMS AND DAMS J.RADICANAL CHEM —-ENERGIES BLUE BCGK FOR OTHER
ENERGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G ENERGY
127,300 3,000 306. 800 91.000 544,500 8.000 626.£00
2 TC10Z 432 4,5¢C M B~ 0 0.0 0.0 1 O0.41FE
REF ¢ BLACHCT ETAL MNUCL PHYS Al13¢ (196¢) P434 GE DIODE
eNERGY INTENSITY G ENERGY INTENSITY G ENERGY INTERSITY G ENERGY
415.300 2.0N0 4184800 3,400 474,800 160,000 627.500
£30L.E00 29,700 691,800 2.700 695,600 3.000 504.000
G21.900 0.0 10464200 12.800 1073.00¢C 1.000 1104,£20
11154000 0.7 1128.0C0 0.0 1197.80¢C 10.8C0 1292.6C0
1221.80v 3,700 1811.C00 24500 1595.7C0 £+100 1612.,800
1€17.060 11.1C0 1710.000 5.500 1810,00¢C T.000 1907.000
is5C.000 0.0 2141.600 1.C0¢C 2227.C00 7«100 2242.2C0
233%,200 10.060 242 7.800 c.C
3e TC10Z 43 5.000 S B- 0 0.0 0.0 ¢ Ds415E
REF 0 BLACHOT ETAL NUCL PHYS Al3G (196¢) P4324 GE DIOCCE

ENERGY INYENSTITY G ENERGY INTENSITY G ENERGY INTENSITY 5 ENERGY
“4E8 4000 15.000C 4TE.000 1CC.000 £28.000 15.000 1102.€00
11¢e.000 3.000

01 22 A

INTENSITY G
95.100
1.450
0,005
13.760
0.002

01 22 D

INTENSITY G
64000
25.000
25000
3.0C0
1.0C0

01 4 A
DATA
INTENSITY G
l1.000

01 30 R

INTENSITY G
17.000
.0
12,200
5. ‘006
T« 300
00
13.5C0

01 5 R

INTENSITY G
11.€00

TCO99M

TC-101

43 99

g9l
=7 PULG2 44 2%.6(0 D B~ grUl02 314610 le4CC 2 C.2C0E @1 7T A
rEF ~ NUCLEAE DATA TABLES SEC A 1970
tNeR oY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G
2%.559 D.08% 524110 Ce340 294,880 Calll “43,850 Ce3£C
L9 . 90T £9.0CC 556,900 0.8CC €10, 200 400
b RULCS 44 4e430 H F- RUL104 16,580 C.480 2 0.5C0GE €O 17 A
REF &2 GUNNINK ETAL UCID 1542¢ 14JAN 1S5¢5
ENERGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G
1254530 4,800 149.040 1.580 2624840 €.550 2164502 10,200
32€4100 1.070 3304500 D.72F 2FC.18C 1.220 393.32€0 2.8¢€C
“134E1% 2.200 46943280 17.5CC 499,289 24270 575.190 1.020
€EEo.154 1.920 6764320 18,0082 724,20C 44,5C0 875.8C0 24940
26T 4,400 2.000
g FPH102M 45 £7.0C0 M IT RU102 31,610 le4CO 2 0.0 1 A
kEF 0 NUCLEAF DATA TABLES SEC A 1970
ENERGY INTENSITY G
364550 C.MNiBE
<0 RH1GSM 45 45,000 S IT C 0.C 0.0 1 0.850E QO 1 D
KEF 17 LEDERER,HOLLANDER,PERLMAN TABLE OF ISOTOPES (1%568)
eNERGY INTENSITY G
125.430 100,200
41 RH105 45 1.475 D B- RULO4 l18.580 Ce480 2 0.900€E 00 6 A
REF 4z GUNNINK ETAL UCIC 15439 14JAN 1966
ENERGY INTENSIYY G ENERGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY &
212.€E00 0.113 280,540 0.180 306.310C Ee44C 319.240 194600

442.20C 0.037 427,000 0.05¢

LE107

RH105

45103

69T
42 FH106 45 367,000 D B- 0 0.C 0.0 1 0.2S0E 00 106 A

REF 13 STRUTZ,STRUTZ,AND HAMMERSFELD, ZePHYSIK 221 231(1969)

ENERGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G
22B.3200 0.000 427.900 0.092 435.C00 0,047 498,500 0004
511.800 2045C0 £44,C00 0.C03 £5249C0 0.C02 565,700 0.001
56G.200 04002 578,000 C.008 588.100 C.001 602.500 0,002
£1€£.200 0.810 621.800 9. TE0Q 6614260 C.015 680.100 G.007
684.30C 0.002 7164200 0.016 717.100 0.01% 749,500 0.002
774.600 0.006 873,1C0 0.414 1045, 70C 0.004 1056.,100 1,450

1062.100 0.024 1108.900 0.C05 1114.¢0C0 0.008 1128.000 . 0.383

11504100 0,000 1159.700 0.000 1179,100 0.014 1180.600 0,014

1194.300 01.053 1220. 400 0.000 1231.700 0.001 1257.000 0.001

126€.c00 0.001 13C5.6C0 C.003 1310.700C 0.0C3 1316.800 0.004

1355,000 0.001 1359,900 0.001 1397,.300 C.003 1457.800 0.001

14688, 5C0C 0.002 14964400 0.026 1562 .C00 Cal46 1574.400 0,001

12T7€.£20 0.002 1592, 700 0.008 16007060 C.008 160G5.700 0.001

1609.800C C.000 1686.100 0.001 1730.000 C.001  1733.100 0.002

1766.,200 0.023 1775,000 0.001 1784, €00 C. 00 1789.000 0.006

1796.70C 0.02% 1854,800 o.00¢® 19044000 0.000 1909.400 0.006

1926.8C0 7,013 1927, 200 0.0113 15504400 0.001 1956.900 G.001

1$88.500 D.027¢ 2014.,000 C.000 2033.90C C.0CO 2041.900 Q.000

2064.100 0.006 2169, 200 0.000 2193,200 C.00k 2194.000 C.005

221C.700 0.0C0 2230.400 0.000 22424400 0.001 2271.700 0.001

308,700 G.006 2309,400 C.006 2317.¢€00 0.006 23664400 G.021

22904800 0007 2406.C00 0.012 2438,500 0.004 2485,700 0.001

2£25.200 0N 2543,400 0.002 25714200 C.001 26514600 0.001

27064600 0.0032 2709.400 0.003 2740.,500 G.00C Z2783.500 0.0

£80%.50U 0.004 28214200 0.001 2504.100 €.00¢C 2917.9C0 0.001
2027.100 0.0 3036.8C0 0.001 3055460C C.0CC 3085.,900 0.0
2161.500 0.0 3169,000 C.0

4z KH107 45 4,200 M E- c C.0 0.0 1 C.190EF 0C g PD-107

KEF 12 GRIFFIN,PIERSON, BULL«AMusPHYS4SOCa13,NO.4 (196€8)

ENERGY IMTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G
112.600 0.0 202.700 0.0 312.200 C.0 348,200 Ne0
Z8l.70C 0.0 39244C0 0.0 56T7«6CC .0 670.000 0.C

L4 AGL1ICM 47 260,000 O EB- AG1OS 48.180 2,000 ¢C 3.0 15 A

REF 38 BEMIS,ETAL, NUCLe. PHYS Al2% (1969) NO 1 P217
ENEKGY INTENSITY G EMERGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G
a4€ . T70 3,200 6£204229 24740 6574710 C4,35C ETT7.550 11.510
tBELECD 6o 8N 7064 €R0 16,230 744,190 44150 763.280 224640
Bl8,.0C9 Te?€C 684,670 7.1C0 G37.480 244440 1284.220 2¢.120

1475,730 G250 15C4.200 12.260 15624220 1,250

0LT
&5 £Bi25

FEF ¢

ENERGY
35,450
116.379
178,78y
221.120
443,620
£J64 820

46 SB126

REF 42

ENERGY

c23.SC0
EEE.20C
20,000
674a9CC
£56. 700

“7 $Biz27

KREF “#2
tNERGY
252.70C
2G2.1C0
465,200
603,8C2
£84,900
753,470
216,400
1289.50C

«B S$8129

REF 0
ENERGY
18¢.0C0
523.000
760,000
967.000
1200.020
2070.000

4< TEL12TM

KEF ¢
ENERGY
58.C02

£l

NUCLFAR DATA
INTENSITY G

e A00
0e260
C.043
Oe44C
Ne280
52C0

51

GUNNINK ETAL
INTENSITY G

1.7¢0
1.480
1.15C
4e 320
17.600

51

GUNNTNK
INTENSITY G

6e7C0
l1.000
4+300
4. 4CC
36.800
D.0r7
0250
04340

51

INTENSITY G

24¢2C0
240090
1.600
5. 7CO
1,600
D800

52

INTENSITY G

0.010

FeGED C.100

INTENSITY G
D.068
C.200
C.l9C
C.240
C.250
l.800

C.0

INTENSITY G
4,TF0
CL00C
2e830

32.000
£.500

G.C

INTENSITY G
1.800
C.780
C.670
C.260
1.800
C.140
C.5CC

Cal

INTENSITY G
543C0
2.8C0
2900
1,060
4.100

18,710 0.100

3.210E-C1 z3 &
ENERGY INTENSITY G
111,060 C.0G8
176,299 €.3CC
227.<00 0.1C0
427,950 294600
6004770 15.400
0.320E-C1 1¢ A
ENERGY INTENSITY G
414,7C0 £1.000
£05.000 24480
6664200 100000
T2N.4€0 564000
0.137TE CO 20 A
ENERGY INTENSITY G
309.60C 0.210
440.700 CasT0
542,2C0 3.000
6664900 G.64C
7434700 0.150
8064000 C.C36
1140.3C0 C.3280
0.1C0E 01 21 ¢C
ENERGY INTENSITY G
412.0C0 2.000
682,000 54300
€16,.,000 17.000
1240,000 2.800
1840, 000 1.000
N.127E 0O 4 A

LEDERER, HOLLANDER GAMMA AB

INTENSITY G
C.3200

2.770 ¥ IN124
TABLES SEC A 1670
ENERGY INTENSITY G ENERGY
81,800 1.00G 109,270
122,430 0.03¢ 172. €GGC
204,070 0.250 208,120
380.¢°10 le4nC 408.1G0
463,510 10.000 489,800
€3€.15¢C 11.200 £T1.560
12.50C D 0 0.C
UCID 15429 14JAN 1969
ENERGY INTENSITY G ENERGY
278,00 242850 297200
57T3.70C T«300 5934CC0
639,009 1.350 €56,2C0
695,100 10C.000 587,00C
554,200 12.3200 $89. 500
3.500 O 0 0.0
ETAL UCID 15439 14JAN 19669
ENERGY INTENSITY G ENERGY
28C. 800 Ce4CO 220.60C
412,000 3.500 417.500
473.200 24.E00 501.9C0
638.20¢C 0.500 652. 700
£97,90¢C 24500 72144CC
782.600 5.0€C T87.900
829,100 0.249 924,200
1379,000 0.068
4,350 H ¢ 0.0
YOSHIHIRD TAGISHI JePHYS.SOC.JAPAN 21,2429(156¢€)
ENERGY INTENSTTY G ENERGY
296,C00 1.600 3584000
£44,00C 16.0C0 652,000
813,000 4C.S0C 8764 CCO
1028.000 10,000 1048.,000
1520.000 24400 1730.00C
169,000 D TElR2E€
ADAMS ,DAMS J RADICANAL CHEM.3,271(196°)DICDE
ENERGY INTENSITY G ENERGY
361,000 0.050 417,400

ENERGY
563,000

INTENSITY G

Cel

5zl2%

TELZ7

52129

LT
0 TE129 52 76,000 M B~ TEl28 31790 0.140 2 0.900E

REF 0 DICKINSON FT.AL, NUCLEAR PHYS A123 481-496(1969)

cNERGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G ENERGY
27,800 17.660 208,980 0.199 250, 65C C.430 270.300
2784430 0.614 281.160 0.161 3424600 C.008 342,800

459.6030 T.€80 4874390 le460 £31.830 0.062 551.500
559,7C0 1.008 624,400 0.092 7164800 0.002 T41.100
768,500 0.004 794,500 0.002 B02.170 C.209 8294900
833.4C0 0.045 g2%5.800 0.001 982.4C0 0.017 1003,4600

1C12.830 C.,001 10834990 0.606 1111, 740 C.238 1204,200

1233.000 J.009 1254.200 0.001 1260.840 0.012 1264,400

1282.100 N.001

5i TE129M 52 33.000 D B-IT TeEl28 31,750 0.017 2 0.350¢E

REF 0 DICKINSON FT.ALs NUCLEAR PHYS A123 4R81-496(198&9)

ENERGY INTENSITY G ENEKGY INTENSITY G ENERGY INTENSITY G ENERGY
105.500 C.230 £E564650 0176 €72.G30 c.028 695,980
701.800 0.029 705.600 c.00C8 729.620 1,170 T4l1.100
7684500 D.005 617.200 C.145 844,900 C.05¢ 1022.600

iC5C.«00 Ue028 1373,800 0.001 1401.60C C.0G7

5o TE131M 57 1,20 D 78~ TE13C 344,480 0.030 2 D.440E

REF b2 GUNNINK ETAL UCID 15439 L4JAN 1969

ENERGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G ENERGY
102.2G0 7.500 149.800 244200 182.4C0 2.100 200.700
240507 84.40C 334.300 11.1C0 4524400 te5C0 665.100
773,720 46.000 7824500 7.€00 753,80C 1%.900 822.800
€52.200 25.600 1125.500 14.8C0 1206.€00 11.8C0 164£.8C0
i687.7CO 1,70C 2001.000 2700 ‘

t= TE131 52 2%.0060 M B- TE13C  ®kdkddok*kx 0.2C0 2 De260E

REF C ADAMS ,DAMS J RADIOANAL CHEM.3,271(1969)DIODE LEDERER HOLLANDER

ENERGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY . ENERGY
149,7C0 69.000 343.000 0.C 384.000 0.0 4524400
+92,70C S.0C0 €02.10¢C 4.000 654.4C0 1.7C0 Q24,C00
€49.009 C.0 ©97.200 4.C0N loc8.C00 0.0 1147.800

Ha TE132 52 78,000 H B- 0 0.0 CaC 1 0.433E

REF n NUCLEAR DATA TABLES SEC A 1970
ENERGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G ENERGY

49,7293 1Zz.970 111.760C 1.80C 11643600 1.900 22841690

00 33 D 53129

INTENSITY G
0.005
0039
0.012
0.035
0.008
0.001
0.000
G.009

o 15 D 5312%

INTENSITY G
44510
0.045
0.031

Q0 18 A 1131

INTENSITY G
T«&00
5.000
6. 700
1.500

01 lz & 52131
GAMMA AR '
INTENSITY G
16,000
0.0
£.000

01 4 A 53132

INTENSITY G
ES.000

cLT
5E TE132M 52 55.40C M B- 1 0 G.0 0.0 1 N.65CE 01 £3 v TE-1332
kZF 2C MCTSAAC PHYS.REV VOL172 P1252 4 HEATH IN-1218
ENERGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSTITY G ENERGY INTENSITY G
T4..00 J.8CC 81.5G0 U.800 88,000 2.000 94,500 £.0C0
164,340 1.600 168,870 12.000 177.10C 1.5C0 178.2C0 1.000C
184,450 Fe400 193,220 l.2CC 138.200 C.600 213456C 44200
22Ue940 0500 2244030 C.400 244,280 Ca70GC 2514460 Ue €00
257.€40 i1.009 261, 550 14,CC0 234,140 1£.000C 344,500 1.500
247,220 1.300 255,570 0.609 262.81C l.10¢C 376.820 0.6C0
296496C 1.70C 429.020 3460G 444 .90C 44400 4624110 2.40C
471,850 2.30C 478+ €90 1.8C0 519.60C C.5C2 534,850 24060
5T4.C40 34600 622.030 1.600 6474400 24,000 7024752 443C0
T3i.590 1.700 T32.89%0 3.300 71¢. 750 2.90C 795,700 1l.50C
SuL. 510 2.200 863,910 29.000 982.8307 4. 800 BG7,700 C. 500
Sl2.530 1C0.000 Gl4.720 13,000 S34.40C 1,50C ©78.,199 %200
SEu.4L0 2.7CC 982.900 1,300 1026,.,800 l.800 1061.830 3.100
12438,200 20900 1459,100 2.5C0 1516.100 1100 i531.600 1.06C
1587.450 2ec!0Q 1683, 30¢C 6.7C0C 17C4,400 1.1C0 1855,700 1,300
c004.900 5.400 2027.700 1.400 204G, 200 1.7C0
e TEL133 52 12.400 M c 0.C 0.0 1 0.650E 01 31 A I-123
REF ie PARSA,GOGRDONsWALTERS, NUCL.PHYS.ALl10,€74 (1668)
eNERGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G
212,100 70,000 384, 600 0.400 292.900 C.8C0O 407.900 31.000
475,000 1.100Q 5464400 C.800 587.100 0. 7C0 613.6C0 0.400
€96+500 Do P 7204100 8.300 787,000 T.2C0 844,500 446C0
$3C.580 5.400 1000.5C0 4500 1021.0CC 24400 1¢61.000 1.800
1296.500 G.0 P 1252.1C0 1.4G0 1307.700 1.30C 1313.5C0 1.1G0
1332,300 11.090 1405, 6C0 0.800 1474.,00C0 " Qe.ECO 1518. 600 0.700
1588.200 C.400 1717.570 3.2G0 18254100 G.8CO 1881.450 1.3CC
2i34.500 0.400 2228,000 €400 2540,60C 0.100
57 TE134 52 41,800 M B- 0 0.0 C.0 1 0.4£90F 01 13 A 1-1324
KEF %1 TE-124 C.E. BEMIS,ET.ALe ARK. FYS. 37, 202 (1968)
ENERGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G
79.500 53.000 101.40C 0.500 181.100 22.000 201.500 9.400
213.800 25,000 278,100 21,000 434,800 18.000 4604700 Be9CO
46k 400 44300 565.600 15.0C0 712,500 £.100 742.000 14,000
766.700 27.000
568 I 131 53 8,060 D B- TEL30 34.450 0.240 2 0.291E 01 15 A 54131
KEF Q NUCLEAR DATA TABLES SEC A 1970
ENERGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G
80.164 2.450 163,570 0,023 177.230 C.270 272.300 0.070
284,312 5.8920 318.009 0.090 325.00C 0.033 225.750 C.280
238.500 0.01¢ 3644490 82.40G0 405,000 Ce. 066 5024940 0.330

€36.900 ¢£.9C0 €43.000 0,150 722.51C 1.620

eLT
£G 1 132

REF 39
ENERGY
147.200
2104500
4464000
£05.940
£21.000
670.000
780.200
Bl2.290
954,550
1096.800
1143.40G0
12©5,200
12724100
1476.800
1715.,400
1985,500
22234150

&0 1133

KEF 42
ENERGY
£t2.500
6174560
£20.500
123€e000

1896.%CG

1134

5460C

CRNL~-11C-4 GERRARD:+PINAJIAN,BAKER
INTENSITY G ENERGY INTENSITY G ENERGY
04240 255,000 G.150 2844600
0.160 363. €00 C.504 416.800
0+ 680 473.400 0.270 477.960
44960 522. 640 16,400 547,100
2.000 630,210 14,000 650.50C
4400 671,700 6.100 727.000
1200 784.500 0.300 792.100
5700 8764900 1.100 S10.300
17.900 5844500 0.730 1035.20¢C
0.03¢ 1112.500 0,063 1126,.,600
1.4CC 1148,2€0 0.210 1173.300
1.800 1298.000 0.77¢C 1314.300
2s+3CC 13984,570 64800 1442.560
0.130 1542, 200 0.010 1593,100
0.053 1727.,200 0.057 1757.500
N.008 2002,310 1.090 2CB86.840
Ne098 2249.000 0.030 2487T.600
53 20,900 H B- 0 0.0
GUNNINK ETAL UCID 15439 14JAN 1969
INTENSITY G ENERGY INTENSITY G ENERGY
04370 422,800 0.300 $10.530
D560 680,410 0.650 T06.71C
N«1320 8564470 1l.220 875.540
1.500 1298.400 24240
53 £3.000 M B~ ¢ 0.0
GUNNINK ETAL UCID 15439 14JAN 1969
INTENSITY G ENERGY INTENSITY G ENERGY
34260 139,100 C.650 162.470
1.740 4054 44C Te3EN 433,300
1.610 514.3E0 2.380 540,800
1C,. 900 628,000 24500 £77e340
3.830 847.030 96,000 857,280
4,000 9744630 4,930 1040,G00
S«150 1455.500 2960 1613.7CC

0.0

INTENSITY G
0.800
0.470
C.100
1.200
24500
64500
€.080
C. 906G
C.570
0.052
1.100
0. 060
1.420
04045
C.340
0.23C
0.C02

0.0

INTENSITY G
24000
1.530
4,430

0.0

INTENSITY G
0e260
44450
8.6C0
84200
€.600
2. 000
4,000

1

0s430E

ENERGY
306,600
431.900C
487.500
590,900
667,680
772.600
802,800
927.700

10864300

1136.020

1290, 800

1317.200

1456.500

16204600

1921.100

21724670

25254120

0. 669E

ENERGY
529.910
T68.5G0

1052.500

0.T780€F

ENERGY
128.420
4594000
59544C0
7304600
884,080

1072.530

1741400

01 68 A

INTENSITY &
0.110
0.460
0.180
0.060

99.220
75900
2.700
0.410
0.070
2.900
1,100
0.090C
0.050
€.030
1.200
0. 200
0.036

01 14 A

INTENSITY G
89.000
0.450
0.495

01 29 A

INTENSITY G
0. 680
l.430

11.200
2.20Q
66.000
14,300
3.100

XE133

7LT
bz 1158
REF 42
ENERGY
158,190
425,500
5454563
G72.21C
1131.570
12€7.72C
15¢L.600
1832¢,800

63 112¢

REF 22
ENERGY
187.7C0
381,7GQ
1213,.,20¢
2287.800
2576.800

53

GUNNINK ETAL

INTENSITY G
Ge 500
0,339
O« 400
8.400
24700

264800
Qs 750
16540
0700

53

LUNCAN,STIVOL A,

INTENSITY G
44200
8.2CH0

82.000
7400
1.800

o4 XEL131IM 54
NUCLEAR DATA TABLES SEC A 1970
ENERGY INTENSITY G

RE F 0

80.164

2.450

&5 XEL133M 54

REF 0
ENERGY
233,5C¢

¢é XE133

REF 0

ENERGY
79.550

14.000

54

ENERGY
163.970
2,260 D

5290 D

IT
Pe ALEXANGER NUCL PHYS
INTENSITY G

B

INTENSITY G

0,023

NUCLEAR DATA TABLES SEC A 1970

INTENSITY G

04040

67 XE13EM 54

REF 0
ENERGY
. 526,800

68 XEl35

KEF 23

ENERGY
156,500
IT:a10C
654,630

INTENSITY G
80,060

54

INTENSITY G
D262
0.012
De0?25

ENERGY
80.995

15,60C M
ADAMS DAMS,J«RADIDANAL.CHEM,

9,100 H
ALEXANDER,

ENERGY
199,900
408,200
721.9C0

IT

B-

INTENSITY G
364600

0.020
Ce339
0.0%50

XEl32
A121 612 (1968)

XE132

XE1l34
3,271(1969)

XEl>4
LAU NUCL «PHYS.A121,612
INTENSETY G

6.700 H B~ ¢ 0.0
UCID 15439 14JAN 1966

ENEPGY INTENSITY G ENERGY
2204400 <¢20C 229.€8C
408,000 O.58C 414.800
433,500 0,600 453.60C
707.900 0.850 769,350
1038.810 1C.0C0 1101.580C
1165.100 1,290 1240.,400
1368€£4900 C.720 1457.,¢10
16784260 11.800 17C6,70C0
1927.300 0370 245,000
83.0C0 S B- c 0.0

ANNe ACAD, SCI. FENNe,

ENERGY INTENSITY G ENERGY
219.700 0.600 345,500
295,800 1.000 1246.400
1320.,200 18,009 1767.,200
2414.,000 2.800 2631.800
32039.500 1.200 3153.70C

11,900 O IT XELr30

ENERGY
1€0.5Q0

(1968
ENERGY
249,650
573.200
812.€00

C.0

INTENSITY G
0.250
C.3¢0
Ce320
C.0 P
1.8C0
1.020

1C.CCC
4.50C
C.900

G.C

SER. A VI,NO.

INTENSITY G
Z2.100
34300
l.4C0
4.20G
1,200

4,080 £.000

26.890 0.022

2€.890 0.200

INTENSITY G
0.051

10440 0.003

104440 54200

INTENSITY G

G2.0C00

0.055

0.617E C1 38 A
ENERGY INTENSITY G
288, 3€0 44 00C
417,660 4. 1CC
526.540 1£,409
836,880 8.000
1124.000 49200
12604500 34,900
1502.800 14240
1791.400 Se4C0

0.640E 01 20 R

2€88,1-15{(1368).

ENERGY INTENSITY G
371,200 241G0
12568.600 Te 200
1¢35,0C0 1.4C0
28564500 24700
3238.000 1.700
0e22CE~01 2 A
0.160E €O 1 a
D.6€69E 01 3 A
0.180F Q1 1 A
0.6405E 01 12 A
ENERGY INTENSITY G
258, 600 0e239
€084 600 20630
1043,000 0.003

XE135

XE133

54135

€S5-135

GLT
69 XEL137
REF 2%
ENERGY
394,000
849,000
969.000
1117.000
1230.000
1657.000
15164000
2E52.000
3914.,000

70 XE128

REF 28
ENERGY
153,200
401.600
2013.000

71 XE136

REF 29
ENERGY
121.000
<B85.900
294,200
549,000
722.000C

Te XElaC
KEF 27
ENERGY
474300
123.00C
162,600
202.600
281.2C0
260.100
433.,€00
£03,.200
€57.200
€35.900
205.40"
1136.30C
14124100

72 CSiz24M

REF 43
ENERGY
104500

01 32

INTENSITY
04020
0.070
0.070
0.250
0.010
0.480
0.100
0,010

01l 9

INTENSITY
16,000
14.0C0

01 l¢g

INTENSITY
2.000
0.500
24100
0.900

o1 50

INTENSITY
1.000
0.400
44900
44200
44000
44000
G.700
64600
3.900

21.3C0
164200
34, 10C

)

54 4200 M B- XE13¢ 8.870 0.280 2 0.600E
GeHOLM,y, ARKIV FYSIK 37 1{(19¢8)

INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G ENERGY
0.170 455,600 30.000 595.0040 0.090 684,900
04650 B75.000 0.010 886.000 C.020 934,000
0.030 S82.000 0.220 1068.C00 C.080 1107.000
0.200 1185,000 0.080 1197.000 0.070 1275.000
0.020 1576. 000 O.170 1615.00C 0.160 1635,000
0.040 1668,000 c.080 1686.000 0.040 1784.000
0.110 2004.000 0.020 2017.000 0.040 2396,000
0«260 2524.000 0.0320 3162.000 0.010 3697.,000
0.0G10

54 17,000 M B~ 0 0.0 0.0 1 0.590E

XE138, NAGAHARA ET, AL, Jo PHYS SOC JAPAN 26, 232 (1969)

INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G ENERGY

26.000 242,800 S.600 258.200 1C0. 000 396.50C0
9.000 4344400 48.00C 1769.000 36,000 2002.000
29,000

54 41,000 § B~ G 0.0 0.0 1 0.540E

XE139, G.HOLM ET4 AL+ ARKIV FYSIK 34, 447 {1667)

INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G ENERGY
1,000 174.900 29,000 219.0600 77.000 225.500
10.0G0 29¢.700 24.000 339.000 0.7C0 358.0C0
8,000 456,000 0.1004 491,000 1.400 514.000
0.6C0 613,000 4400 649,000 0.0 723.000
1.200 788,000 34200

54 14,300 S B~ c G.0 0.0 1 0.380€
ALVAGER PHYS REV 167,1105(1968)

INTENSITY G ENERGY INTERSITY G ENERGY INTENSITY G ENERGY
5.200 719,400 224400 874700 1,900 92.£00
11.700 111.¢400C 18.30¢C 117.5CC 21.6C0 158.600
1.4C0 166.70C 5.900 182.000 1.300 197.6C0
C.800 212.200 14,400 258.400 l.2CGC 277.400
10.700 290. 760 3.400 331,200 2.000 373.500
9.2C0 396. 500 44700 411.€C0 C.50C 42%9.400
19.220 445,100 4.200 454,500 0.5C0 461700
2,000 514,800 5700 518,900 €.200 547.800
28B.7C0 £C8.100 13.80C 622.0CC 40.3C0 £27.400
T420C 646.,100 0.8C0 6534400 214400 T73.900

10C.0N0 879,500 11.600 ¢51.200 3.0CD 588.900
T«8C0 1207.400 5.50¢C 1308. 80C 3C.CCC 1314.600
51.200 142¢6.,0C0 2,700

55 2,500 H IT €s132 1C0.0C0 2.€C0 C C.0

ADAMS AND [AMS J,RACIOANAL CHEM —-ENERGIES BLUE BOOK FE2 OTHER LCATA

INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G
0.0 127.400 14,000 137.,40C 0.¢

A CS-137

G

R CS138

G

DO (s1z2¢

G

R CSls&0

G

9LT
T4 513«

REF 42
ENERGY
475,342
7954790
1365,120

7t €s1z27

REF o
ENERGY
6€1 4640

7€ CS13s

REF 27
ENERGY
138.300
402600
14N8.4049

77 CS1zs

REF 27
ENERGY
10l.60¢C
1107.400

78 C5140

REF 27
ENERGY
528.200
820.7C0
1075500
1221.200
1408.200
1852,6C0
2237.400
£522.400
305%.1090

7¢ BA133

REF 23
ENERGY

52.,17¢

223.430
384,090

80 BALZTM

REF 43
ENERGY

661 .60C

ADAMS AND DAMS J.KADIOANAL CHEM —ENERGIES
INTENSITY G
89.000

55 2.100 Y B- €s13z2 160.C00 320.0CC 0.C 10 A
GUNNINK ETAL UCID 15439 14JAN 196¢
INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G
1.540 563,220 8.8290 569,320 15.8C0 £04,700 38,0600
69,0060 801.87C 3.500 1028.,4610 1.0&0 1167.210 1.850
2,000 1400.5GC0 ¢.080
5¢ 30,000 Yy B~ 0 0.0 Ce 0 Q.620F C1 1 A 56137
NUCLEAR DATA TABLES SEC A 1¢70
INTENSITY G
84,6C0
55 32,260 ™ B~ 0 CuD 0.0 D.670E 01 12 A
ALVAGER ETAL PHYS.REV.167, 1105 (1968) MASS+SPEC.SEPERATED
INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G ENEPGY INTENSITY G
2.0C0 192,500 0.800 227,900 1.€00 409,100 3.000
23.000 546.6C0 8.00¢C 1464200 £.000 870.700 44000
25.000 1426.000 73.,C00 2217.0C0 18.000 2630.,000 9.0C0
55 S.500 M B~ o C.C 0.0 0.647E 01 & R BAlZe
ALVAGER ET., AL, PHYS REV 167, 1105 (1968) MASS SEPARATED

INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G

C.400 6264600 5.0C0 7244060 24400 732.400 3.1C0

13.600 1284,00C 84800

55 65.700 S B- o 0.0 0.0 0.5¢0E 01 33 R BAl4C
Te ALVAGER ET AL PHYS REV 167,1105(19¢€8)

INTENSITY G ENERGY INTENSETY G ENERGY INTENSITY G ENERGY INTENSITY G
5,800 602,200 1¢0.000 ET1,9500 244C0 735,500 1,000
0.0 908, 20C 16,000 $25,.,000 0.0 10084500 2.000
0.0 1130.3C0 550C 1200.000 6.7C0 1215.500 0.0
44200 1246.200 1.200 13¢8,800 C.0 1290.800C 24400
0.0 1499,5C0 6.100 1633.800 2.4C0 1827.CCO 24500
8.100 1948,200 20090 2031,000 0.0 2099,.C00 4700
3.100 23294600 3.600 2348.400 0.0 2429.400 44600
5.000 2580.807 OeC 27724700 0.0 2924.300 0.0
C.0

56 10.660 Y EC Bal32 0.0 0.0 D40 9 R
ALEXANDERyLAU NUCL.PHYS,A121,£612 (1968)

INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G
3.300 794600 11.000 80.997 5540C0 1604660 1.200
0.740 276.430 12.00202 303.090 3Cs6C0 356,260 10C.CCO

14.200
te 2+.554 M 17 BAl3¢ 7.810 0.01C 0.,580E 01 1 A

BLUE BOOK FCR OTHER DATA

LLT
8l BAl29 56 83,000 M B- BAl38 Tl.6€0 0,400 2 O.649E

REF ra Te ALVAGER ET AL PHYS REV 167,1105(1968) HOLLANDER FOR HIGHER
ENERGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G ENERGY
165,700 22.000 1090.000 C.120 1270.000 0.200 1430.000
82 BAl40 56 12.800 0O B- G 0.0 0.0 1 0.630E
REF 0 NUCLEAR DATA TABLES SEC A 1970
ENERGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G ENERGY
13.850 1.290 29.960 14,000 34,150 1.240 118,9C0
132.7C9 0.170C 144,000 C.040 162.900 6,200 177.000
304.€20 4.500 4234690 34200 437.550 2,100 466.000
498.000 0.400 512.000 C.260 537.380 23.800 £02.000
627,000 03060 661,000 0.700
g3 BAl4l 56 18,000 M B~ 0 0.C 0.0 1 0.610E
REF 0 ADAMS UCAMS J.RADIOANALYTICAL CHEM 3,271 (1969} DIODE
ENERGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G ENERGY
18%.800 100.000 2764900 5C.000 304.400 60,000 344,000
457,00 30.000 4624500 20.000 46T7.0CC C.0 647.900
738.000 0.0 1197.000 0.0
b4 £ A14C 57 40,200 H B~ LAL3S $9.950 8,900 2 O.634F
REF Q NUCLEAR DATA TABLES SEC A 1970
ENERGY INTENSIYY G ENERGY INTENSITY G ENERGY INTENSITY G ENERGY
24,4595 0.0 68,916 0.060 109.417 0.220 131.121
173,543 0.140 241,960 C.€00 266550 0.700 306.900
228,770 21.00C 297, 800 0.100 432.550 343C0 487,030
6lz.t00 0.040 751.790 44400 815,830 234100 B6T7.840
€1%.509 24500 625.230 €.9C0 55G. 500 0. 640 1596, 5600
25484400 Ne.B60 2822.00¢0 3.300 2847.500 0.110 2899,7CO
3118.500 0.027
o5 LALG]L 57 3,900 H B~ 0 G0 0.0 1 0.610E
REF 42 GUNNINK ETAL UCID 15432S 14JAN 1969
ceNERGY INTENSITY G
1275.900 2.0C0
56 LAl&4Z 57 1.400 H B~ c 0.0 C.0 1 0.584E
REF 42 GUNNINK ETAL UCID 15439 14JAN 1569
ENERGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G ENERGY
CETTeGED 1.229 641,210 46,500 8&la €40 1,770 87G.480
BG4 4, 50 8.50C 1011.48¢0 2.540. 1043.800 24670 11604200
1165%5,200 1,580 1233,000 2.300 1354.600 2430 1263.100
1220.300C 6.8E0 154¢€¢.100 24520 1722.500 l.,€00 175645CC
1G21.407 645C0 1949,300 40730 2055.3C0 24450 21874300
22%1.500C 24140 2398.,000 11,700 25424500 E.800 25604.400

z€114230 24550 26674100 l.610 25T71.300 24740

01 4 R
ENEGIES
INTENSITY G
1.800

0l 18 A

INTENSITY G
0.048
0.190
0.210
0.600

01 10 R

INTENSITY G
20,000
10.0G0

ol 25 A

INTENSITY G
0.800
04060

454000
5.500
35,600
0.070

01 1 2

01 2T A

INTENSITY G
1.17¢
1.650
24020
24690
4.050
1a860

57140

57141

CE-14!

CElaz

+

QLT
87 CElal 58 32.380 D B~ CElaq 884480 O«.£CC C D.£10E 01 1 A
REF 0 NUCLEAR DATA TABLES SEC A 1970

ENERGY INTENSITY G
145,450 49.0C0
ge Celas 58 l.370 D B~- CEla2 11,070 1.000 2 0.591E 01 10 A PR1432
REF 42 GUNNINK ETAL UCID 15439 14JAN 1969
ENERGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G
5T.400 12.500 231.600 24130 293,200 46,4500 3504600 3550
432.800 0.180 490.400 2.210 5687.400 C.220 6644590 &.000
7214980 +« 600 880,400 1.100
g9 CEiss 58 284,000 O B~ c 0.0 0.0 1 0.541€ 01 10 A 59144
REF 0 NUCLEAR DATA TABLES SEC A 1970
ENERLY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G
32,870 0.220 4C.930 Q.390 £3.410C Oe 140 59.030 0.0
80.120 14540 96,950 C.038 133,520 1C.,8C0 696,480 1.51C
1469.140 0.250 2185,720 Ce 740
G0 RAZ26 88 1620.000 Y c 0.0 0.0 4 0.0 66 A
FEF 41 ADOPTED VALUES BY ORNL NUCLEAR DATA CENTER RA-226 IN EQUIL W DAU
ENERCY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G
464510 0.0 53.230 24200 186.000 3.900 241,920 Te260
258.800 0.5%0 274.800 C.47C 295.220 17.£C0 351.990 34,600
285.800 0.360 388,900 Ce 350 405,900 C.15C 455, G00 0.280
462,100 5.210 470000 Cs130 480,500 0,320 48T.2G0 C«400
533.800 0170 5804300 C.320 60G.400 424200 665,500 1.500
702.1C0 0.¢70 718.800 0.410 753.00C 0,110 768.400 44950
786.000 0.8¢0 786,100 C.290 806.200 14150 52143200 0.160
EZ€.0200 0.13% 836,106 C.£80 234,100 2,000 S€4.200 . 380
LGB2.000 D240 1070.000 0,260 11¢4,000 Cel6C 11204400 13.70C
1123.800 0.250 11£5.3C0 1.700 1207.7°C 04506 1238,200 6+ 300
12814102 1.600 - 1303,.800 0,140 1377.8CC 44,200 1285.400 Ue92C
1401.7C0 14500 1408.0CC 2.4C0 1416.,1C00 Ce0 15209.5C0 2.10C
1£432,400 Co26C 1582,500 $a720 1594,90C Ca250C 1599,6CC Ce370
1EA1.500 1.120 1684,200 Ca250 1729.9CC 34100 1764,700 164000
1236,.,4C0 0.3572 1847.70C 2.100 1£72,€00 C.240 123904 600 €a110
18964800 0.190 2110.4CC C.100C 2118.20C l.28¢C 22044500 £.300

22593, 7C0 04350 2448.000 1740

6LT
“1 TH252

REF n

ENERGY
3G.43F0
Ba o500
121.,6C0
176.00C
232345350
270,00
200.€10
09,700
t8%el40
785.“'{.3
C‘:’Bo‘fjo
1C79.100
1£13,100
1808.0060

S FAZ33

FEF 432

ENEKGY
15280
1G2.8560
271 540
275.400

Gz U 232

REF 4
ENERGY
1158.0C0
511.C00
2410.000

94 u 2
REF 4
ENERGY
124.000
154,000
211.4600

NUCLEAR DATA TABLES SEC A 197G
INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY EMERGY
l.060 57.300 0.0 F3.000 0.0 78. 100
1.200 98.000 0.0 115,190 C.570 128,000
0e1RQ 181. 600 0.£00 1664500 C.180 1764700
0.0 184.C00 1.600Q 2084 00C 2.700 216,100
« 120 238,620 40.000 2404980 441CC 2524600
4,000 27174240 2+40C 282.00C 0.300 288.1C0
3,000 327.000 1.900 328, €00 0,172 137.000
1.600 452.800 0.400 461,000 2.700 51D.720
2(.C00 7274200 7.10C0 162,30C C.E1N 7804009
1.000 793, G00 Ze500 £33.000 3.200 350100
0.400 909.0CH AC.000 553.00C 0.110 96T.0C0
Qe 6CC 063,960 C.150C 14664000 1.30¢C 100,000
Ce?1C 1595.,000 5.600C 1620, 800 l.8CC 1£36,000
0,110 2E14,.660 28,930
91 27.400 © B- TH232 10C.0C0 T.400 O 0.C
GUNNINK ETAL UCID 15439 14JAN 1669
INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G ENERGY
0.680 B6.61D0 1.860 94, 66C Be&430 984440
0.810 111030 4,850 1144620 l.680 2484300
0,33C 200.110 6.6C0 211.900 38,000 240,470
D« 655 398. 500 12,900 415,750 14720
G2 T2.000 Y B- o 0.0 0.0 4 0.0
HERTZ MLM-1400 RADIATIONS EMITYED BY U-333,U-232 MIXTURES
INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G ENERGY
44500 239.000 81.0C0 277.000 2400 3200.000
T.800 583, C00 29.000 727.C00 4. 500 860.000
5.500 2614.470 3¢€.,000
92 162000.000 Y B- c 0.0 0.0 4 0.0
HERTZ MLM-1400 RADJATIONS EMITTED BY U=-333,U-232 MIXTURES
INTENSITY G ENERGY INTENSITY G ENERGY INTENSITY G ENERGY
20,000 132,000 S.000 137.000 15,100 143,000
20.500 156.000 11.200 180.,00C 1.80C 193,000
9,000 164000 15.060 437.000 0.0

54 A

INTENSITY G
0.0
1.600
0.200
G.400
0.270
0.450

11.000
8,000
1.40C
49400

27.060
2.100
34200

15 A

INTEHSITY 6
13.500
0.075
44430

08T

10 A +

INTENSITY G
44500
44.1C0

11 A +

INTENSITY 6
7.500
14,600
181

Appendix B

PRODUCTION AND STANDARDIZATION OF '*°MAg IN SILVER TUBING
FOR MSRE GAMMA SPECTROMETRY

Silver tubing was irradiated in the Oak Ridge Research Reactor (ORR)
to produce a cylindrical source (6.1 in. long, 0.5 in. OD, and 0.035 in.
wall thickness) of about 24 Ci of **°"Ag., Gamma spectra of the source,
placed in a mockup of the MSRE heat exchanger, were made to obtain a curve
of counting efficiency versus photon energy for resolution of spectra of
fission products in MSRE components. Since ''°"Ag has several peaks in
the energy range from 446 to 1562 keV of well-known energy and abundance,
this source is very suitable for efficiency calibrations.

Irradiations were made in hydraulic tubes of the ORR. Because of the
large effect of the silver tubing on the reactivity of the ORR, only 2.3 in.
of tubing could be irradiated in a single hydraulic tube at any one time.
It was estimated that rapid withdrawal of a longer section of tubing would
scram the reactor. No provision was available for slow withdrawal., Since
a source at least 6 in. long was desired, it was decided to expedite the
source production by irradiating three 2.3-in. sections in three separate
hydraulic tubes (12, 13, and 25). To ensure that the silver tube sections
were exposed to equal neutron fluences and that the neutron flux gradients
were not excessive in the hydraulic tubes, Co-Al flux-monitor foils were
irradiated first, and results obtained from measurements on these were used
to adjust the irradiation times of the silver tubes for equal neutron flu-
ences., Details of all measurements and source preparation and evaluation

are given below.

Flux Monitors

Flux monitors were prepared by cutting 3/8- by 2-in. sections from a
0.01-in. foil of cobalt-aluminum alloy containing 0.14%Z cobalt. The foils
were weld-sealed in aluminum rabbits and irradiated 20 min in hydraulic
tubes 12, 13, and 25 on May 6, 1969. After irradiation, the foils were
weighed and their °°Co contents measured by gamma spectrometry using a

Nal detector. The total areas of both the 1.17- and 1.33-MeV peaks were
182

taken as a measure of the °°Co content. The final results, which are given
in Table B.l, were reduced to counts per milligram of foil. The values in
the table were used to normalize the irradiation times of the silver tube

sections to expose the tubes to equal neutron fluences.

Table B.1l. Results of count-rate measurements
for total Co-Al flux monitors

 

Sample Weight

No. (mg) Counts Counts/mg
D-25 96. 4 56,439 585.5
D-12 95.7 46,771 488.7
D~-13 103.5 48,484 468.4

 

To obtain a measure of the neutron flux gradient in the hydraulic
tubes, sections of about 1/4 in., were cut from the ends and middle of each
foil, weighed, and assayed for °°Co by gamma spectrometry. The count rates
of each foil were compared with the count rate of a standard ¢°Co source
to permit calculation of their disintegration rates and the neutron fluxes
to which the foil sections were exposed. These results are given in Table

B,2.

Silver Tube Irradiation

 

From previous experience it was estimated that the activity of the
source tube should be in the order of 3.5 to 4.5 Ci of *'*°"Ag per inch
of tube.

The amount of **°™Ag, A, that would be formed in 1 in. of silver
tubing after one week of exposure to a flux of 2 x 10'* neutrons cm—?

sec™' was calculated by
183

Table B.2. Measured results for foil sections
cut from Co-Al flux monitors

 

 

Sample Flux
No. Counts/mg/sec Disintegrations/mg/sec  (neutrons cm—2? sec—?')

D-25-1 7.45 720 2.70 x 10t*®
D-25=-2 7.20 696 2.62 x 10**
D-25-3 6.86 663 2.49 x 10**
Mean 7.17 693 2.60 x 10**
D-12-1 6.15 595 2.24 x 10"
D-12-2 6.10 590 2.22 x 10**
D-12-3 6.11 591 2.22 x 10**
Mean 6.12 592 2.23 x 10"
D-13-1 5.68 549 2.06 x 10**
D-13-2 5.94 575 2.16 x 10**
D-13-3 6.14 594 2.23 x 10**

Mean 5.92 573 2.15 x 10**

 
184

W (AB) (AVY F S Ho
4 = Ci,
100 (4W) x 3.7 x 10*°

 

where

W = weight of 1 in. of tubing, 8.80 g,

AW = atomic weight of silver, 107.87,

AB = natural abundance of '°®Ag, 48.18%,

AV = 6.023 x 1023,
F = neutron flux, 2 x 10'%,
o = saturation factor, 0.693 x 7 days/260 days = 0.0186,
H = the neutron self-shielding factor, assumed value = (.6.

Using the values in the equation it was predicted that 4.28 Ci of
119MAg per inch of tube would be formed. Therefore it was decided that
the irradiation in hydraulic tube 25 would be made for 5.00 days and those

in hydraulic tubes 12 and 13 for

585.5

 

 

5 x 788 7 5.99 days
and
588.5
5 x L68.5 6.28 days,

respectively, which would then give an activity of approximately 4.3 Ci
per inch of tube., The normalization factors were obtained from Table B.l.
The silver tube sections were prepared 2.3 in. long with a 0.25-in,
center section that could be removed for destructive assay of the **%"Ag,
Silver caps were placed on the ends of the tubes to prevent neutrons from
entering and causing increased activation of the ends of the tubes. The
tubes were weld-sealed in aluminum rabbits and irradiated according to the

schedule in Table B.3.
185

Table B.3, Irradiation schedule for silver tubes

 

 

 

Hydraulic Dates and times
tube No. In Out AT (hr)
25 5/7/69, 0819 5/12/69, 0855 120.5
12 5/7/69, 0826 5/13/69, 0845 144.3
13 5/7/69, 0830 5/13/69, 1455 150.4

 

Measurement of '*°7Ag in Silver Tube Sections

After irradiation, the rabbits containing the silver tubes were trans-
ferred to a hot cell and loaded into the source holder shown in Fig. 5.1.
The silver tubes irradiated in hydraulic tubes 12, 13, and 25 will be de-
noted as samples 12, 13, and 25 respectively. Because it was desirable
to maintain the length of the final source as long as possible, the center
sections of only two tubes were removed. Since the flux gradient in hy-
draulic tube 12 was the smallest, the center section of this silver tube
was left in place. The order of placement of the silver tubes in the
source holder was 13, 12, and 25,

The center section of silver tubes 13 and 25 were dissolved in separate
l-liter volumetric flasks containing nitric acid. The solutions were di-
luted with distilled water and mixed well, and aliquots of 10 ml were taken
from each, placed in other l-liter flasks, and diluted to capacity. From
these solutions aliquots of 5 ml were taken for analysis. From each of
these final solutions, three aliquots of 0.1 ml were assayed for *'°"Ag by
gamma spectrometry using a Ge(Li) detector. The peak area of the 657-keV

gamma ray of *'°"Ag was measured. The photon emission rates of the '*°"Ag
186

samples were obtained by comparing their count rates with the count rate

i37

of a Cs source of known photon emission rate. The results of these

measurements are given in Table B.4.

Table B.4. Measured and calculated quantities for
silver tube samples

 

 

Sample Counts/sec Aliquot Curies
No. per 0,1 ml volume® (ml) Gammas/sec per inch of tube
13~1 5.835
=2 5.961
-3 5.851
Mean 5.882 9.9 3.336 x lO“b 3.861
25-1 6.457
-2 6.355
-3 6.064
Mean 6.329 9.8 3.591 x lO“b 4,198

 

aAliquot taken from first liter of solution in which silver tube
sections were dissolved.

b ,
Calculations based on count rate of 26.6 counts/sec for *2?7Cs source

with a photon emission rate of 1,526 x 10° gammas/sec. The values were
corrected for decay to June 1, 1969.
187

The photon-emission rates of Table B.4 were calculated by

G
4 = =20,
Cs
where
4 = gammas/sec of 0.1-ml aliquot of '*°Ag solution,
cc - &ammas/sec of 137Cs source, 1.526 x 10°,
Cog = counts/sec of **®7Cs source, 26.6,
¢ = counts/sec of 0.1-ml aliquot of **°"Ag solution,
D = decay factor for **°"Ag for May 27, 69, to June 1, 69 = 0.989.

Values for curies of **°"Ag per inch of silver tube given in Table B.4

were calculated by

A x 4 x 10°

 

Ci/in, =
0.1V (y/DY 3.7 x 10*°

where the factor 4 converts the results from 1/4 to 1 in., of tubing, the
factor 10° is a dilution factor for the two l-liter dilutions, the factor
0.1 is the volume in milliliters of solution assayed, the factor 3.7 x 10'°

converts from disintegrations per second to curies,

A = gammas/sec for the 0.l-ml aliquot,
v/D = branching ratio of the 657-keV gamma ray = 0.9435,
V = aliquot volume, in ml, taken from liter of solution in which

silver sample was dissolved and added to second l-liter flask,

9.9 ml for sample 13 and 9.8 ml for sample 25.

<
I
188

AEEendix C

ABSOLUTE EFFICIENCY CURVES OF THE GAMMA-RAY DETECTION SYSTEM
USED FOR THE CALCULATION OF ABSOLUTE AMOUNTS OF NUCLIDES DEPOSITED

We intended to keep the detector system counting dead time more or
less constant (between 15 and 257%). Since the intensity of radiation from
the various measured components differed and varied with time, we adjusted
the shielding configurations to obtain a relatively constant count rate.
Tables C.1l and C.2 present the various shielding configurations employed
in surveying the primary heat exchanger and the main reactor off-gas line
(jumper line). Figures C.1l to C.3 present the efficiency curves used in

the analysis of the spectra.
189

Table C.1. Shielding materials employed during survey
of the main reactor off-gas line

 

Diameter of

 

Shielding collimator insert Shielding material
case (in.)
1 1/16 1/8 in. steel
2 1/8 1/8 in. steel, 1 in. Cu
3 1/16 1/8 in. steel, 1 in. Cu,
2 in. paraffin, 1/8 in. Cd
4 1/8 1 in. Al
5 1/8 None
6 1/16 None
7 1/8 1/8 in. Cd
8 1/8 1/8 in., Cd, 1/8 in. steel
9 1/16 1/8 in., Cd, 1/8 in. steel
10 1/16 1/8 in. Cd, 1/8 in. steel,
2 in. paraffin, 1/2 in. Pb
11 1/16 1/8 in. Cd, 1/8 in. steel,
2 in. paraffin, 1/2 in. Pb
12 1/16 1/8 in. Cd, 1/8 in. steel,

2 in, paraffin

 
190

Table C.2. Shielding materials employed during survey
of primary heat exchanger

 

Diameter of
Shielding case collimator Shielding material
insert (in.)

 

1 1/8 None

2 1/8 1 in. Al

3 1/8 1/2 in, Al, 1/2 in. Cu

4 1/8 1 in. Al, 1/2 in. Cu

5 1/16 None

6 1/16 1/8 in. Cd

7 1/8 1/8 in., Cd

8 1/8 1/8 in. Cd, 1/8 in. steel

 
191

ORNL-DWG 70-9608

 

 

 

 

  
  

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

-8
{0 e - -
o 4 +—1r = =
R S, — COUNTS REGISTERED _ .
' DETECTOR SYSTEM EFFICIENCY = 5uoroRc EMITTED PER IN. OF TUBE  {-— - -
5 . B S
]
2
10°° | — - .
L . - + 1
S e Lo - 4 - 1
5 P i L _
/ . \ . _
S S b ] i : y .
| T~——__CASE 9 |
2 e .hi?f o
\o\% P——
5107 - = — L —
@ - - : T CASE T T = —
9 T 7_ / ] 1 TTTmTrTT 4 i - 1 e
o /S CASE 10 P ]
p—
i; o . . ’/ . T _ 3. . _
T T —
o ol . ,/ CASE NO.  COLLIMATOR SHIELDING MATERIAL B
S DIAMETER
5 (in.)
: I
g*dfl,, — — 7 g Ygin. Cd ]
"‘“4;:j#f T I s Ya Ygin. Cd + g in. STEEL —
S et~ 9 Yig Ygin. Cd +Ygin. STEEL -
R o {0 Vig  Ygin. Cd+Ygin. STEEL+2in.Pat¥%2in.Pb
£ 1/16 '/8 in. Cd +‘/8 in. STEEL+2in. Pa +'/4 in, Pb
T a2 Vi Ygin. Cd + Vg in. STEEL+ 2in. Pa |
L T i ! - ]
e T T T | i
T - -
S I ] i
—t - +
t f
5 i
| i
‘0—13

100 300 500 700 900 100 1300 {500 {1700 4900 2100 2300 2500 2700 2900
PHOTON ENERGY (kev)

Fig. C.1. Absolute efficiency of gamma-ray detection system for
main reactor off-gas line.
192

ORNL-DWG 70-9607

 

 

 

 

10
: COUNTS REGISTERED
5 N ' DETECTOR SYSTEM EFFICIENCY = PHOTONS EMITTED PER [IN. OF TUBE
2
1078
5
2
5 o107°
P
E
o
i
L 5
=
w
-
w
)_
w
o 2
O
-
Q
E o
i 40
5
2 e 5 ~ CASE NO. COLLIMATOR SHIELDING MATERIAL
; DIAMETER
{in.}
=41
10 Yhe Yain. STEEL
- ) ) g tin.Cu+1gin STEEL
5 : an Vig fin.Cu+2in. Pa+gin. Cd+ Ygin STEEL
1/8 fin. Al
Vg NONE
2 o Yie NONE
10—12

100 300 500 700 9C0o 1400 4300 t500 1700 4900 2100 2300 2500 2700 2900
PHOTON ENERGY (kev)

Fig. C.2. Absoclute efficiency of gamma-ray detection system for
main reactor off-gas line.
DETECTOR SYSTEM EFFICIENCY

-6 ORNL-DWG 70-9609

COUNTS REGISTERED - —-
"DETECTOR SYSTEM EFFICIENCY = b

PHOTONS EMITTED PER c¢m® OF HEAT EXCHANGER TUBE M————;—CASE 7

 

CASE . CASE 1

 

 

 

1077
5
> R i .
| | | [ i | |
_8 CASE NO.  COLLIMATOR SHIELDING MATERIAL
10 DIAMETER
{in.}
NONE
> {in, Al
Yoin.Cu + 5 in. Al
; Yoin.Cu+1in. Al
2 e NONE
Ygin. STEEL
"/ain. Cd
-9 . R
10 Yg in. STEEL + g in. Cd
5

 

100 300 500 700 900 1100 1300 1500 1700 {900 2100 2300 2500 2700 2900
PHOTON ENERGY (kev)

Fig. C.3. Absolute efficiency of gamma-ray detection system of
primary heat exchanger.

€61
194

Appendix D

CALCULATIONS OF COUNTING EFFICIENCIES FROM FIRST PRINCIPLES

Objective

We have computed from first principles counting efficiencies for the
MSRE main off-gas line (see Sect., 5.4) and the MSRE heat exchanger (see
Sect. 5.5). Initially we believed that such calculations could be done by
models that were of necessity so simplified that only gross agreement be-
tween theory and experiment would be achieved. However, this has not been
the case; very close agreement has been found over a wide gamma-energy
range, Although significant differences still exist below 200 keV and
above 1500 keV (precbably still due to the simplified calculational model),

the calculations serve to confirm the validity of the empirical calibratiouns,

MSRE Main O0ff-Gas Line

 

Since the MSRE main off-gas line was only 1 in. in diameter we used
a line source as a model in our calculations. The physical configuration
of the detector-collimator-source is shown in Fig. D,l1. Distances of items

labeled in the figure are as follows:

Lco = geometrical collimator length

D = collimator diameter = 1/8 in.

o
S

12 in. = 30.48 cm,
0.317 cm,

fi

= distance from source to source face of collimator = 15 ft =
457 cm ,

L

length of source subtended by collimator.

Due to penetration of the edges of the collimator on both source and de-
tector ends, the length of the source subtended by the collimator is not
constant but depends on the gamma-ray energy. Mather,' in a study of col-
limators found that, to a first approximation, the effect of edge penetra-
tion is to reduce the length of the collimator by two mean free paths of

the gamma photons. The effective collimator length Lc thus can be computed

by

 

'R. L. Mather, J. Appl. Phys. 28, 1200 (1957),
195

L
co
‘(30.48 in, ™™
r/\N\ Dc = 1/8 in.
__SOURCE- COLLIMATOR

DETECTOR
— Y

2 457 .2 N e—

 

 

 

 

 

 

 

 

 

e

 

 

 

 

 

 

Fig. D.1l. Schematic of detector——collimator—line source arrangement.

LC = LCO—Z/Ut s (1)

where My is the linear attenuation coefficient for the collimator material
(lead).

The subtended source length can then be obtained from

2D

L
= ¢ £
L - Lc (H + 2) . (2)

Thus, in our experimental arrangement, the source length subtended varied

from 3.8 in. for 200-keV photons to 4.4 in. for 2000-keV photons.

Let S, = source strength per inch = 1 gamma in.~* sec~! and
Ac = geometrical area of collimator hole,
7 (0.317)% cn® = 0.07197 cn’.

Then if we consider a differential source length, dx, at point x on the
source, the photon intensity dG due to dx at the distance 7 + Lco (source-

to-detector distance) is given by
dq = —— (3)
Gn(H + L )?
co

If the coliimator were not between the source and the detector, the photon

current (due to dG) that would hit the detector over an area Ac would be

dIl = A4 _dG .
e

However, since the collimator is present, only those photons emitted from
the central portion of the source can go through any portion of the de-
tector end of the collimator hole. Photons emitted by parts of the source
that are off the source~collimator axis are partially shadowed from the
detector by the collimator and can only go through a fraction of the area
Ac' Thus Eq. (2) must be multiplied by the fraction of the area 4 that

is not shadowed by the collimator. Let this fraction be F(x), and we have

dT = 4 R@) d6 , (4)

where R(x) is a function of the distance x that the segment dx lies from
the collimator axis.

Now the quantity A(x) is a radial weighting function and corresponds
to the R(x) of Section 5.3. Although R(x) is a gaussian-type function, we
can approximate it by a linear function that has a value of unity at the
center of the source and a value of zero at a distance L/2 from the axis.
Within these boundary values, RF(x) is given by

B(x) = 1 — 2

f.’L’ . (5)

Thus by combining Egqs. (3), (4), and (5), we have an expression for the

photon current dI that leaves the segment dx and hits the detector, viz.,
197

2
S, A (1 —=ux) dx
a7 = e’ L : (6)
4r(H + L )*
co

 

We can now integrate this expression and obtain an equation for the total

photon current that hits the detector:

o 51 4, L/2 )
I = ————— 2/ (1 —-% ) doe (N
2 0 L
4n(H + L )
coO
S, A, L

2
8n(H + Lco)

Since the photon current through an area equal to Ac without the collimator

being present is

SZ Ac L

2
4 (H + LCO)

3

we see that collimator shadowing for a line source allows only one-half
of the photons directed toward the collimator exit hole (detector end of
collimator) to be transmitted through the collimator. The fraction of

photons P' that interact with the detector is given by

P' = 1 — exp (-uGe DGe Yy , (8)

where is mass attenuation coefficient of germanium (cm®/g), DGe is

Ge
density of germanium (5.33 g/cm®), and Y is diameter of the germanium de-

tector (2.7 cm). Thus the total count rate in the detector is given by

CR = IP' . (9)
198

To obtain the count rate in the photopeak, we must multiply by the fraction
of the counts that fall in the photopeak. Because these peak-to-total
ratios are difficult to calculate, we used values that were measured with
our collimated detector with the sources ®'Cr, *®7Cs, *?v, and 2®Al. These
sources each emit single gamma rays with energies of 320, 662, 1434 and
1778 keV respectively. The corresponding peak-to-total ratios found were
0,195, 0.106, 0.0551, and 0.046. When these peak-to-total values were
plotted versus energy on log-log coordinate paper, a straight line was ob-

tained that could be represented by

P = 26.828 0:822L (10)

where P is the peak-to-total ratio for photons with energy E.

Now, combining Egs. (2), (7), (9), and (10), we have for the count rate,

 

2D L r
q 21 & - ( -0.8521
5, 4, [ T 3 J [1 exp(-u,, Dy, ¥)| (26.82F )
CR = - .
8r(H + L )*?
co
(11)
Since the counting efficiency EF is simply
G
L
and SZ is assumed to be 1 gamma in.”* sec ', the right side of Eq. (11) is

also equal to the counting efficiency.

A computer program was written to evaluate Eq. (11) and obtain count-
ing efficiencies at several energies over a range from 100 to 2700 keV.
The program allowed for photon transmission through the edges of the col-
limator according to Egs. (1) and (2). Results of the calculated effici-

encies are shown in Table D.1l, along with experimental values. As can be
199

Table D.1. Calculated versus experimental counting efficiencies
for MSRE off-gas line

 

Efficiencies (x 10°)

 

 

Energy (keV) Measured Calculated Percent difference®
100 15.0 27.2 -81.3
200 12.1 13.7 -13.2
300 8.4 9.5 -13.1
400 6.2 6.2 0
500 5.1 5.1 0
600 4,2 4,1 2.4
700 3.6 3.4 5.5
800 3.2 3.0 6.2
900 2.9 2.6 10.3

1100 2.4 2,1 12.5
1300 2.1 1.8 14.3
1500 1.9 1.5 21.0
1700 1.7 1.3 23.5
2000 1.5 1.1 26.6
2700 1.04 0.79 24.0

 

aPercent diff. = 100(meas. — calc.)/meas.
200

seen from the table, agreement between calculated and measured efficiencies

is good from 200 to 1300 keV; below 200 keV the difference becomes very large.

MSRE Heat Exchanger

The MSRE heat exchanger® is composed of a 16 3/4=-in.-0D shell that
encloses a 14,5-in. diam tube bundle. The shell is made of Hastelloy N
and is 1/2 in. thick; the tube bundle, also made of Hastelloy N, consists
of 156 cooling U-tubes (312 tubes) plus twenty-three 1/2-in. dummy rods.

A typical cross section also contains eight 7/16-in, baffle-plate spacer
rods. Although these tubes and rods coated with fission products represent
a complex radioactive source, we have calculated counting efficiencies for
the heat exchanger using a model in which the tubes and their radioactive
coatings are assumed to be homogeneously dispersed throughout the tube
bundle.

The detector—collimatecr—heat exchanger arrangement is schematically
represented in Fig. D.2. The source region subtended in the heat exchanger
is a truncated cone with a front face diameter of 3.9 in. and a back face
diameter of 4.3 in, As in the case of the line source, because of trans-
mission of photons through the collimator edges around the collimator hole,
the diameter of the subtended area increases with increasing energy. This
effect is actually larger in the heat exchanger because of attenuation in
the heat exchanger tubes and shell. However, we will neglect the latter
effect and only acccunt for collimator edge transmission in the same manner
as before. Thus, the diameter of the subtended area will be determined
according to Eqs. (1) and (2). In addition, we will approximate the trun-
cated cone by a cylinder with a diameter equal to that of the front face
(detector end) of the cone. This approximation is reasonable since most

of the radiation measured comes from the front portions of the cone.

 

See R. C. Robertson, MSERE Design and Operations Report. Part I.
Description of Reactor Desigr, ORNL-IM~728, pp. 162~72, for a more com-
plete description of the heat exchanger,
201

 

 

 

 

 

 

 

 

  

i L" 1/2 in. i
I 12 inepee— 15 ¢ -
Pb —_—
Collimator
S S |

 

 

(Hole Diam = 1/8 in.)
Heat /

Detector Exchanger x—%J

i 14 ins

 

 

 

 

 

Fig. D.2, Detector-collimator assembly shown subtending
a cone~shaped source region in heat exchanger.

Because the collimator partially shadows all but the central portion
of the source from the detector, we must again obtain an average radial
weighting function for the radiation transmitted through the collimator.
Because this function is dependent only on the distance from the source-
collimator axis (and not on depth of the source), we can derive the func-
tion for a two-dimensional, circular area source. Consider the front face
of the cone with a differential area, o dp df, located a distance p from
the source-collimator axis. The photon current dI through the collimator

due to this area is given by

A48,
ar = —CS2 R o do d8 , (12)

2
4w (H + Lco)

where SA is the source strength per unit area and R(p) is the radial weight-
ing function. Other terms were previously defined.
We approximate F(p) by a linear function that is unity at the source

center and zero at the circumference. Thus
202

P
Rp) = 1—775 (13)

L/2 °
Substituting this result into Eq. (12) and integrating, we have for the
total photon current:
Ac S L/2 27 0

— —_—_—-.—-——.—.—.-.—-A - ———
ro- 2l [ =gz e do do (14)

0
4 (H + LCO)
4, 5y m(L/2)2
2
4r(H + Lco) 3

The photon current through an equivalent area Ac without the collimator

interposed is

Ac SA

—_— e a(L/2)?

4 (H + LCO)2
We see therefore that the collimator transmits only one-third of the pho-
tons directed toward its exit (detector side) hole.

We must now account for the fraction of those photons that are produced
in the source and directed toward the collimator that actually pass through
the source without being degraded. Consider the cross-sectional slice
(Fig. D.2) through the cone of thickness dx and at distance x from the

3 sec-l),

cone's front face. Denoting the source strength as SV (photons cm™
we obtain for the fraction dF of photons from the slice transmitted dis-
tance x

dF = 5, Ag eH T g ,

where AS is the area of the circular source surface facing the detector,
fl.is mass attenuation coefficient of Hastelloy N (taken as iron), and D is
average density of the MSRE tube bundle. Integrating over the diameter T

of the tube bundle we have
203

5,4, L —e™ D2
F o= . (15)

WD

We can now write an expression for the count rate in the photopeak
for those photons that are emitted by the subtended cone (cylinder in this

approximation) and transmitted by the collimator.

A e« S5 A _ =u DT
CR = ___Q___l/__fi__ e 1/3 - (_l_..:e?___l . []_ — eXP(—uGe DG Y] P! P,
4 (H + Lco)z u D ©

(16)

where P' is the fraction of photons transmitted through the heat exchanger
shell; other terms have previously been defined.

Once we have derived values for the volumetric source strength SV and
the average density of the MSRE tube bundle, we can use Eq. (16) to compute
the counting efficiency for the heat exchanger. Because our experimental
efficiencies were expressed as counts per photon per square centimeter of
tube, we assume that each square centimeter of tube in the tube bundle emits
1 photon cm™2 sec”!. From the number of tubes in the bundle and the bundle
diameter, we obtain a value of 1.09 cm? of tubing per cubic centimeter of
bundle. Thus SV = 1.09 photons cm~ 3 sec™!. Similarly we compute, for

the average density of metal in the bundle,-B = 0.976 g/cm3. The count-

ing efficiency is

where § = 1 photon cm™? sec”!,

A computer program was written to evaluate the counting efficiency at
several energies over the energy range 100 to 2700 keV and also to compute
the counting efficiency for an area source corresponding to the inner sur-
face of the heat exchanger shell. The results of our computations are sum-

marized in Table D.2, along with the experimentally measured efficiencies.
20U

As can be seen, the agreement between calculated and measured efficiency
values is excellent between 400 and 1500 keV. The contribution of the
heat exchanger shell varies from about 12% at 200 keV to about 4% at 2700
keV., The effect of the shell was not accounted for in our empirical cali-

bration but can now be seen to be a negligible factor.

Table D.2. Calculated and measured counting efficiencies
for the MSRE heat exchanger

 

 

 

Energy (keV) Measured Calculated tube bundle Calculated shell D1 D2
100 0.4 0.3 0.08 25.0 26.7
200 1.2 3.3 0.39 -175.0 11.8
300 2.7 3.8 0.36 -40.7 9.5
400 3.9 3.8 0.31 2.6 8.1
500 3.9 3.6 0.27 7.7 7.5
600 3.4 3.5 0.24 =29 6.8
700 3.1 3.4 0.22 -9.6 6.5
800 3.0 3.2 0.20 -6.6 6.2
900 3.0 3.1 0.18 -3.3 5.8

1100 3.0 2.9 0.16 3.3 5.5
1300 2.9 2.8 0.14 3.4 5.0
1500 2.8 2.6 0.13 7.1 5.0
2100 2.7 2.1 0.09 22,2 4.3
2700 2.5 1.6 0.07 36.0 4.3
9p1 = 100 (meas. - calc.)/meas.
bDZ = 100 calc. shell/calc. tube bundle.
ek
SO0~y P W N

NN HERFER R
NHEOW~OU ™ WN R

23,
24,
25-34,
35.
36.
37.
38.
39.
40.
41.
42,
43,
44,
45,
46-48.
49,
50.
51.
52,
53.
54.
55.
56.
57.

un
co

SO E m:zig nwrHON e

*

> GGy O;?:fifflgflfqrdpd?>2:C>>»h DU G Sy =S

205

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