ORNL-4865.txt

ORNL-4865

Fission Product Behavior
in the Molten Salt Reactor Experiment

E. L. Compere
S. S. Kirslis

E. G. Bohlmann
F. F.
W,

Blankenship
R. Grimes

0
{/

OAK RIDGE NATIONAL LABORATORY

OPERATED BY UNION CARBIDE CORPORATION = FOR THE U.S. ATOMIC ENERGY COMMISSION

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ORNL-4865%

UC-76 — Molten Salt Reactor Technology

Contract No, W-7405-eng-26

MOLTEN-SALT REACTOR PROGRAM

FISSION PRODUCT BEHAVIOR
IN THE MOLTEN SALT REACTOR EXPERIMENT

E. L. Compere E. G. Bohimann
S. S. Kirslis ~ F. F. Blankenship -
W. R. Grimes

OCTOBER 1975

OAK RIDGE NATIONAL LABORATORY
Oak Ridge, Tennessee 37830
: operated by
UNION CARBIDE CORPORATION
for the
ENERGY RESEARCH AND DEVELOPMENT ADMINISTRATION
ORNL-2474
ORNL-2626
ORNL-2684
ORNL-2723
ORNL-2799
ORNL-2890
ORNL-2973
ORNL-3014
ORNL-3122
ORNL-3215
ORNL-3282
ORNL-3369
ORNL-3419
ORNL-3529
ORNL-3626
ORNL-3708
ORNL-3812
ORNL-3872
ORNL-3936
ORNL-4037
ORNL-4119
ORNL-4191
ORNL-4254
ORNL-4344
ORNL-4396
ORNL-4449
ORNL-4548
ORNL-4622
ORNL-4676
ORNL-4728
ORNL-4782

This report is one of periodic reports in which we describe the progress of the program. Other reports
issued in this series are listed below.

Period Ending January 31, 1958
Period Ending October 31, 1958

Period Ending January 31, 1959 -

Period Ending April 30, 1959
Period Ending July 31, 1959
Period Ending October 31, 1959

Periods Ending January 31 and April 30, 1960

Period Ending July 31, 1960
Period Ending February 28, 1961
Period Ending August 31, 1961
Period Ending February 28, 1962
Period Ending August 31, 1962
Period Ending January 31, 1963
Period Ending July 31, 1963
Period Ending January 31, 1964
Period Ending July 31, 1964
Period Ending February 28, 1965
Period Ending August 31, 1965
Period Ending February 28, 1966
Period Ending August 31, 1966
Period Ending February 28, 1967
Period Ending August 31, 1967
Period Ending February 29, 1968
Period Ending August 31, 1968
Period Ending February 28, 1969
Period Ending August 31, 1969
Period Ending February 28, 1970
Period Ending August 31, 1970
Period Ending February 28, 1971
Period Ending August 31, 1971
Period Ending February 29, 1972
iii

< , CONTENTS
AB ST R AT . e e e e e 1
FOREW AR D . L e e 1
L. INTRODUCTION . . e et e e e e e et e e e e e e e 2
2. THE MOLTEN SALT REACTOR EXPERIMENT ...................... e e e e 3
3. MSRE CHRONOLOGY . ... et e e e e e e e e e e e e e et e i 10
3.1 Operation with 235U FUel ... ... .. .. . it e e e 10
3.2 Operation with 233U Fuel . ... .o . e e e e 12
4. SOME CHEMISTRY FUNDAMENTALS ... . e i 14
5. INVENTORY o e e e e 16
6. SALT SAMPLES . . e e e e 19
6.1 Ladle Samples . ... .. . e e 19
6.2 Freeze-Value Samples. .. ... e e e e e et e P 22
i 6.3 Double Wall Capsule . . ... .. e e e e 24
6.4 Fission Product Element Grouping . ................ ..o oo, e e e 27
.. 6.5 Noble-Metal Behavior . ... ... .. i e e e e 27
X
6.6 NIODIUI . . ot e e e e 28
6.7 lodine .. .............. A e 28
6.8 TellUmiUm . ..o e e e e e 29
7. SURFACE DEPOSITION OF FISSION PRODUCTS BY PUMP BOWL EXPOSURE . ................. 37
o Cable L e e e s 37
7.2 Capsule SUITaCes . . ... o e e e e 37
7.3 Exposure EXperiments .. ... .. ..ottt it i e ettt FETTI 37
' ] G 41
8. GAS SAMPLES .« e e 48
8.1 Freeze-Valve Capsule ................. e e e e e e e s 48
8.2 Validity of Gas Samples . ... .. . e e e 438
8.3 Double-Wall Freeze Valve Capsule ...... ... ... . i, . 49
8.4 EffectofMist ............ e e e e e e e e e e e e e e e e 50
9. SURVEILLANCE SPECIMENS . .. e e e e e e e e 60
- O.1 Assemblies I—4 ... e e e e 60
. 0. . Preface .. 60

9.1.2 Relative deposit intensity .. ... ...t i e 63
9.2

9.3

9.4

iv

Final Assembly . .. ... . . . e e e e e 70
0.2 DSIEN e e e e e e 70
9.2.2 Specimens and flow .. ... e e 70
0.2.3  Fission 1ecoil ... ... e e e e 73
9.2.4 Salt-seeking nuclides . . . ... . .. et 73
9.2.5 Nuclides with noble-gas precursors . .. .. ... .. .. ittt e e 73
9.2.6 Noble metals: niobium and molybdenum .. ...... ... ... ... ... ... ... . . ... 74
9.2 7 Ruthenium .. ... e e e 74

C9.2.8  Tellurium ..o e e e 74
9.2.9 dodine . . ... e e e e e 74
0.2.10 Sticking factOrs . .. .. .. e 74

Profile Data . ... . e e e e 75
9.3.1  Procedure . ... e e 75
0.3, RESUIS ..ot e e e e e 78
9.3.3 Diffusion mechanism relationships ... ....... ... ... .. .. . .. .. . 78
9.3.4 Conclusions from profiledata ....... ... ... ... . . . . . i e 84

Other Findings on Surveillance Specimens . ... ... ... .. i it e e 85

10. EXAMINATION OF OFF-GAS SYSTEM COMPONENTS OR SPECIMENS

11.

REMOVED PRIOR TO FINAL SHUTDOWN . . ... e e e 91
10.1 Examination of Particle Trap Removed after Run7 . ... ... ... ... .. ... . .. ... .. ... .. ....... 91
10.2 Examination of Off-Gas Jumper Line Removed after Run 14 .. ... ... ... ... .. ... ............. 92
10.2.1 Chemical analysis . .. ... ... e 93
10.2.2 Radiochemical analysis . ... ... ...ttt e e e e e 93
10.3 Examination of Material Recovered from Off-Gas Line afterRun16 .................... e 98
10.4 Off-Gas Line Examinations after Run 18 .. ... ... ... .. .. . . .. . . . . . . . 100
10.5 Examination of Valve Assembly from Line 523 after Run 18 .. .............. ... ... .. ....... 104
10.6 The Estimation of Flowing Aerosol Concentrations from Deposits on Conduit Wall .............. 106
10.6.1 Deposition by diffusion . ... ... .. 106
10.6.2 Deposition by thermophoresis ................ P JE 107
10.6.3 Relationship between observed deposition and reactor loss fractions . . . ................. 107
10.7 Discussion of Off-Gas Line Transport . . .. ...ttt e e e e 109
10.7.1 Salt constituents and salt-seeking nuclides ............ . ... . . . .. 109
10.7.2 Daughters of noble gases . ... .... ... ... e e 109
10.7.3 Noble metals . ... ... e e e e, 110
POST-OPERATION EXAMINATION OF MSRE COMPONENTS . ..... .. ... ... . i, 112
11.1 Examination of Deposits from the Mist Shield in the MSRE Fuel Pump Bowl . ............... ... 112
11.2 Examination of Moderator Graphite from MSRE . ... ... ... ... . .. . . 120
11.2.1 Results of visual examination . . ... ... .. ... . e e 120
11.2.2 Segmenting of graphite Stringer. . .. . ... .. .. e e 120
11.2.3 Examination of surface samples by x-ray diffraction ............... .. ... ... ....... 121
11.2.4 Milling of surface graphite samples .. ... ... ... . i e 121
11.2.5 Radiochemical and chemical analyses of MSRE graphite . ......................... ... 121
11.3 Examination of Heat Exchangers and Control Rod Thimble Surfaces ............ ... ... ... ... 125
11.4 Metal Transfer in MSRE Salt Circuits . ... ... ... i i e e e e 126
11.5 Cesium Isotope Migration in MSRE Graphite . . ... ... ... . . .. .. . i 127
12.

11.6 Noble-Metal Fission Transport Model . .. ... ... . . i i e e 128
11.6.1 laventory and model .. ... ... . . . e e e 128
11.6.2 Off-gasline deposits . .. ... ot i e et e et e et e e 130
11.6.3 Surveillance specimens . .. .. ... .. . e e 130
11.6.4 Pump bowlsamples . .......................... A 131

SUMMARY AND OVERVIEW . . . i e e e el e 135

12.1 Stable Salt-Soluble Fluorides . ... ... ... . . . e .. 135
12.1.1 Saltsamples . ... . e e e 135
12.1.2 Deposition . ... e e e 135
12.01.3 Gas SamIPIes .. .ot e e e e 136

12.2 Noble Metals .. ... e e e e 136
12,21 Salt-bOrme .« . e e e e e 136
12.2.2 NIODIUM .« .. e et et e e e e e e 136
12.2.3 Gas-DOINE .. ..ttt e e e e e e e e 136

12.3 Deposition in Graphite and Hastelloy N . .. .. ... . . o 137

12,4 T0odine . ... e e e e e e e e 138

125 Evaluation .. ... e e e e e e 138
Figure 2.1
Figure 2.2
Figure 2.3
Figure 2.4
Figure 3.1
Figure 6.1
Figure 6.3

Figure 6.4
Figure 6.5
Figure 7.1
Figure 7.2
Figure 7.3
Figure 8.1
Figure 8.2
Figure 9.1
Figure 9.2
Figure 9.3
Figure 9.4
Figure 9.5
~ Figure 9.6
Figure 9.7
Figure 9.8
Figure 9.9

Figure 9.10
Figure 9.11

Figure 9.12
Figure 9.13
Figure 9.14
Figure 9.15

Figure 9.16

vi

LIST OF FIGURES

Design flow sheet of the MSRE . .. .. ... . . 3
Pressures, volumes, and transit times in MSRE fuel circulatingloop .. ..................... 4
MSRE reactor vessel . .. .. .. e 5
Flow patternsin the MSRE fuel pump .. ... ... . . e e 6
Chronological outlines of MSRE operations . .. ....... ... ... . . . i 11
Sampler-enricher schematic ....... ... . . . . .. e 19
Apparatus for removing MSRE salt from pulverizer-mixer to

polyethylene sample bottle ... ... ... .. . e 20
Freeze valve capsuie ........................................................... 22
Noble-metal activities of salt samples .. .... ... ... e e 30
Specimen holder designed to prevent contamination by contact with transfer tube . ... ....... 39
Sample holder for short-term deposition test ... ........ ... ... ... i 41
Salt droplets on a metal strip exposed in MSRE pump bowl gas space for IOhr ............. 42
Freeze valve capsule ... ... ... . e e e 48
Double-wall sample capsule . ... .. ... .. e 50
Typical graphite shapes used in a stringer of surveillance specimens ...................... 60
Surveillance Specimen StriNEEr . ... ... . ittt i ettt et e et e e 61
Stringer cONtANMENt . . .. ... ... e e e 61
Stringer assembly .. ... .. e e 62
Neutron flux and temperature profiles for core surveillance assembly ..................... 63
Scheme for milling graphite samples .. . ... ... ... ... e e 64
Final surveillance specimen assembly . ... .. .. ... . . . . L e 70
Concentration profile for ' *7Cs in impregnated CGB graphite, sample P-92 ................ 75
Concentration profiles for #?Sr and '*°Ba in two samples of CGB, X-13 wide face,

and P- 58 . L e e 75
Concentration profiles for #®Srand '#°Ba in pyrolytic graphite . .................oo. ... 76
Concentration profiles for ®**Nb and ®5 Zr in CGB graphite, X-13, |

double exposure . .................... e e e e e e e e 76
Concentration profiles for ' 3 Ru on two faces of the X-13 graphite

specimen, CGB, double exposure .. ....... ... . i e 77
Concentration profiles for °¢Ru, !4 Ce, and ***Ce in CGB

graphite, sample Y-0 . .. .. e e 77
Concentration profiles for ' ! Ce and '**Ce in two samples of CGB graphite,

X-13,wide face, and P-38 .. ... ... .. e 78

Fission product distribution in unimpregnated CGB (P-55) graphite specimen irradiated
in MSRE cycle ending March 25, 1968 . .. .. .. ... .. . . e 79

Fission product distribution in impregnated CGB (V-28) graphite specimen irradiated
in MSRE cycle ending March 25, 1968
Figure 9.17

Figure 9.18
Figure 9.19
Figure 9.20
Figure 9.21

Figure 9.22
Figure 9.23
Figure 9.24
Figure 9.25

Figure 10.1
Figure 10.2
Figure 10.3
Figure 10.4
Figure 10.5
Figure 10.6
Figure 10.7
Figure 10.8

Figure 11.1
Figure 11.2
Figure 11.3
Figure 11.4
Figure 11.5

Figure 11.6

Figure 11.7
Figure 11.8

vii

Distribution in pyrolytic graphite specimen irradiated in MSRE for

cycle ending March 25, 1968 . . .. . . ... e e e 81
Uranium-235 concentration profiles in CGB and pyrolytic graphite . ............ e 84
Lithium concentration as a function of distance from the surface, specimen Y-7 .. ........... 87
Lithium concentration as a function of distance from the surface, specimen X-13 .. .......... 87
Fluorine concentrations in graphite sample Y-7, exposed to molten fuel .

salt in the MSRE fornine months .. ... ... .. ... .. . . . . . . 87
Fluorine concentration as a function of distance from the surface, specimen X-13 ........... 88
Comparison of lithium concentrations in samples Y-7and X-13 . ........ ... ... .......... 88
Mass concentration ratio, F/Li, vs depth, specimen X-13 ... ... ... ... ... ... . ... ... . . ..., 89
Comparison of fluorine concentrations in samples Y-7 and X-13, a smooth line having been

drawn through the datapoints ..... ... .. ... . . . . . 89
MSRE off-gas particle trap ...................... e e e e e e e e e 91
Deposits in particle trap Yorkmesh ......... ... .. .. o oL, e 92
Deposits on jumper line flanges after run 14 . .. .. ... .. . 94
Sections of off-gas jumper line flexible tubing and outlet tube afterrun 14 ................. 95
Deposit on flexible probe . .. ... . e Q8
Dust recovered from upstream end of jumper line after run 14 (16,000 X) . ................ 99
Off-gas line specimen holder as segmented after removal, followingrun 18 ................. 101
Section of off-gas line specimen holder showing flaked deposit,

removed after TUN 18 . .. ... . e 104
Mist shield containing sampler cage from MSRE pump bowl ...... .. ... ... ... ... ... ... 113
Interior of mist shield . . .. .. . .. e 114
Sample cage and mistshield . ............... ... ... ... ... ... I 115
Deposits On Sampler Cage . .. ..ot ittt e e e e 116
Concentration profiles from the fuel side of an MSRE heat exchanger tube measured

about 1.5 years after reactor shutdown ... .......... e e e 126
Concentration of cesium isotopes in MSRE core graphite at given distances

from fuel channel surface ... ... ... . .. .. . . . e 127
Compartment model for noble-metal fission transport in MSRE . ... ..................... 129
Ratio of ruthenium isotope activities for pump bowlsamples ......... .. ... ... ... ... 132
Table 2.1
Table 2.2
Table 3.1
Table 4.1
Table 5.1
Table 6.1

Table 6.2
Table 6.3

Table 6.4

Table 6.5
Table 6.6

Table 6.7
Table 6.8
Table 7.1
Table 7.2

Table 7.3
Table 7.4
Table 7.5

Table 7.6

Table 8.1
Table 8.2
Table 8.3
Table 9.1
Table 9.2
Table 9.3
Table 9.4
Table 9.5

Table 9.6

Table 9.7

viii

LIST OF TABLES

Physical properties of the MSRE fuelsalt ...... ... ... ... .. . . ... 8
Average composition of MSRE fuel salt . ... ... ... . ... . . 9
MSRE run periods and power accumulation .......... . ... . .. i 12
Free energy of formation at 650°C (AG g, keal) ...t 14
Fission product data for inventory calculations ... ... ... ... . i, 17
Noble-gas daughters and salt-seeking isotopes in salt samples from MSRE pump bowl

during uranium-235 Operations . . . . . ..ottt i e 21
Noble metals in salt samples from MSRE pump bowl during uranium-235 operation ......... 21
Data on fuel (including carrier) salt samples from MSRE pump bowl during

uranium-233 Operation . .. ... .. . e e e e 23
Noble-gas daughters and salt-seeking isotopes in salt samples from MSRE pump bowl

during uranium-233 Operation .. ..... ...ttt e e e i 24
Moble metals in salt samples from MSRE pump bowl during uranium-233 operation ......... 25
Operating conditions for salt samples taken from MSRE pump bowl during

uranium-233 Operation . . . .. .. . i e 26
Data for salt samples from pump bowl during uranium-235 operation .................... 31
Data for salt samples from MSRE pump bow! during uranium-233 operation ............ 32--36
Data for graphite and metal specimens immersed in pumpbowl ......................... 38
Deposition of fission products on graphite and metal specimens in float-window

capsule immersed for various periods in MSRE pump bowl liquid salt ................... 40
Data for wire coils and cables exposed in MSRE pumpbowl ........... ... ... ... ... .... 44
Data for miscellaneous capsules from MSRE pump bowl . .. ... ... ... ... . . . ... 45
Data for salt samples for double-walled capsules immersed in salt in the MSRE pump bowl

during uranium-233 Operation ... .. .. .. .. ... ..t et 46
Data for gas samples from double-walled capsules exposed to gas in the MSRE pump

bowl during uranium-233 operation .. ..... ... ... e 47
Gas-borne percentage of MSRE productionrate ........... ... ... .. . i 51
Gas samples 225U Operation ... .. ...ttt e e 53
Data for gas samples from MSRE pump bowl during uranium-233 operation ............ 54--59
Surveillance specimen data: first array removed afterrun7 ... ... ... ... ... .. .. 0., 66
Survey 2, removed after run 11, inserted afterrun 7 . . ... ... ... . 67
Third surveillance specimen survey, removed afterrun 14 ..., ... ... . ... ... .. ... . ....... 68
Fourth surveillance specimen survey, removed afterrun 18 . .. ... ... .. ... ............ 69
Relative desposition intensity of fission products on graphite surveillance specimens

from final core specimen array .. ... ... ... .. i e 71
Relative deposition intensity of fission products on Hastelloy N surveillance specimens

from final core specimen array .. ....... ... .ttt e 72
Calculated specimen activity parameters after run 14 based on diffusion

calculations and salt INVENLOTY ... ... . ittt e e e 84
. Table 9.8

Table 9.9
Table 9.10
Table 10.1
Table 10.2
Table 10.3
Table 10.4
Table 10.5
Table 10.6
Table 10.7
Table 10.8
Table 10.9
Table 10.10

Table 10.11
Table 11.1

Table 11.2

Table 11.3
Table 11.4
Table 11.5

Table 11.6

Table 11.7
Table 11.8
Table 12.1

Table 12.2
Table 12.3

ix

List of milled cuts from graphite for which the fission product content could be
approximately accounted for by the uranium present

...............................

.......................

Spectrographic analyses of graphite specimens after 32,000 MWhr

................

Percentage isotopic composition of molybdenum on surveillance specimens

Analysis of dust from MSRE off-gas jumper line

.....................................

Relative quantities of elements and isotopes found in off-gas jumper line

------------------

Material recovered from MSRE off-gas line afterrun 16, . ... ... ... ... . ... .. .uu ...,
Data on samples or segmenté from off-gas line specimen holder removed following run 18 .. ...

Specimens exposed in MSRE off-gas line, runs 15—18

.................................

Analysis of deposits from line 523 . ... . .. . e

Diffusion coefficient of particles in 5 psig of helium

..................................

Thermophoretic deposition parameters estimated for off-gas line

........................

Grams of salt estimated to enter off-gas system

......................................

Estimated percentage of noble-gas nuclides entering the off-gas based on deposited daughter

activity and ratio to theoretical value for full stripping . .. .......... P
Estimated fraction of noble-metal production entering off-gas system

....................

Chemical and spectrographic analysis of deposits from mist shield in
the MSRE pump bowl

Gamma spectrographic (Ge-diode) énalysis of deposits from mist shield in the
MSRE pump bowl

Chemical analyses of milled samples

.......................................................

...........................................................
..............................................

..................................

Radiochemical analyses of graphite stringer samples

Fission products in MSRE graphite core bar after removal in cumulative values
of ratio to inventory ;

---------------------------------------------------------

Fission products on surfaces of Hastelloy N after termination of operation expressed
as (observed dis min ™' ¢m™?)/(MSRE inventory/total MSRE surface area)

Ruthenium isotope activity ratios of off-gas line deposits

------------------------------

Ruthenium isotope activity ratios of surveillance specimens

----------------------------

Stable fluoride fission product activity as a fraction of calculated inventory
in salt samples from 233U operation

............................................

......................................

Relative deposition intensities for noble metals

Indicated distribution of fission products in molten-salt reactors
o
FISSION PRODUCT BEHAVIOR IN THE MOLTEN SALT REACTOR EXPERIMENT

E. L. Compere E.G. Bohlmann
S. S. Kirslis F. F. Blankenship
W. R. Grimes

ABSTRACT

Essentially all the fission product data for numerous and varied samples taken during operation of
the Molten Salt Reactor Experiment or as part of the examination of specimens removed after
particular phases of operation are reported, together with the appropriate inventory or other basis of
comparison, and relevant reactor parameters and conditions. Fission product behavior fell into distinct
chemical groups.

The noble-gas fission products Kr and Xe were indicated by the activity of their daughters to be
removed from the fuel salt by stripping to the off-gas during bypass flow through the pump bowl, and
by diffusion into moderator graphite, in reasonable accord with theory. Daughter products appeared
to be deposited promptly on nearby surfaces including salt. For the short-lived noble-gas nuclides,
most decay occurred in the fuel salt.

The fission product elements Rb, Cs, Sr, Ba, Y, Zr, and the lanthanides all form stable fluorides
which are soluble in fuel salt. These were not removed from the salt, and material balances were
reasonably good. An aerosol salt mist produced in the pump bowl permitted a very small amount to be
transported into the off-gas. '

lodine was indicated (with less certainty because of somewhat deficient material balance) also to
remain in the salt, with no evidence of volatilization or deposition on metal or graphite surfaces.

The elements Nb, Mo, Tc, Ru, Ag, Sb, and Te are not expected to form stable fluorides under the
redox conditions of reactor fuel salt. These so-called noble-metal elements tended to deposit
ubiquitously on system surfaces — metal, graphite, or the salt-gas interface — so that these regions
accumulated relatively high proportions while the salt proper was depleted.

Some holdup prior to final deposition was indicated at least for ruthenium and tellurium and
possibly all of this group of elements.

Evidence for fission product behavior during operation over a period of 26 months with 2350 fuel

"{more than 9000 effective full-power hours) was consistent with behavior during operation using 233y
fuel over a period of about 15 months {(more than 5100 effective full-power hours).

FOREWORD"

This report includes essentially all the fission product
data for samples taken during operation of the Molten
Salt Reactor Experiment or as part of the examination
of specimens removed after completion of particular
phases of operation, together with the appropriate
inventory or other basis of comparison appropriate to
each particular datum.

It is appropriate here to acknowledge the excellent
cooperation with the operating staff of the Molten Salt
Reactor Experiment, under P. N. Haubenreich. The
work is also necessarily based on innumerable highly
radioactive samples, and we are grateful for the con-
sistently reliable chemical and radiochemical analyses
performed by the Analytical Chemistry Division (J. C.
White, Director), with particular gratitude due C. E.

Lamb, U. Koskela, C. K. Talbot, E. I. Wyatt, J. H.
Moneyhun, R. R. Rickard, H. A. Parker, and H.
Wright,

The preparation of specimens in the hot cells was
conducted under the direction of E. M. King, A. A.
Walls, R. L. Lines, S. E. Dismuke, E. L. Long, D. R,
Cuneo, and their co-workers, and we express our
appreciation for their cooperation and innovative assist-
ance.

We are especially grateful to the Technical Publications
Department for very perceptive and thorough editorial
work.

We also wish to acknowledge the excellent assistance

received from our co-workers L. L. Fairchild, J. A.
Myers, and J. L. Rutherford.
1. INTRODUCTION

In molten-salt reactors (or any with circulating fuel),
fission occurs as the fluid fuel is passed through a core
region large enough to develop a critical mass. The
kinetic energy of the fission fragments is taken up by
the fluid, substantially as heat, with the fission frag-
ment atoms (except those in recoil range of surfaces)
remaining in the fluid, unless they subsequently are
subject to chemical or physical actions that transport
them from the fluid fuel. In any event, progression
down the radioactive decay sequence characteristic of
each fission chain ensues.

In molten-salt reactors, this process accumulates
many fission products in the salt until a steady state is
reached as a result of burnout, decay, or processing.
The first four periodic groups, including the rare earths,
fall in this category.

Krypton and xenon isotopes are slightly soluble gases
in the fluid fuel and may be readily stripped from the
fuel as such, though most of the rare gases undergo
decay to alkali element daughters while in the fuel and
remain there. i

A third category of elements, the so-called noble
metals (including Nb, Mo, Tc, Ru, Rh, Pd, Ag, Sb, and
Te) appear to be less stable in salt and can deposit out
on various surfaces.

There are a number of consequences of fission
product deposition. They provide fixed sources of
decay heat and radiation. The afterheat effect will
require careful consideration in design, and the associ-
ated radiation will make maintenance of related equip-
ment more hazardous or difficult. Localization (on
graphite) in the core could increase the neutron poison
effect. There are indications that some fission products

(e.g., tellurium) deposited on metals are associated with
deleterious grain-boundary effects,

Thus, an understanding of fission product behavior is
requisite for the development of molten-salt breeder
reactors, and the information obtainable from the
Molten Salt Reactor Experiment is a major source,

The Molten Salt Reactor Experiment in its operating
period of nearly four years provided essentially four
sources of data on fission products:

1. Samples — capsules of liquid or gas — taken from the
pump bowl periodically; also surfaces exposed there.

2. Surveillance specimens — assemblies of materials
exposed in the core. Five such assemblies were
removed after exposure to fuel fissioning over a
period of time.

3. Specimens of material recovered from various sys-
tem segments, particularly after the final shutdown.

4. Surveys of gamma radiation using remote collimated
instrumentation, during and after shutdown. As this
is the subject of a separate report, we will not deal
with this directly.

Because of the continuing generation by fission and
decay through time, the fission product population is
constantly changing. We will normally refer all measure-
ments back to the time at which the sample was
removed during fuel circulation. In the case of speci-
mens removed after the fuel was drained, the activities
will normally refer to the time of shutdown of the
reactor. Calculated inventories will refer in each case
also to the appropriate time.
2. THE MOLTEN SALT REACTOR EXPERIMENT

We will briefly describe here some of the character-
istics of the Molten Salt Reactor Experiment that might
be related to fission product behavior.

The fuel circuit of the MSRE'™ is indicated in Figs.
2.1 and 2.2. It consisted essentially of a reactor vessel, a
circulating pump, and the shell side of the primary heat
exchanger, connected by appropriate piping, all con-
structed of Hastelloy N.5-®Hastelloy N is a nickel-based
alloy containing about 17% molybdenum, 7% chrom-
ium, and 5% iron, developed for superior resistance to
corrosion by molten fluorides.

The main circulating “loop” (Fig. 2.2) contained
69.13 ft* of fuel, with approximately 2.9 ft* more in
the 4.8-ft*> pump bowl, which served as a surge volume.
The total fuel-salt charge to the; system amounted to
about 78.8 ft?; the extra volume, amounting to about
9% of the system total, was contained in the drain tanks
and mixed with the salt from the main loop each time
the fuel salt was drained from the core.

Of the salt in the main loop, about 23.52 ft* was in
fuel channels cut in vertical graphite bars which filled
the reactor vessel core, 33.65 ft> was in the reactor
vessel outer annulus and the upper and lower plenums,

6.12 ft*> was in the heat exchanger (shell side), and the
remaining 6.14 ft> was in the pump and piping.

About 5% (65 gpm) of the pump output was recir-
culated through the pump bowl. The remaining 1200
gpm (2.67 cfs) flowed through the shell side of the heat
exchanger and thence to the reactor vessel (Fig. 2.3).
The flow was distributed around the upper part of an
annulus separated from the core region by a metal wall
and flowed into a lower plenum, from which the entire
system could be drained. The lower plenum was
provided with flow vanes and the support structure for
a two-layer grid of 1-in. graphite bars spaced 1 in. apart,
covering the entire bottom cross section except for a
central (10 X 10 in.) area. One-inch cylindrical ends of
the two-inch-square graphite moderator bars extended
into alternate spacings of the grid. Above the grid the
core was entirely filled with vertical graphite moderator
bars, 64 in. tall, with matching round end half channels,
0.2 in.deep and 1.2 in. wide, cut into each face. There
were 1108 full channels, and partial channels equivalent
to 32 more. Four bar spaces at the comers of the
central bar were approximately circular, 2.6 in. in diam-
eter; three of these contained 2-in. control rod thim-

ORNL-DWG 5%-11410R

. COOLANT

PUMP LEGEND

——— FUEL SALT

ez COOLANT SALT

HELIUM COVER GAS
RADIDACTIVE OFF -GAS

".
OFF-GAS }
HCLCUP :
OVERFLOW TANK
1170 °F
ABSOLUTE = 0 A
FILTERS 1200 GPM. e e
aLDG ! REACTOR . Wigeser
T VENTILATION [ VESSEL POWER FREEZE FLANGE (TYP)
L ©; i ! B M
STACK  FAN ot n
| FR {
|~~~ COOLANT i ~FREEZE VALVE (TYR)
i é:’ SYSTEM $1
|
| I___i__._.l e
t i ‘ : ————— - e e i ————— e e e e et o I
] I o i b
| ? % ................................................ - LS SR SR 2 i o
\ LI ABSOLUTE
: M WATER  STEAM E:D__‘__%D fi_fi_?—o FILTERS
1 )
{ MAIN DR l ek
\ CHARCOAL | B
| €0 AUX. ]
| CHARCOAL 7 |
h BED —'l :
i - !
: FUEL |
] DRAIN FLUSH L 2_
! J TANK NO. 1 TANK Lo
} S
!
' |
! |
! )
s 4 SODIUM
FLUORIDE BED

Fig. 2.1. Design flow sheet of the MSRE.
ORNL-~DWG 70-5192

LOOP DATA AT 12C0°F, 5 psig, 1200 gpm
DIFF. TRANSIT
POSITION VOLUME TIME PRESSURE
(£13) {sec) {psia)
10 ' 20.4
1 11 0.41 733
> .76 0.28 69.7 )
3 6.42 2.29 44 4 (@/
a4 2.18 0.81 434
5 9.72 3.63 o
6 12.24 4.58 403
7 23.52 8.79 355
a i1.39 4.26 25.8
9 1.37 0.51 L
(0 0.73 0.27 0.4

Fig. 2.2. Pressures, volumes, and transit times in MSRE fuel circulating loop.

ble tubes, and the fourth contained a removable tubular

surveillance specimen array.

At reactor temperatures the expansion of the reactor
vessel enlarged the annulus between the core graphite

and the inner wall to about Y, in.

Model studies,”® indicated that although the Reyn-
olds number for flow in the noncentral graphite fuel
channels was 1000, the square-root dependence of flow
on salt head loss implied that turbulent entrance
conditions persisted well up into the channel.

Fuel salt leaving the core passed through the upper

plenum and the reactor outlet nozzle, to which the
reactor access port was attached. Surveillance speci-
mens, the postmortem segments of control rod thimble,

and a core graphite bar were withdrawn through the
access port.

The fuel outlet line extended from the reactor outlet
nozzle to the pump entry nozzle.

The centrifugal sump-type pump operated with a
vertical shaft and an overhung impeller normally at a
speed of 1160 rpm to deliver 1200 gpm to the discharge
line at a head of 49 ft, in addition to internal
circulation in the pump bowl, described below, amount-
ing to 65 gpm. Because many gas and liquid samples
were taken from the pump bowl, we will outline here
some of the relevant structures and flows. These have

been discussed in greater length by Engel, Haubenreich,
and Houtzeel
ORNL -LR-DWG ©1097RIA

FLEXIBLE CONDUIT TO
CONTROL ROD DRIVES

,

GRAPHITE SAMPLE ACCESS PORT \1

COOLING AIR LINES

ACCESS PORT COOLING JACKETS

FUEL QUTLET REACTOR 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 ~//

_) 8 —— CORE WALL COOLING ANNULUS
REACTOR CORE CAN

|
REACTOR VESSEL —

ANTI-SWIRL VANES
MODERATOR

VESSEL DRAIN LINE SUPPORT GRID

Fig. 2.3. MSRE reactor vessel.
Some of the major functions of the pump and pump
bowl were:

. fuel circulation pump,

. liquid expansion or surge tank,

. point for removal and return of system overflow,

. system pressurizer,

l

2

3

4

5. fission gas stripper,
6. gas addition point (helium, argon, oil vapor),
7. holdup and outlet for off-gas and purge gas,
8. fuel enricher and chemical addition point,

9. salt sample point,

10. gas sample point,

11. point for contacting specimen surfaces with liquid
or gas during operation,

12. point for postmortem excision of some system
surfaces.

The major flow patterns are shown in Fig. 2.4.

Usually the pump bowl, which had a fluid capacity of
4.8 {t*, was operated about 60% full. Although the
overflow pipe inlet was well above the liquid level and

OFFGAS
LINE

100 oy

o _

LEVEL
SAMPLE SCALE
CAPSULE o

(%)
CAGE
OVERFLOW
PIPE

was protected from spray, overflow rates of several
pounds per hour (0.1 to 10) resulted in the accumula-
tion, in a toroidal overflow tank below the pump, of
overflow salt, which was blown back to the pump bowl
at the necessary intervals (hours to weeks). The
overflow tank was connected to the main off-gas line,
but because the pump bowl overflow line extended to
the bottom of the overflow tank, little or no off-gas
took this path except when the normal off-gas exit
from the pump bowl had been appreciably restricted.

It was desirable® to remove as much of the xenon and
krypton fission gases as possible, particularly to miti-
gate the high neutron poison effect of '3°Xe.
Consequently, about 50 gpm of pump discharge liquid
was passed into a segmented toroidal spray ring near the
top of the pump bowl. Many Y% ¢ and %-in. perfora-
tions sent strong jets angled downward spurting into
the liquid a few inches away, releasing bubbles,
entraining much pump bowl gas — the larger bubbles
of which returned rapidly to the surface — and
vigorously mixing the adjacent pump bowl gas and
liquid. An additional “fountain” flow of about 15 gpm
came up between the volute casing seal and the impeller
shaft. Other minor leakages from the volute to the
pump bowl also existed. At a net flow of 65 gpm (8.7

ORNL—DWG 69~ 10172aA

\\N

/ REFERENCE
L LINE
/4] BuBBLER
; /’///
/ -~ | !
',/,_-: .__:_"%_.‘ :_’ _ &
X7 7
= = V)
- CR- ™ ___:___'." ._47'0 e .- ’J 7
SRR ; e — = S
_ NI P " DISCHARGE
- {_}— - .~ SALT

-4 <mmm RADIOACTIVE GAS

SUCTION == CLEAN GAS

Fig. 2.4. Flow patterns in the MSRE fuel pump.
cfm) into a pump bowl salt volume of 2.9 ft3, the
average residence time of salt in the pump bowl was
about 20 sec.

The pump bowl liquid flowed past skirts on the
volute at an average velocity of 0.11 fps, accelerating to
1.7 fps as the salt approached the openings to the pump
suction. Entrained bubble size can be judged by noting
that bubbles 0.04 c¢cm (0.016 in.) in diameter are
estimated by Stokes’s law to rise at a velocity of 0.1 fps
in pump bowl salt. '

The salt turbulence also provided an underflow entry
to the spiral metal baffle surrounding the sample
capsule cage. The baffle, curved into a spiral about 3 in.
in diameter, with about one-fourth of a turn overlap,
extended from the sample transfer tube at the top of
the pump bowl, downward through both gas and liquid
phases, to the sloping bottom of the pump bowl, with a
half-circle notch on the bottom near the pump bowl
wall to facilitate liquid entry. Subsequent upflow
permitted release of associated gas bubbles to the vapor
space, with liquid outflow through the/ ' -in. spiral gap.

Around the entry from the sample transfer tube at
the top of the pump bowl was a cage of five vertical
1/4-in. rods terminating in a ring near the bottom of the
pump bowl. Sample capsules, specimen exposure de-
vices, or capsules of materials to be dissolved in the fuel
were lowered by a steel cable into this cage for varying
periods of time and then withdrawn upward into the
sample transfer tube to be removed. Normally (when
not in use) a slight gas flow passed down the tube, due
to leakage of protective pressurization around closed
block valves (gas was also passed down the transfer tube
during exposure of many of the above-mentioned
items).

Gas could enter the sample baffle region from the
liquid and by diffusion via the spiral gap. The rate of
passage has not been determined, but some evidence
will be considered in connection with gas samples.

Purge gas, normally purified helium, entered the
pump bowl gas space through the annulus between the
rotating impeller shaft and the shield plug, normally at
a rate of 2.4 std liters/min. Some sealing oil vapor, of
the order of a few grams per day, is indicated to have
entered by this path. Two bubbler tubes (0.37 std
liter/min each) and a bubbler reference line (0.15 std
liter/min) also introduced gas into the pump bowl. With
an average pump bowl gas volume of 1.9 ft* at 5 psig
and 650°C, a flow of 3.3 std liters/min corresponds to a
gas holdup time of about 6.5 min.

In order to prevent spray from entering the overflow
line or the two %-in. off-gas exit lines in the top of the
pump bowl, a sheet metal skirt or roof extended across

the pump bowl gas space from the central shaft housing
to the top of the toroidal spray ring. That some aerosol
salt or organic mist still was borne out of the pump
bowl was indicated by the occasional plugging of the
off-gas line and by the examination of materials
recovered from this region, to be discussed in a
subsequent section,

The areas of the Hastelloy N surfaces exposed to
circulating salt in the MSRE fuel loop were given'? as
follows:

Pump 30 ft?
Piping 45 ft?
Heat exchanger 346 ft?
Reactor vessel 431 ft?

852 ft2 (0.7915 x 10% cm?)

The areas of the graphite surfaces in the core of the
MSRE are estimated from design data® to be:

Fuel channels 132.35 m?
Tops and bottoms 3.42 m?
Contact edges 80.25 m?
Support lattice bars 8.95 m?

224.97 m? (2.2497 x 10% ecm?)

Thus the total surface area of the MSRE fuel loop is
3.041 X 10° cm?.

Properties of the MSRE fuel salt have been given by
Cantor,’ ' Grimes,!? and Thoma.'® ‘Some of these are
given in Table 2.1. As given by Thoma,!? the average
composition of the fuel (as determined by chemical
analysis) was that shown in Table 2.2.

The power released in the reactor as a result of
nuclear fission was evaluated both from heat balance
data!4:15 and from changes in isotopic composition."

An originally assigned full power of 8 MW, corrected
for various small deviations in fluid properties and
instrument calibrations, gave a new heat balance full
power of 7.65, and the value based on isotopic changes
was 7.4 MW. Uncertainties in physical and nuclear
properties of the salt and in reactor instrument calibra-
tion are sufficient to account for the difference.

At a reactor power of 7.4 MW the mean power
density in the circulating fuel was 3.6 W per cubic
centimeter of salt. About 88% of the|fissions occurred
in the core fuel channels, about 6% in the upper head,
and 3% each in the lower head and in the outer
downflow annulus.!® The average thermal-neutron flux
in the circulationg fuel was about 2.9 X 10'? neutrons
cm™? sec”! for 22°U fuel at full power. The ther-
mal-neutron flux was about 3 X 10'3 neutrons cm ™

sec! near the center of the core and declined both
Table 2.1. Physical properties of the MSRE fuel salt

Property Value Estimated precision

Viscosity n(centipoises) = 0.116 exp [3755/T(°K}] 17
Thermal conductivity 0.010 watt cm ™! °C~} *10%
Electrical conductivity K=-2.22+681x 10731(°C)a +10%
Liquidus temperature 434°C 13°C
Heat capacity

Liquid Cp=057calg™ °C™! 3%

Solid Cp=0.31+3.61X107*T(°C) calg™* °C™" *3%
Density

Liquid p=2575-5.13 X 107*T(°C) *1%

= 139.9 1b/ft3 at 650°C

Expansivity 2.154 X 107%/°C at 600°C *10%
Compressibility Bpr(°K) =2.3X 10712 exp [1.0 X 10737TC°K)] cm?/dyne Factor 3
Vapor pressure log P(torrs) = 8.0 — 10,000/T(°K) Factor 50 from 500 to 700°C
Surface tension ¥ =260 ~ 0.127(°C) dynes/cm +30, -10%
Solubility of He, Kr, Xe TCC) He Kr Xe > Factor 10

Isochoric heat capacity, C,,

Sonic velocity

500 6.6 0.13 0.03
600 106 0.55 0.17
700 151 1.7  0.67
800 201 44 20

X 1078 moles cm™ melt atm ™!

Cy
o Cp
TCO calg™ calgmole™ calgatom™ —
°K—l OK"I OK"I CV
500 0.48, 16. 6.9, 1.1,
600  0.48, 15.5 6.8, 1.1

700 0.475 15.4 6.7, 1.20

500°C: u = 3420 m/fsec
600°C: 1= 3310 m/sec
700°C: i= 3200 mfsec

Thermal diffusivity

500°C:D =209 X 1073 cm?/sec
600°C: D =2.14 X 1073 cm?/sec
700°C: D =2.15 X 1073 cm?/sec

Kinematic viscosity

500°C: ¥=7.44 X 1072 cm?/sec
600°C: v=4.34 X 1072 cm?/sec
700°C: v=2.84 X 1072 cm?/sec

Prandt! number

500°C: Pr=35.¢
600°C: Pr=20.4
700°C: Pr=13.,

A pplicable over the temperature range 530 to 650°C. The value of electrical conductivity given here was estimated by
G. D. Robbins and is based on the assumption that ZrF4 and UF4 behave identically with ThF4; see G. D. Robbins and
A. S. Gallanter, MSR Program Semiannu. Frogr. Rep. Aug. 31, 1970, ORNL4548, p. 159;ibid., ORNL4622, p. 101,
Table 2.2. Average composition of MSRE fuel salt

Runs 4-14° Runs 16-20°
LiF, mole % 64.1 + 1.1 64.5+1.5
BeF,, mole % 30,L0+1.0 304+1.5
Z1F g4, mole % 5.0+0.19 4.90 £ 0.16
UF,4 , mole % 0.809 £ 0.024 0.137  0.004
Cr, ppm 64 + 13 (range 35-80) 80 14 (range 35-100)
Fe, ppm 130 £ 45 157 £ 43
Ni, ppm 67 + 67 46 = 14

20Operation with 235y fuel.
bOperation with 233U fuel.

radially and axially to values about 10% of this near the
graphite periphery.The fast flux was about three times
the thermal flux in most core regions.

B. E. Prince'” computed the central core flux for
2330 to be about 0.8 X 10'? neutrons cm ™2 sec ™! per
megawatt of reactor power, or about 6 X 103 at full
power. The relatively higher flux for the 233U fuel
results from the absence of 238U as well as the greater
neutron productivity of the 233U.

Across the period of operation with 235U fuel, 23°Pu
was formed more rapidly than it was burned, and the
concentration rose until about 5% of the fissions were
contributed by this nuclide. During the 233U opera-
tions, the plutonium concentration fell moderately but
was replenished by fuel addition. The resultant effects
on fission yields will be discussed later.

References

1. P. N. Haubenreich and J. R. Engel, “Experience
with the Molten Salt Reactor Experiment,” Nucl. Appl.
Technol. 8, 118—37 (February 1970).

2. R. C. Robertson, MSRE Design’ and Operations
Report, Part 1. Description of Reactor Design, ORNL-
TM-728 (January 1965).

3. J. R. Engel, P. N Haubenreich, and A. Houtzeel,
Spray, Mist, Bubbles, and Foam in the Molten Salt
Reactor Experiment, ORNL-TM-3027 (June 1970).

4. W. B. McDonald, “MSRE Design and Con-
struction,” MSR Program Semiannu. Prog. Rep. July
31, 1964, ORNL-3708, pp. 22-83.

5. H. E. McCoy et al., “New Developments in
Materials for Molten-Salt Reactors,” Nucl. Appl. Tech-
nol. 8, 156—69 (February 1970).

6. A. Taboada, “Metallurgical Developments,” MSR
Program Semiannu. Progr. Rep. July 31, 1964, ORNL-
3708, pp. 330--72.

7. D. Scott, Jr., “Component Development in Sup-
port of the MSRE,” MSR Program Semiannu. Progr.
Rep. July 31, 1964, ORNL-3708, pp. 167-90.

8. R. J. Kedl, Fluid Dynamic Studies of the Molten
Salt Reactor Experiment (MSRE) Core, ORNL-
TM-3229 (Nov. 19, 1970).

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

10. J. A. Watts and J. R. Engel, “Hastelloy N Surface
Areas in MSRE,” internal memorandum MSR-69-32 to
R. E. Thoma, Apr. 16, 1969.(Internal document — no
further dissimination authorized.)

11. S. Cantor, Physical Properties of Molten-Salt
Reactor Fuel, Coolant and Flush Salts, ORNL-TM-2316
(August 1968).

12. W. R. Grimes, “Molten Salt Reactor Chemistry,”
Nucl. Appl. Technol. 8(2), 137—55 (February 1970).

13. R. E. Thoma, Chemical Aspects of MSRE Opera-
tion, ORNL-4658 (December 1971).

14. C. H. Gabbard, Reactor Power Measurement and
Heat Transfer Perfornmnce in the MSRE, ORNL-TM-
3002 (May 1971).

15. C. H. Gabbard and P. N. Haubenreich, “Test of
Coolant Salt Flowmeter and Conclusions,” MSR Pro-
gram Semiannu. Progr. Rep. Feb. 28, 1971, ORNL-
4676, pp. 17-18.

16. J. R. Engel, “Nuclear Characteristics of the
MSRE,” MSR Program Semiannu. Progr. Rep. July 31,
1965, ORNL-3708, pp. 83—-114.

17. B. E. Prince, “Other Neutronic Characteristics of
MSRE with 233U Fuel,” MSR Program Semiannu.
Progr. Rep. Aug. 31, 1967, ORNL-4191, pp. 54-61.
10

3. MSRE CHRONOLOGY

A sketchy chronology of the MSRE, with an eye
toward factors affecting fission product measurements,
will be given below. More complete details are avail-
able.!3

3.1 Operation with ?>>U Fuel

The MSRE was first loaded with flush salt on
November 28, 1964; after draining the flush salt, 452
kg of carrier salt ("LiF-BeF,-ZrF,, 62.4-32.3-5.3 mole
%; mol. wt 40.2) was added to a drain tank followed by
235 kg of "LiF-?38UF, eutectic salt (72.3-27.7 mole
%, mol. wt 105.7) in late April. Circulation of this salt
was followed by addition of "LiF-233UF, (93% en-
riched) eutectic salt beginning on May 24, 1965.
Criticality was achieved on June 1, 1965. Addition of
enriched capsules of "LiF-223UF, eutectic salt con-
tinued_throughout zero-power experiments, which in-
cluded controlled calibration. The loop charge at the
beginning of power operation consisted of a total of
4498 kg of salt (nominal composition by weight, ”Li,
11.08%; Be, 0.35%; Zr, 11.04%; and U, 4.628%), with
390 kg in the drain tank (ref, 2, Table 2.15). -

Operation of the MSRE was commonly divided into
runs, during which salt was circulating in the fuel loop;
between runs the salt was returned to the drain tanks,
mixing with the residual salt there,

Run 4, in which significant power wa$ first achieved,
began circulation in late December 1965; the approach
to power has been taken arbitrarily as beginning at
noon January 23, 1966, for purposes of accounting for
fission product production and decay.

Significant events during the subsequent operation of
the MSRE until the termination of operation on
December 12, 1969, are shown in Fig. 3.1. The time
period and accumulated power for the various runs are
shown in Table 3.1.

Soon after significant power levels were reached,
difficulty in maintaining off-gas flow developed. De-

-posits of varnish-like material had plugged small pas-
sages and a small filter in the off-gas system. A small
amount of oil in the off-gas holdup pipe and from the
pump had evidently been vaporized and polymerized by
the heat and radiation from gas-borne fission products.
The problem was relieved by installation of a larger and
more efficient filter downstream from the holdup pipe.
On resumption of operation in April 1966, full power
was reached in run 6 after a brief shutdown to repair an
- electrical short in the fuel sampler-enricher drive. The
first radiochemical analyses of salt samples were re-
-ported for this run. Run 7, which was substantially at
full power, was terminated in late July by failure of the

blades and hub of the main blower in the heat removal
system. While a replacement was redesigned, procured,
and installed, the array of surveillance specimens was
removed, and examinations (reported later) were made.
Some buckling and cracking of the assembly had
occurred* because movement resulting from differential
expansion had been inhibited by entrapment and
freezing of salt within tongue-and-groove joints. Modi-
fications in the new assembly permitted its continued
use, with removals after runs 11, 14, and 18, when it
was replaced by an assembly of another design.

Run 8 was halted to permit installation of a blower;
run 9, to remove from the off-gas jumper flange above
the pump bowl some flush salt deposited by an overfill.

During run 9 an analysis for the oxidation state of the
fuel resulted in a U**/U*" value of 0.1%. Because values
nearer 1% were desired, additions of metallic beryllium
as rod (or powder) were made’ using the sampler-
enricher, interspersed with some samples from time to
time to determine U¥*/U*%".

During run 10 the first “freeze-valve” gas sample was
taken from the pump bowl. The series of samples begun
at this time will be discussed in a later section. Run 10
operated at full power for a month, with a scheduled
termination to permit inspection of the new blower. .

Run 11 lasted for 102 days, essentially at full power,
and was terminated on ‘schedule to permit routine
examinations and return of the core surveillance speci-
men assembly. During this run a total of 761 g of 235U
was added (as LiF-UF, eutectic salt) without difficulty
[through the sampler-enricher, while the reactor was
in operation at full power. After completion of main-
tenance the reactor was operated at full power during
run 12 for 42 days. During this period, 1527 g of
235U was added using the sampler-enricher. Beryl-
lium additions were followed by samples showing
U3 /U* of 1.3 and 1.0%. Attempts to untangle the
sampler drive cable severed it, dropping the sample
capsule attached to it, thus terminating run 12. The
cable latch was soon recovered; the capsule was sub-
sequently found in the pump bowl during the final
postmortem examination. Run 14 commenced on
September 20, after some coolant pump repairs, and
continued without fuel drain for 188 days; the reactor
was operated subcritical for several days in November
to permit electrical repairs to the sampler-enricher.
Reactor power and temperature were varied to deter-
mine the effect of operating conditions on '35Xe
stripping.® During run 14 the first subsurface salt
samples were taken using a freeze-valve capsule.
SALT IN
FUEL LOCP

FUEL S
FLUSH

FOWER

0 2 4 6 8 10
POWER (Mw)

]

DYNAMICS TESTS

INVESTIGATE
OFFGAS PLUGGING

REPLACE VALVES
AND FILTERS

RAISE POWER
REPAIR SAMPLER
ATTAIN FULL POWER

CHECK CONTAINMENT

FULL - POWER RUN

MAIN BLOWER FAILURE

REPLACE MAIN BLOWER
MELT SALT FROM GAS LINES
REPLACE CORE SAMPLES
TEST CONTAINMENT

RUN WITH ONE BLOWER
INSTALL SECOND BLOWER

ROD OUT OFFGAS LIiNE
CHECK CONTAINMENT

30-day RUN
AT FULL POWER

REPLACE AIR LINE
DISCONNECTS

SUSTAINED OPERATION
AT HIGH POWER

REPLACE CORE SAMPLES
TEST CONTAINMENT

} REPAIR SAMPLER

11

SALT iN
FUEL LOOP

N

1968 —

FUEL RSN
FLusH [ ]

POWER

N

Fig. 3.1. Chronological outline of MSRE operations.

——

S e . ot gt

ORNL-DWG 69— 7293R2

XENON  STRIPPING
EXPERIMENTS

MAINTENANCE
REPLACE CORE SAMPLES

1 } INSPECTION AND

TEST AND MODIFY
FLUCRINE DiSPOSAL -
SYSTEM

PROCESS FLUSH SALT
PROCESS FUEL SALT

LOAD URANIUM-233
REMOVE LOADING DEVICE

233, 2ERO- POWER
PHYSICS EXPERIMENTS

INVESTIGATE FUEL
SALT BEHAVIOR

CLEAR CFFGAS LINES
REPAIR SAMPLER AND
CONTROL ROD DRIVE
233 DYNAMICS TESTS

INVESTIGATE GAS
IN FUEL LOOP

HIGH-POWER OPERATION
0 MEASURE 233U o /o

REPLACE CORE SAMPLES
REPAIR ROD DRIVES
CLEAR QFFGAS LINES

INVESTIGATE COVER GAS,

L XENON, AND FISSION

PRODUCT BEHAVIOR
ADD PLUTONIUM

IRRADIATE ENCAPSULATED U

MAP F.P. DEPOSITION WITH
GAMMA SPECTROMETER
MEASURE TRITIUM,
SAMPI.E FUEL

REMOVE CORE ARRAY
PUT REACTOR [N STANDBY
12

Table 3.1. MSRE run periods and power accumulation

Cumulative total

Cumulative effective full-

Run Date started Date drained Run hours h a
ours power hours
4 1-23-664 1-26-66 80.8° 80.8 4
5 - 2-13-66 2-16-66 55.0 567.5 5
6A 4-8-66 4-22-66 342.1 2,146.3 54
6B 4-25-66 4-29-66 107.5 2,312.7 115
6C 5-8-66 5-28-66 475.0 3,005.0 377
7A 6-12-66 6-28-66 348.1 37114 684
7B 6-30-66 7-23-66 553.0 4.355.7 1,055
Surveillance specimen assembly removed. New assembly installed.
8 10-8-66 10-31-66 546.1 6,747.6 1,386
9 11-766 11-20-66 301.2 7,213.0 1,545
10 12-14-66 1-18-67 827.2 86285 2,262
11 1-28-67 5-1167 2461.4 11,340.6 4,510
Surveillance specimen assembly removed. Reinstalled.
12 6-18-67 8-11-67 1277.8 13,548.3 5,566
13 9-15-67 9-18-67 77.8 14,471.7 5,626
14 9-20-67 3-25-68 4468.2 18,997.0 9,005
Surveiliance specimen assembly removed, reinstalled. Off-gas specimen installed.
235U removed from carrier salt by fluorination. 2339 fuel added.
15 10-2-68 11-28-68 1372.1 24 .956.1 9,006.5
16 12-12-68 12-17-68 111.0 25,404.0 9,006.5
17 1-13-69 4-10-69 2085.8 28,146.1 10,487
18A 4-12-69 4-15-69 74.7 28.269.3 10,553
18B 4-16-69 6-1-69 1104 .4 29.402.6 11,547
Surveillance specimen assembly removed. New assembly installed.
19 8-17-69 11-2-69 1856.7 33,098.7 12,790
20 11-25-69 12-12-69 396.7 34,0553 13,172

Final drain. Surveillance specimen assembly removed. System to standby.

Postmortem, January 1971. Segments from core graphite, rod thimble, heat exchanger, pump bowl, freeze valve. System to

standby.

9From beginning of approach to power, taken as noon, Jan. 23, 1966. Prior circulation in run 4 not included.

After the scheduled termination of run 14, the core
surveillance specimen assembly was removed for exami-
nation and returned. The off-gas jumper line was
replaced; the examination of the removed line is
reported below. A specimen assembly was inserted in
the off-gas line.

All major objectives of the 235U operation had been
achieved, culminated by the sustained final run of over
six months at full power, with no indications of any
operating instability, fuel instability, significant corro-
sion, or other evident threats to the stability or ability
to sustain operation indefinitely.

3.2 Operation with 23U Fuel

It remained to change the fuel and to operate with
233U fuel, which will be the normal fuel for a

molten-salt breeder reactor. This was accomplished
across the summer of 1968. The fuel, in the drain tanks,
was treated with fluorine gas, and the volatilized UF,
was caught in traps of granular NaF. Essentially all 218
kg of uranium was recovered,? and no fission products
(except ®SNb), inbred plutonium, or other substances
were removed in this way. The carrier salt was then
reduced by hydrogen sparging and metallic zirconium
treatment, filtered to remove reduced corrosion prod-
ucts, and returned to the reactor. A mixture of 233 UF,
and "LiF was added to the drain tanks, and some
238UF, was included to facilitate desired isotope ratio
determinations. _
Addition of capsules using the sampler-enricher per-
mitted criticality to be achieved, and on October 8,
1968, the U.S. Atomic Energy Commission Chairman,
Glenn Seaborg, a discoverer of 233U, first took the
reactor to significant power using 233U fuel.

The uranium concentration with 233U fuel (83%
enriched) was about 0.3 mole %. The fuel also
contained about 540 g of ?3°Pu, which had been
formed during the 233U operation when the fuel
contained appreciable 238 U.

During the final months of 1968, zero-power physics
experiments were accompanied by an increase in the
entrained gas in the fuel. Beryllium was added to halt a
rise in the chromium content of the fuel. Some finely
divided iron was recovered from the pump bowl using
sample capsules containing magnets. During a subse-
quent shutdown to combine all fuel-containing salt in
the drain tank for base-line isotopic analysis, a stricture
in the off-gas line was removed, with some of the
material involved being recovered on a filter.

At the beginning of run 17 in January 1969, the
power level was regularly increased, with good nuclear
stability being attained at full power. Transients attrib-
uted to behavior of entrained gas were studied by
varying pump speed and other variables; argon was used
as cover gas for a time. Freeze-valve gas samples and salt
samples were taken, and a new double-wall-type sample
capsule was employed. Further samples were taken for
isotopic analysis. The lower concentrations of uranium
in the fuel led to unsuccessful efforts to determine the
U3*/U* ratio. However, beryllium additions were con-
tinued as Cr** concentration increases indicated.

In May 1969, restrictions in the off-gas lines appeared
and subsequently also in the off-gas line from the
overflow tank. Operation continued, and run 18 was
terminated as scheduled on June 1.

Surveillance specimens were removed, and an assem-
bly of different design was installed. This assembly
contained specially encapsulated uranium, as well as
material specimens. A preliminary survey of the distri-
bution of fission products was conducted, using a
collimated Ge(Li) diode gamma spectrometer.® This
was repeated more extensively after run 19,

After completing scheduled routine maintenance, the
reactor was returned to power in August 1969 for run
19. Plutonium fluoride was added, using the sampler-
enricher, as a first step in evaluating the possibility of
using this material as a significant component of
molten-salt reactor fuel.

At the end of run 19, the reactor was drained without
flushing to facilitate an extensive gamma spectrometer
survey of the location of fission products.

The fate of tritium in the system was of considerable
interest, and a variety of experiments were conducted
and samples taken to account for the behavior of this
product of reactor operation.” 8

13

Because salt aerosol appeared to accompany the gas
taken into gas sample capsules, a few double-walled
sample capsules equipped with sintered metal filters
over the entrance nozzles were used in run 20.

After final draining of the reactor on December 12,
1969, the surveillance specimen assembly was removed
for examination, and the reactor was put in standby.

In January 1971 the reactor cell was opened, and
several segments of reactor components were excised
for examination. These included the sampler-enricher
from the pump bowl, segments of a control rod thimble
and a central graphite bar from the core, segments of
heat exchanger tubes and shell, and a drain line freeze
valve in which a small stress crack appeared during final
drain operations. The openings in the reactor were
sealed, and the reactor crypt was closed.

References

1. P. N. Haubenreich and J. R. Engel, “Experience
with the Molten Salt Reactor Experiment,” Nucl. Appl.
Technol 8(2), 18—36 (1969).

2. R. E. Thoma, Chemical Aspects of MSRE Opera-
tion, ORNL-4658 (December 1971).

3. MSR Program Semiannu. Progr. Rep. (a) July 31,
1964, ORNL-3708; (b) Feb. 28, 1965, ORNL-3812;(c)
Aug. 31, 1965, ORNL-3872; (d) Feb. 28, 1966,
ORNL-3936; (¢) Aug. 31, 1966, ORNL-4037; (f) Feb.
28, 1967, ORNL4119; (g) Aug. 31, 1967, ORNL4191;
(h) Feb. 29, 1968, ORNL4254; (i) Aug. 31, 1968,
ORNL-4344; (j) Feb. 28, 1969, ORNL-4396; (k) Aug.
31, 1969, ORNL-4449; (I) Feb. 28, 1970, ORNL-4548;
(m) Aug. 31, 1970, ORNL-4622; (n) Feb. 28, 1971,
ORNL-4676;(0) Aug. 31, 1971, ORNL-4728.

4. W. H. Cook, “MSRE Materials Surveillance Test-
ing,” MSR Program Semiannu. Progr. Rep. Aug. 31,
1966, ORNL-4037, pp. 97—103.

5. J. R. Engel and R. C. Stefty, Xenon Behavior in
the Molten Salt Reactor Experiment, ORNL-TM-3464
{October 1971). :

6. A. Houtzeel and F. F. Dyer, A Study of Fission
Products in the Molten Salt Reactor Experiment by
Gamma Spectrometry, ORNL-TM-3151 (August 1972).

7. P. N. Haubenreich, (g) “Tritium in the MSRE:
Calculated Production Rates and Observed Amounts,”
ORNL-CF-70-2-7 (Feb. 4, 1970); (b) “A Review of
Production and Observed Distributions of Tritium in
MSRE in the Light of Recent Findings,” ORNL-
CF-71-8-34 (Aug. 23, 1971). (Internal documents — no
further dissemination authorized).

8. R. B. Briggs, “Tritium in Molten Salt Reactors,”
Reactor Technol 14(4), 335—42 (Winter 1971-72).
14

4. SOME CHEMISTRY FUNDAMENTALS

Discussions of the chemistry of the elements of major
significance in molten-salt reactor fuels have been made
by Grimes,! Thoma,? and Baes.®> Some relevant high-
lights will be summarized here.

The original fuel of the Molten Salt Reactor Experi-
ment consisted essentially of a mixture of 7 LiF-BeF,-
ZrF,-UF, (65-29-5-0.9 mole %). The fuel was circu-
lated at about 650°C, contacting graphite bars in-the

reactor vessel and passing then through a pump and

heat exchanger. The equipment was constructed of
Hastelloy N, a Ni-Mo-Cr-Fe alloy (71-17-7-5 wt %).
Small amounts of structural elements, particularly
chromium, iron, and nickel, were found in the salt.

The concentration of fission product elements in the
molten salt fuel is lower than that of constituent or .

structural elements. The following estimate will indicate

the limits on the concentration of fission products

evenly distributed in the fuel.
For a single nuclide of fission yield y, at a given

power P, the number of existing nuclide atoms in the -

7 “salt is
A=FXPXyXrTt,

where F is the system fission rate at unit power and  is
the effective time of operation. This is the actual time

of operation for a stable nuclide and equals 1/X at

steady state for a radioactive nuclide. The contribution
to the mole fraction of the fission product nuclide in
4.5 X 10% g of salt of molecular weight 40 is then

___ A
6 X 1023

4.5 X 108

X 40

As an example, for a single nuclide of 1% yield and
30-day halfife at 8 MW, 4 = 9.4 X 102! atoms and
X = 1.3 X 1077. Because the inventory of a fission
product element involves only a few nuclides, many
radioactive, the mole fractions are typically of the order
of 1 X 107¢ or less.

Traces of other substances may have entered the salt
in the pump bowl, where salt was brought into vigorous
contact with the purified helium cover gas. Flow of this
gas to the off-gas system served to remove xenon and
krypton fission gases from the system. A slight leakage
or vaporization of oil into the pump bowl used as a
lubricant and seal for the circulating pump introduced
hydrocarbons and, by decomposition, carbon and hy-
drogen into the system. For the several times the
reactor vessel was opened for retrieval of surveillance

assemblies and for maintenance, the possible ingress of
cell air should be taken into account.

The binary molten fluoride system LiF-BeF, (66-34
mole %) melts* at about 459°C. The solution chemistry
of many substances in this solvent has been discussed
by Baes.’> Much of the rédox-and oxide precipitation
chemistry can be summarized in. terms of the free -
energy. of formation of undissolved species. . -

- Free energies of formation of various species calcu--
lated at 650°C largely from" Baes’s data are shown in

Table 4.1. The elements of the table are listed in terrs -
of the telative redox stability of the dissolved fluorides. -

Table 4.1. 'F_f_ee energy of formation at 650°C (‘AGof, kcal)-

Lli_+,_Bef, and F~ are at unit activity; all others, -
. activities in mole fraction units

" Dissolved in- .

Solid = LiFBeF, . Gas

LiF SR -126.49

LaF3  -363.36. ~354.49
_ CeF5 —364.67 ~356.19

NdF3 - —341.80 —332.14 .

BeF, © —216.16

BeO ~123.00 -109.37

Bel, . ' ~74.48

UF, -310.92. -300.88

UF, —-389.79. -392.52

uo, '~221.08 S
UFg , —449.89
PuF;5 ~31693 —308.10 o
/,Puy 03 —-185.39 '

Z1F, ~392.92

Z10, -21942

NbF,4 (—296.35)

NbF —366.49
1/,Nb, 05 -179.14

NbC -32.4

CrF, (-150.7) ~152.06

/5Cr3C, ~7.5to -8.5

FeF, ~138.18 ~134.59

NiF, —~121.58 ~113.40

MoF, (—186.3)

MOOQ -99 .81

MoFg -306.65"
TeFy ~259.13
TeF5 “'23226
TeF, ~200.59
TeF, ~98.36
TeF —42.15
T62F10 —446.11
CF4 -189.57
HF ~50.29 -66.12
H,0 -47.04
RuFs ~173.72

As one example of the use of the free energy data, we
will calculate the dissolved CrF, concentration suffi-
cient to halt the dissolution of chromium from Hastel-
loy N if no region of lower chromium potential can be
developed as a sink.

_ For the reaction

Cr°(s) + 2UF.(d) = CrF, (d) + 2UF5(d),
AG = —152.06 — 2[—392.52 — (—300.88)]
= 31.22 keal,
~AG°
2.3RT/1000 4233
‘logK = log CrF, — log Cr° — 2 log (U*/U%") .

log K

If we assume U*/U* ~ 100 and note that the
chromium concentration in Hastelloy N is about 0.08
mole fraction (log Cr® = —1.10).

log CrF, =7.39 + (—1.10)
42X 20=-449=10g(3.2X 107%) .

A mole fraction of 3.2 X 1073 corresponds to a weight
concentration of 52 X 3.2 X 107°/40 = 42 ppm Cr** in
solution.

To obtain a higher concentration of dissolved Cr",
the solution would have to be more oxidizing. Further-
more, the Hastelloy N surface during operation be-
" comes depleted in chromium, and a chromium sink of
lower activity, Cr3C, (equivalent to a mole fraction of
-about 0.016 to 0.01), may be formed; all this would
require a somewhat more oxidizing regime to hold even
this much Cr?* in solution.

The free energy data can be used to estimate the
quantities in solution only when the species shown are
dominant. Thus it is shown by Ting, Baes, and

15

Mamantov® that under conditions of moderate concen-
trations of dissolved oxide, pentavalent niobium exists
largely as an oxyfluoride, which may be stable enough
for this rather than NbF,4 to be the significant dissolved
species under MSRE conditions.

The stability of the various fluorides below chromium
in the tabulation are such as to indicate that at the
redox potential of the U**/U% couple, only the
elemental form will be present in appreciable quantity.

In particular, tellurium® vapor is much more stzib»le,
than any of its fluoride vapors. Unless a more stable’
species than those listed in the table exists in molten
salt, these data indicate that tellurium would exist in -
the salt as a dissolved elemental gas or as a telluride ion.
[No data are available on Te,(g), etc., but such
combinations would not much affect this view.]

References

1. W. R. Grimes, “Molten Salt Reactor Chemistry,”
Nuel Appl 8,137-55(1970).

2. R. E. Thoma, Chemical Aspects of MSRE Opera-
tions, ORNL-4658 (December 1971).

3. C. F. Baes, Jr., “The Chemistry and Thermody-
namics of Molten Salt Reactor Fuels,” Nucl Met. 15,
617—44 (1969} USAEC CONF-690801).

4. K. A. Romberger, J. Braunstein, and R. E. Thoma,
“New Electrochemical Measurements of the Liquidus in
the LiF-BeF, System — Congruency of Li,BeF,,” J.
Phys. Chem. 76, 1154—59 (1972).

5. G. Ting, C. F. Baes, Ir., and G. Mamantov, “The

Oxide Chemistry of Niobium in Molten LiF-BeF,.

Mixtures,” MSR Program Semiannu. Progr. Rep. Feb.
29, 1972, ORNL-4782, pp. 87-93.

6. Free energies for tellurium fluorides given in Table
4.1 were taken from P. A. G. O’Hare, The Thermody-

namic Properties of Some Chalcogen Fluorides,

ANL-7315 (July 1968).
16

5. INVENTORY

Molten-salt reactors generate the full array of fission
products in the circulating fuel. The amount of any
given nuclide is constantly changing as a result of
concurrent decay and generation by fission. Also,
certain fission product elements, particularly noble
gases, noble metals, and others, may not remain in the
salt because of limited solubility.

For the development of information on fission
product behavior from sample data, each nuclide of
each sample must be (and here has been) furnished with
a suitable basis of comparison calculated from an
appropriate model, against which the observed values
can be measured. The most useful basis is the total
inventory, which is the number of atoms of a nuclide
which are in existence at a given time as a result of all
prior fissioning and decay. It is frequently useful to
consider the salt as two parts, circulating fuel salt and
drain tank salt, which are mixed at stated times. It is
then convenient to express an inventory value as
activity per gram of circulating fuel salt, affording for
salt samples a direct comparison with observed activity
per gram of sample.

For deposits on surfaces, it is useful to calculate for
comparison the total inventory activity divided by the
total surface area in the primary system.

Some of the comparisons for gas samples will be
based on accumulated inventory values, and others on
production rate per unit of purge gas flow. These
models will be developed in a later section.

In the calculation of inventory from power history,
we have in most cases found it adequate to consider the
isotope in question as being a direct product of fission,
or at most having only one significant precursor. For
the nuclides of interest, it has generally not appeared
necessary to account for production by neutron absorp-
tion by lighter nuclides. These assumptions permit us to
calculate the amount of nuclide produced during an
interval of steady relative power and bring it forward to
a given point in real time, with unit power fission rate
and yield as factorable items.

In Table 5.1 we show yield and decay data used in
inventory calculations. In the case -of ''°™Ag and
134Cs, neutron absorption with the stable element of
the lighter chain produced the nuclide, and special
calculations are required.

The branching fraction of !2°Sb to '2°Te is a
factor in the net effective fission yield of '2°""Te. The
Nuclear Data Sheets are to be revised! to indicate that
this branching fraction is 0.157 (instead of the prior
literature value of 0.36). All our inventory values and

calculations resulting from them have been proportion-
ately altered to reflect this revision.

The inventories for 235U operation were calculated
by program FISK? using a fourth-order Runge-Kutta
numerical integration method.

Differential equations describing the formation and
decay of each isotope were written, and time steps were
defined which evenly divided each period into segments
adequately shorter than the half-lives or other time
constants of the equation. The program FISK was
written in FORTRAN 3 and executed on a large-scale
digital computer at ORNL. Good agreement was ob-
tained with results from parallel integral calculations.

The FISK calculation did not take into account the
ingrowth of 23°Pu during the operation with 235U
fuel. The effect is slight except for ! °® Ru. We obtained
values taking this into account in separate calculations
using the integral method.

For the many samples taken during operation with
233U fuel, a one- or two-element integral equation
calculation® was made over periods of steady power,
generally not exceeding a day. Because the plutonium
level was relatively constant (about 500 g total) during
the 233U operation, weighted yields were used, as-
suming? that, of the fissions, 93.5% came from 233U,
2.2% from 235U, and 4.3% from 23°Pu.

For irradiation for an interval ¢, at a fission rate ¥
and yield Y, followed by cooling for-a time 7,, the
usual expressions® for one- and two-element chains are

Fission—>A =B -~

FY
Atoms A(z;) =T (1 —e™1f1)e Ml

1
FYA A, ( Ay

A2

l -e

Al

Atoms B(t;) = e M2

1 —e M2l

3‘7’\2’2)

where A, and A, are decay constants for nuclides A and
B.

A program based on the above expressions was
written in BASIC and executed periodically on a
commercial time-sharing computer to provide a current
inventory basis for incoming radiochemical data from
recent samples.

To remain as current as possible, the working power
history was obtained by a daily logging of changes in

A
17

Table 5.1. Fission product data for inventory calculations

Chain Isotope Half-life

Fraction

Cumulative {fission yield?

233y 235y 239p,

89 Sr 52 days 1 5.86 4.79 1.711
90 Sr 28.1 years 1 643 5.717 2.21
91 Sr 9.67 hr 1 5.57 5.81 243
91 Y 59 days (1.0¢ 5.57 5.81 243
95 Zr 65 days 1.0 6.05 6.20 497
95 Nb 35 days 1.0) 6.05 6.20 4.97
99 Mo 67 hr 1.0 4.80 6.06 6.10
103 Ru 39.5 days 1.0 1.80 3.00 5.67
106 Ru 368 days 1.0 0.24 0.38 4.57
109 Ag Stable (91 b + resonance) 0.044 0.030 1.40

110 Ag(m) 253 days

111 Ag 7.5 days 1 0.0242 0.0192 0.232

125 Sb 2.7 years 1 0.084 0.021 0.115
127 Te(m) 100 days. 0.22 0.60 0.13 0.39
129 Te(m) 34 days 0.36% 2.00 0.80 2.00
132 Te 3.25 days 1.0 4.40 4.24 5.10
131 I 8.05 days 1.0 2.90 2.93 3.78
133 Cs Stable (32 b + resonance) 5.78 6.61 6.53

134 Cs 750 days

137 Cs 29.9 years 1 6.58 6.15 6.63
140 Ba 12.8 days 1 540 6.85 5.56
141 Ce 32.3 days ] 6.49 6.40 5.01
"144 Ce 284 days 1 461 5.62 3.93
147 Nd 11.1 days 1 198 . 236 2.07
147 Pm 2.65 years 1 1.98 2.36 2.07

_ ?From M. J. Bell, Nuclear Transmutation Data, ORIGEN Code Library; L. E. McNeese,

. Engineering Development Studies for Molten-Salt Breeder Reactor Processing No. 1,
“ - ORNL-TM-3053, Appendix A (November 1970).

This is the value given in the earlier literature. The revised Nuclear Data Sheets will

indicate that the branching fraction is 0.157.

c R .- R
Parentheses indicate nominal values.

reactor pbWer indicated by nuclear instrumentation
charts; -the- history so obtained agreed adequately with
other determinations.

In practice, the daily power log was processed by the
computer to yield a power history which could remain
stored in the machine. A file of reactor sample times
and fill and drain times was also stored, as well as
fission product chain data. It was thus possible to
update and store the inventory of each nuclide at each
sample time. A separate file for individual samples and
their segments containing the available individual nu-
clide cou’hts_ .and counting dates was then processed to
give corrected nuclide data on a weight or other basis
and, using the stored inventory data, a ratio to the
appropriate reactor inventory per unit weight. The data
for individual salt and gas samples were also accumu-
lated for inclusion in a master file along with pertinent
reactor operating parameters at the time of sampling.
This file was used in preparing many tables for this
report.

Only one ad hoc adjustment was made in the
inventory calculation. Radiochemical analyses in con-
nection with the chemical processing of the salt to
change from 235U to 233U fuel indicated that °5Nb,
which had continued to be produced from the °5Zr in
the fuel when run 14 was shut down, was entirely
removed from the salt in the reduction step in which
the salt was treated with zirconium metal and filtered.
To reflect this and provide meaningful ®SNb inventories
for the next several months, the calculated °5Nb
inventory was arbitrarily set at zero as of the time of
reduction. This adjustment permitted agreement be-
tween inventory and observation during the ensuing
interval as ®SNb grew back into the salt from decay of
the 3 Zr contained in it.

In this report we have normally tabulated the activity
of each nuclide per unit of sample as of the time of
sampling and also tabulated the ratio of this to
inventory. For economy of space we then did not
tabulate inventory; this can of course be calculated by
dividing the activity value by the ratio value.

References

1. D. 1. Hore_fi (ORNL Physics Division), “Decay of
1298, '2°meTe " letter to J. R. Tallackson (ORNL
Reactor Division), May 5, 1972.

18

2. E. J. Lee (ORNL Mathematics Division), Program
FISK, June 18, 1968.

3. J. M. West, “Calculation of Nuclear Radiation,”
pp. 7-14, 7-15 in sect. 7-1, Nuclear Engineering Hand-
book, ed. by H. Etherington, McGraw-Hill, New York,
1958.

4. B. E. Prince, “Long-Term Isotopic Changes and
Reactivity Effects during Operation with 233U’ MSR
Program Semiannu. Progr. Rep. Feb. 28, 1969, ORNL-
4396, pp. 35-37. | | ‘
19

6. SALT SAMPLES

6.1 Ladle Samples

Radiochemical analyses were obtained on salt samples
taken from the pump bowl beginning in run 6, using the
sampler-enricher’ * (Fig. 6.1). A tared hydrogen-fired
copper capsule (ladle, Fig. 6.2) which could contain 10
g of salt was attached to a cable and lowered by
windlass past two containment gate valves down a
slanted transfer tube until it was below the surface of
the liquid within the mist shield in the pump bowl,
After an interval the capsule was raised above the latch,
until the salt froze, and then was raised into the upper
containment area and placed in a sealed transport
container and transferred to the High Radiation Level
Analytical Laboratory. Similar procedures were fol-
lowed with other types of capsules to be described
tater. The various kinds of capsules had hemispherical
ends and were ¥, -in.-diam cylinders, 6 in. or less in
length. Ladles were about 3 in. long.

After removal from the transport container in the
High Radiation Level Analytical Laboratory, the cable

was slipped off, and the capsule and contents were
inspected and weighed. The top of the ladle was cut off.
After this, the sample in the copper ladle bottom part
was placed in a copper containment egg and agitated 45
min in a pulverizer mixer, after which the powdered salt
was transferred (Fig. 6.3) to a polyethylene bottle for
retention or analysis. Data from 19 such samples, from
run 6 to run 14, are shown in Tables 6.1 and 6.2 as
ratios to inventory obtained from program FISK. Full
data are given in Table 6.7, at the end of this chapter.

There will be further discussion of the results.
However, a broad overview will note that the noble-gas
daughters and salt-seeking isotopes were generally close
to inventory values, while the noble-metal group was
not as high, and values appeared to be more erratic.
Noble-metal nuclides were observed to be strongly
deposited on surfaces experimentally exposed to pump
bowl gas. This implied that some of the noble-metal
activity observed for ladle salt sampies could have been
picked up in the passage through the pump bowl gas

ORNL-DWG 63-5848R

—~REMOVAL VALVE AND

[ S P
V/ 7 N /;_-_:9,.‘»-?/j SHAFT SEAL
7 /‘ // //‘f,»i’-_'" 7 //A “L=-PERISCOPE
CAPSULE DRIVE UNIT — - s /WTI.‘.?F’“‘F 4 7 '_-LIGHT

LATCH - ./////iE

ACCESS PORT 1 // 7T

AREA {C
(PRIMARY CONTAINMENT) - T /////“ J‘f

SAMPLE CAPSULE -~ '//

OPERATIONAL AND /7
MAINTENANCE VALVES <

SPRING CLAMP | 1
DISCONNECT — -~ 77 7

TRANSFER TUBE
(PRIMARY CONTAINMENT )~u,_

LATCH STOP -

’/;;i‘ \\\-\ »_-_. \ 2.
Dt |
b MIST SHIELD
CAPSULE GLIDE

_*-{-E E ‘{O*_

Zansy rom
7R Y [H by — il | &
s < AN : ! =L

/ Eyan [ T ‘ o i
e ||

CASTLE JOINT (SHIELDED
/WITH DEPLETED URANIUM)

~—AREA 3A (SECONDARY
CONTAINMENT )

“S— SAMPLE TRANSPORT
CONTAINER

: =i —
Ll j: \\;- == =
; - MANIPULATOR

== AREA 28 (SECONDARY
CONTAINMENT)

/ . CRITICAL CLOSURES
REQUIRING A BUFFERED
SEAL
1 o ! 2

FEET

Fig. 6.1. Sampler-enricher schematic.
20

PHOTO 63984

Fig. 6.2. Container for sampling MSRE salt.

PHOTO 62746

Fig. 6.3. Apparatus for removing MSRE salt from pulverizer-mixer to polyethylene sample bottle.

21

Table 6.1. Noble-gas daughters and salt-seeking isotopes in salt samples
from MSRE pump bowl during uranium-235 operation

Expressed as ratio to amount calculated for 1 g of inventory salt at time of sampling

Noble-gas daughters

Salt-seeking isotopes

Sample Date :
Sr-89 Sr-91 Sr-92 Ba-140 Cs-137 Ce-141 Ce-143 Ce-144 Nd-147 Zr-95
6-17L 5-23-66 0.92 0.81 0.76 0.84
6-19L 5-26-66 0.90 0.63 0.69 0.85
7-07L 6-27-66 0.67 0.63 0.58 0.92 0.79
7-10L 7-6-66 0.67 0.71 0.90 0.95 0.82
7-12L 7-13-66 0.73 0.72 0.89 0.63 0.78 1.20 1.12
8-05L 10-8-66 0.64 0.80 0.77 1.07 0.95
10-12L 12-28-66 0.80 0.71 1.30 0.66 1.04 0.95
10-20L 1-9-67 0.74 0.72 0.59 041 3.50 0.71
11-08L 2-13-67 0.66 0.71 0.69 0.85 1.09
11-12L 2-21-67 0.80 0.82 0.79 0.90 0.98
11-22L 3-9-67 0.91 1.80 1.09
1145L 4-17-67 0.77 0.89 0.23
11-51L 4-28-67 0.69 1.10 0.96 0.87 0.64 1.20 0.86
11-52L 5-1-67 0.69 1.10 0.96 0.42 1.09 0.96
11-54L 5-5-67 ‘ 0.94
11-58L 5-8-67 0.88 1.01 1.03 2.60 1.10 1.10
12-06L 6-20-67 0.76 0.83 1.06 1.04
12-27L 7-17-67 0.75 0.97
14-22L 11-7-67 9.60 0.84 1.02
14-20FV 11-4-67 0.89 0.99 0.76 1.04 1.04
14-30FV 12-5-67 0.87 0.77 0.77 0.55 1.06
14-63FV 2-27-68 0.87 1.14 0.59 1.20 1.12
14-66FV 3-5-68 0.90 45.00 1.26 2.40 0.94
Table 6.2. Noble metals in salt samples from MSRE pump bowl during uranium-235 operation
Expressed as ratios to amount calculated for 1 g of inventory salt at time of sampling
Sample Date Nb-95 Mo-99 Ru-103 Ru-105 Ru-106  Ag-111 Te-129m Te-132 1131 1-133 I-135
6-17 5-23-66 0.57 0.01330 2.50 0.57 0.72 091 0.55
6-19 - 5-26-66 2.77 0.42 9.31 0.51 092 069 0.83
7-07 6-27-66 0.58 0.09086 1.33 0.44 0.81 0.69 0.66
7-10 7-6-66 0.80 0.21 3.77 0.40 0.79 091
7-12 7-13-66 15.51 0.19 0.20 1.44 0.29 0.31 0.33 0.75 0.73 0.64
8-05 10-8-66 2.66 0.03767 0.06205 0.08089 1.10
10-12 12-28-66 0.44 0.02216 0.02659 0.03435 0.12 0.14 091
10-20 1-9-67 0.95 0.28 0.01551 0.01994 0.17 0.17 0.96
11-08 2-13-67 0.03324 1.88 0.12 0.09972 0.47 0.70
11-12 2-21-67 0.30 1.44 0.09972 0.09972 0.31 0.94
11-22 3-9-67 1.03 0.06648 0.07756 0.42 1.33
1145 4-17-67 0.32 0.92 0.21 0.16 0.09972 0.31 1.09
11-51 4-28-67 0.04432 0.44 0.05540 0.03324 0.12 0.17 0.14 098
11-52 5-1-67 0.02216 0.49 1.22 0.08864 0.09972 0.17 0.14 0.96
11-54 5-5-67 0.19 0.21 0.02216 © 0.02216 0.08864 0.82
11-58 5-8-67 0.24 0.03324 0.13 0.12 0.03324 0.17 0.12 0.16
12-06 6-20-67 0.89 0.09972 0.13 0.68
12-27 7-17-67 0.38 0.75 0.12 0.08864 0.17 0.12 0.99
14-22 11-7-67 = 0.00111 0.47 0.06648 0.07756 0.11 0.06648 8.20
14-20FV  114-67 0.00066 0.01440 0.00222 0.00665 0.04432 0.00665 0.74
14-30FV  12-5-67 0.00001 0.00554 0.00111 0.00222 0.01219 0.01219 0.50
14-63FV  2-27-68 0.00003 0.00222 0.00044 0.00078 0.00222 0.61
14-66FV  3-5-68 0.02216 0.00443 0.00033 0.00332 0.01219 -

and transfer tube regions, and indicated that salt
samples taken from below the surface were desirable.

6.2 Freeze-Value Samples

Beginning in run 10, gas samples (q.v.) had been taken
using a “‘freeze-value” capsule (Fig. 6.4).

To prepare a freeze-valve capsule, it was heated
sufficiently to melt the salt seal, then cooled under
vacuum. It was thus possible to lower the capsule
nozzle below the surface of the salt in the pump bowl
before the seal melted; the vacuum then sucked in the
sample.

After the freeze-valve capsule was transferred to the
High Radiation Level Analytical Laboratory, inspected,
and weighed after removing the cable, the entry nozzle
was sealed with chemically durable wax. The capsule
exterior was then leached repeatedly with “verbocit”

ORNL-DWG 67-4784A

STAINLESS STEEL
CABLE

|« —— ¥4-in, OD NICKEL

N

20cc

—

VOLUME é
!

NICKEL CAPILLARY

Fig. 6.4. Freeze valve capsule.

22

(Versene, boric acid, and citric acid) and with
HNOQO,-HF solution until the activity of the leach
solution was acceptably low. The capsule was cut apart
in three places — in the lower sealing cavity, just above
the sealing partition, and near the top of the capsule,
Salt was extracted, and the salt and salt-encrusted
capsule parts were weighed. The metal parts were
thoroughly leached or dissolved, as were aliquots of the
salt.

Four samples (designated FV) were taken late in run
14 using this technique. Results shown near the bottom
of Tables 6.1 and 6.2 show that the values for
salt-seeking isotopes and daughters of noble-gas isotopes
were little changed and were near inventory, but values
for noble-metal nuclides were far below inventory. This
supports the view that the liquid salt held little of the
noble metals and that the noble-metal activity of earlier
ladle samples came from the pump bowl gas (or
gas-liquid interface) or from the transfer tube.

After run 14 was terminated the 35U fuel was
removed by fluorination, and the carrier salt was
reduced with hydrogen and with metallic zirconium,
after which 233U fuel was added, and the system was
brought to criticality and then to power in the early
autumn of 1968.

Radiochemical analyses were obtained on ladle sam-
ples taken during treatment in the fuel storage tank
(designated FST) during chemical processing, and from
the fuel pump bowl after the salt was returned to the
fuel circulation system (designated FP) from time to
time during runs 15, 17, 18, and 19, as shown in Table
6.3. Chemical analyses on these samples were reported
by Thoma.?

The 95Nb activity of the solution was slightly more

than accounted for in samples FP15-6L, as the zir-
conium reduction process had been completed only a

short time before; the niobium inventory was set at
zero at that time. The °3Nb which then grew into the
salt from decay of ®°Zr appeared in these samples to
show some response to beryllium reduction of the salt,
though this effect is seen better with freeze-valve
samples and so will not be discussed here, The various
additions of beryllium to the fuel salt have been given
by Thoma.?

Data for all freeze-valve salt samples taken during
233U operation are summarized as ratios to inventory
salt in Tables 6.4 and 6.5, and Table 6.8 at the end of
the chapter, where various operating conditions are
given, along with the sample activity and ratio to
inventory salt. Analyses for salt constituents as well as
fission products are shown there. On the inventory-ratio
basis, comparisons can be made between any con-
stituents and/or fission products.
Table 6.3. Data on fuel (including carrier) salt samples from MSRE pump bowl! during uranium-233 operation

Ladle capsules

Values shown are the ratio of observed activity to inventory activity, both in disintegrations per minute per gram

Inventory basis, 7.4 MW = full power

Sample Type Sr-89  Sr-91 Y91 Ba-140 Cs-137 Ce-141 Ce-143 Ce-144 Nd-147 Zr-95 Nb-95 Mo-99 Ru-103 Ru-105 Ru-106 Te-129m Te-132 1-131
FST-25, Fuel, pre F, 0.85 1.22 1.22 1.02 0.66
Aug. 14
FST-27, Carrier, end F, 0.83 1.21 099 0171
Aug, 21 :
FST-30, Carrier, end H, 0.94 1.28 1.11  0.50
Sept. 4
FP156L, Fuel 0.92 1.32 1.11  (0.26)*%
Sept. 14
FP159L, Fuel 0.93 1.07 1.26 1.30
Sept. 17
FP15-10L, Fuel 0.93 1.28 1.04 1.02
Sept. 19
FP15-18L, Fuel 0.75 1.04 0.80 0.71
Oct. 4
FP15-26L, Fuel 0.89 1.40 1.04 0.76
Oct. 10
FP17-1L, Fuel, pre power (3.4) 0.84 1.32 1.05 0.63
Jan. 12 )
FP17-4L, Fuel, approach 0.85 0.37 0.82 1.09 0.80 0.44 0.76 1.72 0.25
Jan. 21 power
FP17-9L, Fuel 0.86 1.15 0.81 0.43 0.10 0.28
Jan. 24
FP17-12L, Fuel 1.12  0.26 0.43 0.15 0.04 0.14 1.07
Feb. 6 :
FP17-18L, Fuel 1.17 0.22
Feb. 12
FP17-19L, Fuel 1.22 (0.11)
Feb. 19
FP17-20L, Fuel 1.01 0.11
Feb. 20
FP17-30L, Fuel 1.06 0.46
Apr, 1
18-1L, Fuel 1.06 0.53
Apr. 14
18-5L, Fuel 1.25 0.44
Apr. 23
18-10L, Fuel 1.44 0.15
Apr, 29 i
19-17L, Fuel 1.01 1.36 1.17 0.49 0.94 1.24 1.46 2.18 144 0.45 0.06 1.45 10.7 5.7
Aug. 27

495 Nb removed by fluorination.
bparenthescs indicate approximate value.

€T
Tables 6.4 and 6.5 present only the ratio of the
activity of various fission products to the inventory
value for the various samples.

Two kinds of freeze-valve capsules were used. During
runs 15, 16, and 17 (except 17-32) the salt-sealed
capsule described above was used. In general, the results
obtained with this type of capsule are believed to
represent the sample fairly. However, as discussed
above, the values for a given salt sample represent the
combination of activities of the capsule interior surface

24

with those determined for the contained: salt. To
prevent interference from activities accumulated on the
capsule exterior, as many as several dozen HNO;-HF
leachings were required; occasionally the capsule was
penetrated. Also, the salt seal appeared to leak slightly;
less vacuum inside resulted in less sample.

6.3 Double Wall Capsule

When a double-walled capsule was developed for gas
samples (q.v.) it was adopted also for salt samples. The

Table 6.4. Noble-gas daughters and salt-seeking isotopes in salt samples from MSRE pump bowl during uranium-233 operation

Expressed as ratio to amount calculated for 1 g of inventory salt at time of sampling

Noble-gas daughters

Salt-seeking isotopes

Sample Date

Sr-89 Sr-90 Y91 Ba-140 Cs-137 Ce-141 Ce-144 Nd-147 Zr-95
15-28 10-12-68 0.20 1.36 0.29 0.28
15-32 10-15-68 0.94 0.61 0.84 1.08 0.91
15-42 10-29-68 0.84 0.82 1.16 0.88
15-51 11-6-68 0.79 0.70 0.81 1.13- 1.14
15-57 11-11-68 0.87 0.93 1.25 0.98
15-69 11-25-68 0.84 1.06 0.93
164 12-16-68 1.01 0.89 1.24 1.17 0.69 1.10 1.38
17-2 1-14-69 0.69 0.71 0.66 (1.42)4 0.74 1.10 0.79
17-7 1-23-69 0.48 0.11 0.96 0.80 0.68 0.53 0.70 0.66 0.72
17-10 1-28-69 0.60 0.76 1.22 0.69 0.83 0.94 0.62 0.98 0.89
17-22 2-28-69 0.55 1.31 0.38 0.09776 0.80 1.07 0.99 0.89
17-25 3-26-69 0.77 1.22 1.08 0.91 0.85 1.28 1.22 0.98
17-31 4-1-69 0.63 2.53 0.64 1.05 0.81 0.77 1.11 1.08 0.92
17-32 4-3-69 0.76 0.94 1.01 0.77 0.80 1.16 1.16 0.96
18-2 4-14-69 0.77 1.70 0.92 0.84 091 1.22 1.49 0.97
184 4-18-69 0.78 1.04 1.10 0.88 0.78 1.21 1.15 0.99
18-6 4-23-69 0.60 0.81 0.81 0.71 1.03 0.79
18-12 5-2-69 0.97 1.34 0.72 0.81 1.23 0.94
18-19 5-9-69 0.62 1.38 1.14 0.92 0.81 1.10 0.65 1.01
18-44 5-29-69 0.76 1.11 1.01 0.79 0.78 0.94 0.04148 0.95
18-45 6-1-69 0.75 1.12 1.06 0.91 0.81 1.02 1.20 0.95
1846 6-1-69 0.72 1.02 1.04 0.84 0.75 1.23 1.03 0.90
19-1% 8-11-69 0.00863 0.00389 0.00367 0.02514 0.00176 0.00984 0.00286
19-6 8-15-69 0.01677 0.01612 0.02157 0.09981 0.15 0.02149 0.01439 0.01015
19-9 8-18-69 0.70 0.39 0.85 0.52 0.60 091 0.18 0.72
19-24 9-10-69 0.89 1.37 0.19 0.85 0.75 1.00 0.12 0.86
19-36 9-29-69 0.82 1.00 0.94 1.10 0.76 1.07 0.94 0.86
19-42 10-3-69 0.86 1.12 0.98 1.36 0.84 1.12 1.21 0.93
19-44 10-6-69 0.78 1.21 0.99 0.08140 0.82 1.09 1.18 1.01
19-47 10-7-69 0.89 1.24 0.98 0.80 0.86 1.15 1.20 0.99
19-55 10-14-69 0.77 1.00 1.12 0.69 0.90 1.22 1.34 0.95
19-57 10-17-69 0.75 1.13 1.10 0.85 0.87 1.16 1.22 0.93
19-58 10-17-69 0.73 1.17 1.03 0.86 0.81 1.17 1.36 0.92
19-59 10-17-65 0.65 2.23 1.10 0.80 0.79 1.06 1.16 0.90
19-76 10-30-69 0.73 0.76 1.04 0.98 0.85 1.16 1.17 0.74
20-1 11-26-69 0.63 1.29 1.07 0.75 0.67 1.13 0.92
20-19 12-5-69 0.53 1.02 0.91 0.79 0.69 1.06 0.85

aApproximate value.
bFlush salt.
interior copper capsule was removed without contact
with contaminated hot-cell objects and was entirely
dissolved. The outer capsule could also be dissolved to
determine the relative amounts of activity deposited on
such a “dipped specimen.” Data on capsule exteriors
will be given in a separate section. Salt samples
beginning with sample 17-32 were obtained using the

25

double-walled capsule. Operating conditions associated
with the respective samples are summarized in Table
6.6.

We should note” that very little power had been
produced from 223U prior to sample 15-69; much of
the activity was carried over from 233U operations.
Sample 17-2 was taken during the first approach to

Table 6.5. Noble metals in salt samples from MSRE pump bowl during uranium-233 operation

Expressed as ratio to amount calculated for 1 g of inventory salt at time of sampling

Sb-125

Sample Date Nb-95 Mo-99 Ru-103 Ru-106 Ag-111 Te-129m  Te-132 I-131
15-28 10-12-68 0.74 0.02536 0.03436 0.14

15-32 10-15-68. 1.16 0.00006  0.00015 0.00007

1542 10-29-68 0.72 0.00096  0.00084 0.00391

15-51 11-6-68 0.85 0.00037  0.00026 0.00011

15-57 11-11-68 1.06 0.00433  0.00329 0.00195

15-69 11-25-68 0.02000 0.00004

16-4 12-16-68 0.54 0.09635  0.01241  0.00442 0.00687 0.09176 1.13
17-2 1-14-69 0.52 (2.48)4 0.04953  0.00304 (2.57) 0.26 (3.67) (1.67)
17-7 1-23-69 0.29 0.12 0.03352  0.00165  0.09095 0.21 0.31 0.40
17-10 1-28-69 ~0.02312%  0.53 0.01134  0.00055 0.01981  0.03021 0.37
17-22  2-28-69 —-0.05000 0.00971  0.00076  (.00027  0.00425 0.00233  0.02040 0.46
17-29 3-26-69 0.51 0.00445  0.02199 0.01014  0.01169 0.28
17-31 4-1-69 0.32 0.01360  0.00177 0.13 0.01921  0.03794 0.30
17-32  4-3-69 0.31 0.00344  0.02216 0.08057 0.00348  0.40
18-2 4-14-69 0.46 0.01496 (.52 0.23 0.03367 0.00670  0.10
184 4-18-69 0.22 0.00839 0.00398 0.35
18-6 4-23-69 1.56 0.85 0.07401  0.02958 1.23 1.83 1.80 0.26
18-12  5-2-69 0.04847 0.00812  0.00160 0.06829  0.00383  0.00230 0.59
18-19 59-69 0.01641 0.00773  0.00202 0.06622 0.74
1844 5-29-69 —0.03579 0.03459  0.00153 0.05970 0.00813 0.34
18-45 6-1-69 0.37 1.36 0.13 0.05106 0.14 0.44
18-46  6-1-69 -0.02242 0.12 0.00862  0.00327 0.03976 0.02532  0.04302 0.08227
19-1¢€ 8-11-69 0.11) (0.07066) (0.02304) (0.10) (0.28)
19-6¢ 8-15-69 (—0.00108) (0.00308) (0.00192) (0.01185)

19-9 8-18-69 0.34 0.00071

19-24  9-10-69 0.33 0.22 0.21 0.23 0.74 0.24 0.60
19-36  9-29-69 —-0.08623 0.01916  0.00219 0.12 0.00335  0.00614  0.58
19-42 10-3-69 —0.06114 0.43 0.13 0.08871  0.20 0.18 0.09806  0.65
19-44 10-6-69 -0.05268 0.41 0.04484 0.01308 0.60 0.01808 0.89
19-47 10-7-69 0.02630 0.83 0.15 0.05326 0.10 0.19 0.06358 0.11
19-55 10-14-69 0.63 0.18 0.05759  0.04055 0.13 0.04204 0.44
19-57 10-17-69 0.05050 0.75 0.28 0.11 ' 0.30 0.07974 0.54
19-58 10-17-69 —0.03366 0.01885  0.00367  0.00207 0.00151 (.64
19-59 10-17-69 -0.01349 0.19 0.03270 0.01478 0.02334  0.00680 0.15
19-76 10-30-69 0.02995 0.01412  0.00195 0.00031 0.40
20-1 11-26-69 0.19 3.34 0.29 0.14 0.48 0.22 0.55 0.68
20-19 12-5-69 0.05389 0.78 0.11 0.05074 0.23 0.28 0.21 0.41

9Parentheses indicate approximate value.

bNegative numbers result when ?>Nb, which grows in from 957y present between sampling and analysis time, exceeds that found

by analysis.
€Flush salt.
Table 6.6. Operating conditions for salt samples taken from MSRE pump bowl during uranium-233 operation

Overflow
Sample . Equivalent  Percent of Hours at amp  Voids Pump bowl Previous Purge gas flow
No. Date Time  fuli-power full percent of pm %) level Lb/hr return Std G Pei
hours power power (%) : liters/min as S1g
Day Time

15-28 10-12-68 1726 0.01 0.00 0.5 1180 0.00 63 1.4 1012 0711 3.30 He 5.2 Purgeon
15-32 10-15-68 2047 0.01 0.01 0.1 1180 0.60 66 29 10-15 1830 3.30 He 5.5 Purgeon
15-42 10-29-68 1123 0.01 0.00 9.4 1180 0.00 66 3.7 10-28 1730 3.30 He 4.9  Purge on
15-5t 11-6-68 1531 0.01 0.00 24.3 1180 0.00 66 1.0 114 1630 3.30 He 5.0 Purgeon
15-57 11-11-68 2145 0.01 0.00 0.5 1180 0.60 63 0.8 11-11 1208 3.30 He 5.5 Purgeon
15-69 11-25-68 1700 0.01 0.00 21.4 1180 0.60 56 0.8 11-25 0425 3.30 He 5.0 Purgeon
16-4 12-16-68 0555 1.56 0.00 62.8 1180 0.60 62 0.8 12-15 1758 3.30 He 4.6 Purgeon
17-2 1-14-69 1025 1.69 5.63 1.5 1180 0.60 59 3.8 1-14 0310 3.30 He 3.8 Purgeon
17-7 1-23-69 1320 94.00 57.50 15.1 1180 0.60 57 1.8 1-23 0736 3.30, He 4.2  Purgeon
17-10 1-28-69 0603 155.50 58.75 39.0 1180 0.60 57 1.6 1-27 2315 3.30 He 5.0 Purgeon
17-22 2-28-69 2259 719.63 87.50 31.4 942 0.00 65 0.7 227 1630 3.30 He 54  Purge on
17-29 3-26-69 1506 1144.75 86.25 0.1 1050 0.05 58 1.3 3-25 2132 3.30 He 4.2  Purge on
17-31 4-1-69 1145 1271.00 90.00 140.8 1050 0.05 62 3.0 4-1 1015 3.30 He 8.9 Purge on
17-32 4-3-69 0552 1307.00 90.00 182.9 1050 0.05 60 4.5 4-2 1807 3.30 He 3.2 Purgeon
18-2 4-14-69 1150 1527.75 100.00 41.8 1180 0.60 61 7.4 4-14 0853 3.30 He 4.6 Purgeon
18-4 4-18-69 2119 1601.25 100.00 49.8 1180 0.60 63 4.9 4-18 1901 3.30 He 5.2 Purgeon
18-6 4-23-69 1015 1713.12 100.00 158.7 1180 0.60 63 4.7 4-23 0733 3.30 He 5.5 Purgeon
18-12 5-2-69 1305 1939.00 100.00 376.5 1180 0.60 59 3.0 5-2 0500 3.30 He 5.2 Purge on
18-19 5-9-69 1925 2106.00 100.00 93.8 1180 0.60 59 0.9 5-9 1303 3.30 He 4.8 Purgeon
18-44 5-29-69 0311 2473.00 86.25 28.7 990 0.00 53 0.0 5-28 1833 2.30 He 13.0  Purge on
18-45 6-1-69 0921 2538.63 0.00 0.4 990 0.00 50 0.0 5-31 2223 2.00 Ar 12.8  Purge on
18-46 - 6-1-69 1412 2538.63 0.00 5.2 990 0.00 56 0.0 5-31 2223 2.00 Ar 13.6  Purge on
19-14 8-11-69 0845 2538.63 0.00 0.0 1189 0.00 72 0.0 00 00 3.30 He 2.4  Purgeon
19-67 8-15-69 0413 2538.63 0.00 0.0 1170 0.00 62 0.7 00 00 3.30 He 5.3 Purgeon
19-9 8-18-69 0604 2538.63 0.13 1.1 1189 0.60 65 24 g-18 0219 3.30 He 5.3  Purge on
19-24 9-10-69 1049 2781.87 0.13 19.6 1165 0.70 62 4.7 9-10 0501 2.90 Ar 5.0 Purge off
19-36 9-29-69 1108 2978.50 68.75 66.1 608 0.00 58 0.9 9-27 0316 3.35 He 6.3  Purge off
19-42 10-3-69 1105 3048.87 87.50 49.0 1176 0.53 68 6.3 10-3 1036 3.30 He 5.0 Purge off
19-44 10-6-69 0635 3118.62 100.00 63.3 1188 0.53 64 7.4 10-3 0305 3.30 He 5.5 Purge off
19-47 10-7-69 1033 3148.75 100.00 913 1175 0.53 61 1.8 10-7 0228 3.35 He 5.8  Purge off
19-55 10-14-69 1047 3330.25 100.00 259.5 1186 0.53 63 2.6 10-14 0353 3.30 He 5.2 Purge off
19-57 10-17-69 0620 3395.38 100.00 42.6 1186 0.53 63 3.7 10-17 0105 3.30 He 5.2 Purge off
19-58 10-17-69 0941 3397.13 0.13 1.0 1188 0.53 67 9.0 10-17 0757 3.30 He 5.6  Purge off
19-59 10-17-69 1240 3397.13 0.13 4.0 1189 0.53 65 39 10-17 0757 3.30 He 5.6  Purge off
19-76 10-30-69 1159 3705.63 100.00 140.6 1176 0.53 66 8.6 10-30 0916 3.30 He 5.2 Purge off
20-1 11-26-69 1704 3789.38 100.00 1.7 1190 0.53 64 6.8 11-26 1320 3.30 He 5.5 Purge off
20-19 12-5-69 0557 3990.87 100.00 36.7 1200 0.53 63 5.6. 12-5 0248 3.30 He 5.2 Purge off

~

2F1lush salt.

9¢
sustained high power; the higher values for the shortest-
lived nuclides reflect some uncertainties in inventory
because heat-balance calibrations of the current power
level had not been accomplished at the time — it
appeared more desirable to accept the inventory aberra-
tion, significant only for this sample, than to guess at
correction.

Sample 18-46 was taken 5% hr after a scheduled
reactor shutdown. Samples 19-1 and 19-6 are samples
of flush salt circulated prior to returning fuel after the
shutdown.

6.4 Fission Product Element Grouping

[t is useful in examining the data from salt samples to
establish two broad categories: the salt-seeking elements
and the noble-metal elements. The fluorides of the
salt-seeking elements (Rb, Sr, Y, Zr, Cs, Ba, La, Ce, and
rare earths) are stable and soluble in fuel salt. Some of
these elements (Rb, Sr, Y, Cs, Ba) have noble-gas
precursors with half-lives long enough for some of the
noble gas to leave the salt before decay.

Noble-metal fission product elements (Nb, Mo, Tc,
Ru, Rh, Pd, Ag [Cd, In, Sn?], Sb, Te, and I) do not
form fluorides which are stable in salt at the redox
potential of the fuel salt. Niobium is borderline and will
be discussed later. lodine can form iodides and remain
in the salt; it is included with the noble metals because
most iodine nuclides have a tellurium precursor — and
also to avoid creating a special category just for iodine.
The Nb-Mo-Tc-Ru-Rh-Pd-Ag elements for a subgroup,
and the Sb-Te-I elements another.

Quite generally in the salt samples the salt-seeking
elements are found with values of the ratio to inventory
activity not far from unity. Values for some nuclides
could be affected by loss of noble-gas precursors. These
include: '

Precursor Nuclide affected
3.18-min 3°Kr 89¢r
33-sec °OKr 9081‘
9.8-sec ° 'Kr olgp, 9y
3.9-min 137Xe 1370
16-sec 140Xe 14OBa

The 2°Sr and !37Cs in particular might be expected
to be stripped to some extent into the pump bowl gas,
as discussed below for gas samples. In Table 6.4, ratio
values for °'Y and !*°Ba are close to unity and
actually slightly above.

Values for !*!Ce run slightly below unity, and those
for '**Ce (and '*7Nd) somewhat above. The *5Zr
values average near but a few percent below unity.

Thus the grou;‘) of salt-seeking elements offers no
surprises, and it appears acceptable to regard them as

27

remaining in the salt except as their noble-gas pre-
CUrsors may escape.

The general consistency of the ratio values for this
group provides a strong argument for the adequacy of
the various channels of information which come to-
gether in these numbers: sampling techniques, radio-
chemical procedures, operating histories, fission prod-
uct yield and decay data, and inventory calculations.

6.5 Noble-Metal Behavior

The consideration of noble-metal behavior is ap-
proached from a different point of view than for the
salt-seeking group. Thermodynamic arguments indicate
that the fluorides of the noble metals generally are not
stable in salt at the redox potential of MSRE opera-
tions. Niobium is borderline, and iodine can form
iodides, which could remain in solution. So the ques-
tions are: Where do the noble-metal nuclides go, how
long do they remain in salt after their formation before
leaving, and if our salt samples have concentrations
evidently exceeding such a steady state, how do we
explain it? The ratio of concentration to inventory is
still a good measure of relative behavior as long as our
focus is on events in the salt.

If the fluorides of noble-metal fission product ele-
ments are not stable, the insolubility of reduced
(metallic or carbide) species makes any extra material
found in solution have to be some sort of solid
substance, presumably finely divided. Niobium and
iodine — later tellurium — will be discussed separately,
as these arguments do not apply at one point or
another.

If we examine the data in Table 6.5 for Mo, Ru, Ag,
Sb, and Te isotopes during runs 15 to 20, it is evident
that a low fraction of inventory was in the salt. We
simply need to decide whether what we see is dissolved

-steady-state material or entrained colloidal particulate

material.

If the dissolved steady-state concentration of a
soluble material is low, relative to inventory, loss
processes appreciably more rapid than decay must exist.
If the average power during the shorter period required
to establish the steady state is f;, then at steady state it
may be shown that

fiFy=AQ+L).

It follows that the ratio of observed to inventory
activity will be:

obs A

Obs fi (recent period)
inv. (At+L)

T fi(l—e MmNz

all periods

The amounts in solution should be proportional to the
inverse of half-life, to the current power (vs full), and
to the relative degree of full-power saturation. The
amount does not depend on the other atoms of the
species as long as the loss term is first order.

It follows that samples taken at low power after
operation at appreciable power should drop sharply in
value compared with the prior samples. These include
1845, 18-46, 19-24, 19-58, and 19-59. Of these
samples, only 19-58 appears evidently low across the
board; the criterion is not generally met.

The expression also indicates that after a long
shutdown, the rise in inventory occurring (for a half-life
or so), with fairly steady loss rate, should result in an
appreciable decrease in the ratio. The beginnings of runs
17 and 19 are the only such periods available. Here the
data are too scattered to be conclusive; some of the
data on '2°™Te and '?2Te appear to fit: samples 17-7
and 19-24, respectively, are somewhat higher than
many subsequent samples.

Briggs* has indicated that the loss coefficients should
be

L~(12K + 57K + 7.1K) hr ™!

for mass transfer to graphite, metal, and bubble surface,
respectively, if sticking factors were unity and K the
ratio of the mass transfer coefficient to that of xenon.
For metal atoms, K ~ 1, neglecting bubbles, L ~ 7
hrt,

The ratio to inventory predicted above is dominated
by the first factor, A/(A + L), in the cases (a majority)
where the present power was comparable with the
average power for the last half-life or so. We can then
note for the various nuclides using L = 7 hr™':

Nuclide Half-life (days) MO+ N)
Mo 2.79 0.0015

103Ru 39 6 0.00010
106py 367 0.00001
I1hag 7.5 0.00055

Comparing the observed ratios with these shows that
what we observed in essentially all cases was an order of
magnitude or more greater. This indicates that the
observed data have to be accounted for by something
other than just the steady-state dissolved-atom concen-
tration serving to drive the mass transfer processes.

The concept that remains is that some form of
suspended material contributed the major part of the
activity found in the sample. Because this would
represent a separate phase from the salt, the mixture

28

proportion could vary. The possible sources and be-
havior of such a mixture will be considered in a later
section after other data, for surfaces, etc., have been
presented. We believe that the data on noble-metal
fission products in salt are for the major part explicable
in terms of this concept.

Three elements included in the table of noble-metal
data should be considered separately: niobium, iodine,
and tellurium,

6.6 Niobium

Our information on niobium comes from the 35-day
°5Nb daughter of 65-day °5Zr. Thermodynamic con-
siderations given earlier indicate that at fission product
concentration levels, Nb*" is likely to be in equilibrium
with niobium metal if the redox potential of the salt is
set by U3 /U* concentration ratios perhaps between
0.01 and 0.001. If Nb3* species existed significantly at
MSRE oxide concentrations, the stability of the soluble
form would be enhanced. If NbC were formed at a rate
high enough to affect equilibrium behavior, then the
indicated concentration of soluble niobium in equi-
librium with a solid phase would be considerably
decreased.

Because ?3Nb is to be considered as a soluble species,
direct comparison with inventory is relevant in the
soluble case (when insoluble, it should exhibit a limiting
ratio comparable to 34-day '2° ™Te, or about 0.0001).

The data for ®>Nb in salt samples do appear to have
substantial ratio values, generally 0.5 to 0.3 at times
when the salt was believed to be relatively oxidizing.
When appreciable amounts of reducing agent, usually
beryllium metal, had been added, the activity relative to
inventory approached zero (20.05). Frequently, slightly
negative values resulted from the subtraction from the
observed niobium activity at count time of that which
would have accrued from the decay of °3Zr in the

- sample between the time of sampling and the time of

counting.

6.7 lodine

Iodine, exemplified by *3'I, is indicated to be in the
form of iodide ion at the redox potential of fuel salt,
with little I, being stripped as gas in the pump bowl.?
Thermodynamic calculations indicate that to strip 0.1%
of the 13! as I,, a U**/U3 of at least 10% would have
to exist. As far as is known, the major part of MSRE
operation was not as oxidizing as this (however, because
some dissociation to iodine atoms can occur in the
vapor, stripping could be somewhat easier). '3'I
activities relative to inventory were between 8 and
113%, with most values falling between 30 and 60%.
What happened to the remainder is of interest, It
appears likely that the tellurium precursor (largely
25-min !31Te) was taken from solution in the salt
before half had decayed to iodine, and of this tellurium,
some, possibly half, might have been stripped, and the
remainder deposited on surfaces. Perhaps half of the
1317 resulting from decay should recoil into the
adjacent salt. From such an argument we should expect
about the levels of ' 3'1 that were seen.

6.8 Tellurium

Tellurium is both important and to some extent
unique among the noble metals in that the element has
a vapor pressure at reactor temperature (650°C) of
about 13 torr. Since the fluorides of tellurium are
unstable with respect to the element, at the redox
potential of the fuel, we conclude that the gaseous
element and the tellurium ion are the fundamental
species. As a dissolved gas its behavior should be like
xenon. The mass transfer loss rate coefficients indicated
by Briggs* would apply, L ~ (1.2K + 5.7K + 7K), so
that with high sticking factors, about Y| 4-hr production
would be the steady-state concentration, and half the
tellurium would go to off-gas. The dissolved concentra-
tion for !'®2Te, relative to inventory salt, would be
about 0.0006, and for '2°Te, about 0.00006. Again it
is evident that the observations run higher than this.
Recent observations by C. E. Bamberger and J. P.
Young of ORNL suggest that a soluble, reactive form of
telluride ion can exist in molten salt at a presently
undefined redox potential. Such an ion could be an
important factor in tellurium behavior. However, it is
also plausible that tellurium is largely associated with
undissolved solids, by chemisorption or reaction. Any
of these phenomena would result in lower passage as a
gas to off-gas.

The viewpoint that emerges with respect to noble-
metal behavior in salt is that what we see is due to the
appearance of highly dispersed but undissolved material
in the salt, a mobile separate phase, presumably solid,
which bears much higher noble-metal fission product
concentrations than the salt. Our samples taken from
the pump bowl can only provide direct evidence
concerning the salt within the spiral shield, but if the
dispersion is fine enough and turnover not too slow, it
should represent the salt of the pump bowl and
circulating loop adequately.

We have suggested that the noble metals have behaved
as a mobile separate phase which is concentrated in

29

noble metals and is found in varied amounts in the salt
as sampled. This is illustrated in Fig. 6.5, where the
activities of the respective nuclides (relative to inven-
tory salt) are plotted logarithmically from sample to
sample. Lines for each nuclide from sample to sample
have been drawn. A mobile phase such as. we postu-
lated, concentrated in the noble metals, added in
varying amounts to a salt depleted in noble metals,
should result in lines between samples sloping all in the
same direction. Random behavior would not result.
Thus the noble-metal fission products do exhibit a
common behavior in salt, which can be associated with
a common mobile phase.

The nature and amount of the mobile phase are not
established with certainty, but several possibilities exist,
including (1) graphite particles, (2) tars from decom-
posed lubricating oil from the pump shaft, (3) insoluble
colloidal structural metal in the salt, (4) agglomerates of
fission products on pump bowl surface and/or bubbles,
(5) spalled fragments of fission product deposits on
graphite or metal. As we shall later see, at least some of
the material deposits on surfaces, and it is also indicated
that some is associated with the gas-liquid interface in
the pump bowl.

References

1. R. B. Gallaher, Operation of the Sampler-Enricher
in the Molten Salt Reactor Experiment, ORNL-TM-
3524 (October 1971).

2. R. C. Robertson, MSRE Design and Operation
Report, Part 1. Description of Reactor Design,
ORNL-TM-728 (January 1965).

3. R. E. Thoma, Chemical Aspects of MSRE Opera-
tion, ORNL-4658 (December 1971).

4. R. B. Briggs and J. R. Tallackson, “Distribution of
Noble-Metal Fission Products and Their Decay Heat,”
MSR Program Semiannu. Progr. Rep. Feb. 28, 1969,
ORNL-4396, pp. 62--64; also R. B. Briggs, “Estimate of
the Afterheat by Decay of Noble Metals in MSBR and
Comparison with Data from the MSRE,” internal
ORNL memorandum, November 1968. (Internal docu-
ment no dissemination authorized.)

5. See R. P. Wichner, with C. F. Baes, “Side Stream
Processing for Continuous lodine and Xenon Removal
from the MSBR Fuel,” internal memorandum ORNL-
CF-72-6-12 (June 30, 1972). (Internal document — no
further dissemination authorized.)
ACTIVITY RELATIVE TO INVENTORY SALT

ORNL-DWG 74-6623A

® 99M0 ¢ 132Te A |O3Ru
e} HiAg o 129Te A 106g .

ks r 3
Ji\ ® N\ V4

° ‘
a -/
ST
L d
'fi
I‘c- ’=

\‘ ¢
A\ / N
0.01 ; < —¢ ¢ \\/
0\ ’
¢ ;
$ 4
, '
h F 3 4
A/ \“l
0.001 \if
'y
’
0-000] 76 20-1 19
16-4 17-2 7 10" 22 29 31 32 18-2 4 6 12 19 44 45 46 19-9 24 36 42 44 47 55 57 58 59

SAMPLE NUMBER

Fig. 6.5. Noble-metal activities of salt samples.

0¢
31

Table 6.7. Data for -salt samples from pump bow! during uranium-235 operation

Each entry in the table consists of two numbers. The first number is the radiocactivity of the isotope in the sample expressed in disintegrations per minute per
gram of salt. The second number is the ratio of the isotope to the amount calculated for 1 g of inventory salt at time of sampling.

Sample D MWh Isotopes with noble-gas precursors Salt-seeking isotopes Noble metals Tellurium and iodine
ate X r .
No. Sr-89 Sr-91 Sr92  Ba-140 Cs-137  Zr95  Ce-141 Ce-143 Ce-144 Nd-147 Nb-95 Mo-99 Ru-103 Ru-105 Ru-106 Ag-111 Te-132 Te-129m  [1-131 I-133  I-135
Ladle salt samples
617 5-2366 2872 1.8EI10 1.1E11 98EI1C 2.6E10 4.7E10 1.8E8 4.9EIL0 3.1E10 24E10 1.1E11 9.5E10
0.92 0.81 0.76 0.84 0.51 0.012 23 0.24 0.65 0.82 0.50
6-19  5-25-66 2.2E10 1.2E11 1.2Ell 1.4E11 35E1l 7.1E9 2.5El 4.2E10 42E10 1.3E11 1.5Ell
090 0.63 0.69 0.85 25 0.38 8.4 0.21 0.83 0.62 0.75
77 6-27-66 3.0E10 1.2E11 9.7El0 6.1E10 1.5Ell 9.5E10 24E7 3.5E10 5.2E10 5.0E10 14E11 1.2E11
0.67 0.63 0.58 " 0.92 0.79 0.52 0082 1.2 0.19 0.73 0.62 0.60
7-10 7666 29E10 1.3E1l1 1.5EN 6.7E10 1.4E1l 1.1IE11  6.0E9 9.7E10 3.6E10 4.5E10 1.4E11
0.67 0.71 0.90 0.95 0.82 0.72 0.19 34 0.17 0.71 0.82
7-12 7-1366 40E10 13El11 7.5Ell 3.1EB  6.6E10 6.9E10 1.9E10 24E11? 3.2E10 7.1E9 3.8E10 2.1E8 3.8E10 4.9E8 54E10 14El11 1.1Ell
0.73 0.72 0.89 0.68 1.12 0.78 1.2 14.7 0.17 0.18 1.3 0.26 0.14 0.13 0.68 0.66 0.58
8-5 10-8-66 7,800 3.7E10 4.08E8 6.0E10 7.2E10 1.8E10 4.8E10 1.4E9 5.0E7 1.4E8 7.9E10
0.64 . 0.80 0.95 0.77 1.07 24 0.034 0.056 0.034 099
10-12 122866 13,800 3.8E10 1.3Ell 14E11 S4E10 4.6E10 24E10 6.7E9 3.6E10 8.0E8 ~4.0E7 16E10 1.5E8 5.SEI10
0.80 0.71 1.3 0.95 0.66 1.04 040 0.020 0.024 ~0.031 0.059 0.048 0.82
10-20 1967 15,800 4.7E10 1.3Eill 9.0E10 5.2E10 4.1El10 9.7E10 24E10 4.8E10 6.1E8 2.8E7 2.0E10 ~3.1E8 7.2E10
0.74 0.72 0.59 0.71 041 3.5? 0.86 0.25 0.014 0018 0.073 0.072 (.87
118 2-1367 19,000 4.8EI0 1.3Ell 1.0E11 9.3E10 9.2E10 1.1E9 3.2E11  5.2E9 1.6E8 5.5E10 S.OE10-
0.66 0.71 0.69 1.09 0.85 0.030 1.7 0.11 0.09 0.20 063
11-12 22167 20400 65E10 1.5Ell 1.3E11 9.3E10  1.1E1l 1.2E10  2.5E11? 5.3E9 ~1.9E8 3.8E10 7.6E10
0.80 0.88 0.79 0.98 0.90 0.27 1.3 0.09 ~0.09 0.13 0.85
11-22 3967 7.9E10 2.7E11? 1.1E11 9.1E10 3.E9 1.5E8 2.8E10 8.3E10
0.91 1.8 1.09 0.93 0.07 0.18 1.2
1145% 4-1767 29,000 8.6E10 1.7E11 3.0E10? 96E10 1.5E11 14El10 4.E8 5.1E7 3.5E10 9.2E10
0.77 089 0.23 0.29 0,83 0.19 0.14 009 0.3 0.98
11-51  4-2867 30,800 B8.0E10 1.3EIl 1.8E11 1.2E11  1.5E11 1.1E11 64E10 4.0E9 7.2E10  3.8E9 ~1.E8 6.5E7 1.7E10  5.2E8 8.2E10
0.69 1.1 0.96 0.86 0.87 0.64 1.2 0.04 0.40 0.05 0.03 0.11 0.06 0.07 0.88
11-52  5-1-67 31,250 8.1E10 2.2E11 1.3E11  1.6E11 7.7E10 6.1E10 1.9E9 8.2E10 8.9E1(? 24E8 49E7 16E10 5.0E8 8.2E10
0.69 1.1 0.96 0.96 0.42 1.09 002 0.44 1.1 0.08 009  0.06 0.07 0.87
11-54  5-5-67 32,000 1.3E11 3.1E10 3.5E10 1.8E9 7.5E7 1.1E10 7.0E10
0.94 0.17 0.19 0.02 0.02 0.04 0.74
11-58° 5-8-67 32,650 1.0E1l 1.7E11 1.5E11 1.7E11 18Ell1 6.2E10 2.2E10 3.2E10 9.5E9 36E8 16E7 24E9 S.SE8 1.2E10
' 0.88 1.01 L.10 1.03 26 1.10 0.22 0.03 0.12 0.11 003 005 0.07 0.14
1269 6-2067 32,650 B89EI0 1.5E1l1  14Ell 6.0E10 8.0E10 6.9E9 3.9E8 2.1E9
0.76 1.04 0.83 1.06 0.80 0.09 0.12 0.28
12-27° 7-1767 36,650 6.1E10 1.0E11 1.9E10 1.2E11  5.6E9 2.4E8 1.3E10  3.1E8 17.1E10
0.75 0.97 0.34 0.68 0.11 0.08 0.05 0.07 0.89
1422 11-767 9.2E11 1.3E11 1.7E11 <1E8 8.2E10  3.7E9 ~2.7E8 8.9E9 1.9E8 6.6E1l
9.6 1.02 0.84 <0.001 042 0.06 0.07 0.030 0.045 74
Freeze valve salt samples
14-20FV 11467 8.2E10 1.6E11 1.2E11 14E11 4.2E7 2.2E9 1.4E8 24E7 8.2E8 <I1.1E8 S.5E10
0.89 0.99 1.04 0.76 0.0006 0013 0.002 0.006 0.003  <0.02 0.67
14-30FV 12-5-67 8.5E10 1.2E11 1.3E11 1.5E11 7.0ES 8.5E8 4.9E7 <6.4E6 1.3E9 <29E7 34El10
0.87 0.77 1.06 0.77 0.00001 0.005 0.001 <0.002 0.005 <0.005 045
14-63FV 2-27-68 9.2E10 14E1l 9.3E10 1.5E11 1.9E11 34E6 3.2E8 2.5E7 <3.2E6 2.8E8 4.5E10
0.87 1.14 0.59 1.12 1.2 0.0003  0.002 0.0004 <0.0007 0.001 0.55
1466FV [ 9.1E10 ~9.1E10 1.7E11 1.2E11 9.2E10 ~2.5E9 32E8 <2.2E7 <1.6E7 ~6.2E8
0.90 457 1.26 0.94 24 ~0.02 0.004 <0.0003 <0.003 ~0.005

9Before power; corresponds to end of run 7.
b After addition of 8.6 g of beryllium; no purge.
Pump off 2 hr.
9Before power; corresponds to end of run 11.
€After addition of 38 g of beryllium.

End of run 14,
32

Table 6.8. Data for salt samples from MSRE pump bowl during uranium-233 operation

Sample number
Sample weight, g

Date

Megawatt-hours

Power, MW
Rpm

Pump bowl level, %
Overflow rate, Ib/hr

Voids, %

Flow rate of gas, std liters/min

Sample line purge

Halflife [ Sion

(days) yield
Isotope (%)
Sr-89 5200 546
Sr-90 10,264.00 5.86
Y91 58.80 5.57
Ba-140 1280 540
Cs-137 10,958.00 6.58
Ce-144 284.00 4.61
Zr95 65.00 6.05
Nb-95 3500 6.05
Mo-99 279 4.80
Ru-10% 39.60 1.99
Ru-106 367.00 043
Sb-125 986.00 0.08
Te-132 325 440
1131 8.05 290
Constituent
U-233
U, Total
Li
Be
Zr

3.30 He
On

1.32E9
0.198

5.53E9
1.355
143E10
0.295
4.04E9
0.279
6.39E9
0.740

5.13E7
0.025
1.97E8
0.034
1.62E7
0.143

1.222
0.151
44.35
0.384
25.28
0.378
41.3

0.357

15-328
60.9115
10-15-68

5.98E9
0936
2.48E9
0.609

3.42E9
0.840
5.18E10
1.077
1.28E10
0914
1.04E10
1.159

1.15E5
0.000
8.6ES
0.000
7.48E3
0.000

7.86
0.974

63.3

0.948
109.8
0.949

15428 15-518
51.5684 56.9563
10-29-68 11668

0.1 0.1
0.00 0.00
1180 1180
66.50 66.00
3.7 1.0
0.00 0.00
3.30 He 3.30 He
On On

Fission Product isotopes”®

4 47E9 3.76E9
0.839 0.787

2.86E9
0.703

3.35E9 3.30E9

0.823 0.811
5.41E10 5.17E10
1.163 1.134
1.07E10 1.26E10
0.884 1.135
7.12E9 8.71E9
0.718 0.854

1.45E6 4.77ES

0.001 0.000
4.65E6 1.42E6
0.001 0.000
4.38ES 1.18E4
0.004 0.000

Salt constituents?

7.50
0.929
109.8
0.951
57.9

0.867
104.9
0.907

15-578
32.0493
11-11-68
0.1

0.00
1180
62.50
0.8
0.60
3.30 He
On

3.84E9
0.871

3.77E9
0.926
5.61E10
1.249
1.02E10
0.981
1.08E10
1.059

5.12E6
0.004
1.78E7
0.003
2.16ES
0.002

15698
51.9611
11-25-68

3.42E9
0.842
4.61E10
1.060
8.46E9
0.935
2.02E8
0.020

4.64E3
0.000

7.82
0.969

68.3

1.022
105.3
0.910

164-FVS
56.0914
12-16-68

2.98E9
1.014
3.63E9
0.894
6.44E9
1.236
2.51E8
1.167
2.81E9
0.692
4.55E10
1.099
1.02E10
1.382
5.08E9
0.537
3.17E6
0.096
8.75E6
0.012
2.25E7
0.004
7.42E5
0.007
3.90E6
0.092
1.14E8
1.129

6.53
0.977
785
0.973
108.8
0.942
65.1
0975
87.5
0.756
33

Table 6.8 (continued)

Sample number
Sample weight, g
Date
Megawatt-hours
Power, MW

Rpm

Pump bowl level, %
Overflow rate, Ib/hr
Voids, %

Flow rate of gas, std liters/min
Sample line purge

Haite F
Isotope (%)
Sr-89 5200 5.46
Y91 58.00 5.57
Ba-140 12.80 540
Cs-137 10,95800 6.58
Ce-141 3300 7.09
Ce-144 28400 4.61
Nd-147 11.10 198
Zr95 65.00 6.05
Nb-95 35.00 6.05
Mo-99 279 480
Ru-103 39.60 1.99
Ru-1 66 367.00 043
Ag-111 7.50 0.02
Sb-125 986.00 0.08
Te129m 3400 033
Te-132 3:25 440
1-131 805 290
Constituent
U-233
U-total
Li
Be
Zr

18-2NFVS 184-NFVS 186-NFVS 1812-NFV 18-19-NFV 1844-NFV 1845-NFV 1846-NFV

3.8651
4-14-69

0.60
3.30 He
On

7.36E10°

0.771
144E11
1.702
1.21E11
0.917
4.06E9
0.835
1.29E11
0.908
6.28E10
1.222
7.29E10
1488
8.50E10
0.966
2.40E10
0.462
1.75E9
0.015
193E10
0.521
1.28E9
0230
2.02E7
0.034

7.24E8
0.007
7.33E9
0.101

7.28

1.089
23.54
2917
116.4
1.008
6597
0.988
138.4
1.196

11.1866
4-18-69
12,810.0
8.00
1180
63.10
4.9

0.60
3.30 He
On

7.60E10
0.780

9.02E10
1.043

1.43E11
1.100

4.30E9
0.878
1.12E11
0.783

6.30E10
1.214

5.52E10
1.150

8.96E10
0.994

1.23E10
0.224

9.99E8

" 0.008

4.30E8
0.004
249E10
0.352

7.11

1.064
6.18

0.766
107.5
0931
65.1

0.975
135.1
1.168

6.7612 12.6825
42369 5269
13,7050 15,5120
8.00 8.00
1180 1180
63.20 59.50
47 3.0
0.60 0.60
3.30 He 3.30 He
On On

Fission product isotopes?

624E10  1.12E1l
0.600 0.966
1.38E11
1.340
1.17E11
0.813
4.03E9  3.66E9
0.811 0.718
1.09E11  1.38El1
0.712 0.807
SSO0E10  6.82E10
1.034 1227
7.60E10  101E1l
0.792 0.944
9.09E10  3.17E9
1.562 0.048
131E11  1.38E9
0.851 0.008
297E9  7.13E7
0.074 0.002
1.68E8
0.030
8.35E8
1.232
‘ 1.68E7
0.068
1.29E10  3.02E7
1.828 0.004
249E11  3.56ES
1.804 0.002
211E10  5.56E10
2.260 0.595

Salt constituents?

6.233
0.933
6.936
0.859
64.34
0.557
60.93
0912
914

0.790

6.72
1.005

98.56
0.853

12.6691
59-69
16,848.0
8.00
1180
58.90
09

0.60 -
3.30 He
On

7.51E10
0.621
1.49E11
1.380
1.84E11
1.143
4.74E9
0.915
141E11
0.806
6.25E10
1.096
3.90E10
0.653
1.13E11
1.009

1.17E9

0.01e
1.16E9
0.008

9.30E7
0.002

4 .90E7
0.066

6.59E10
0.739

7.19.
1.076
6.87
0.851
110.7
0.958
70.0
1.048
103.9
0.898

3.6339
5-29-69
19,784.0
6.90

990
52.90
0.0

0.00
2.30 He
On

1.00E11
0.758

1.32E11°

1.109
1.56E11
1.006
4.25E9
0.786
145E11
0.784
5.67E10
0.937
2.36E9
0.041

1.18E11

0.952
-3.10E9
-0.036
4.22E9
0.035
7.58E7
0.002

4.03E7
 0.060

9.27E8
0.008
2.78E10
0.339

6.88

1.029
6.96

0.862
103.2
0.894
62.5

0.936
128.3
1.109

14.0790
6-1-69
20,309.0
0.00

990
50.00
0.0

0.00
2.00 Ar
On

1.01E11
0.754
1.35E11
1.116
1.65E11
1.058

4 98E9
0914
1.52E11
0.813
6.21E10
1.016
6.82E10
1.196
1.20E11
0.952
3.33E10
0.374
1.80E11
1.364

8.76E7
0.128

4 40E8
0.051
1.74E10
0.144
3.62E10
0.438

7.10

1.062
8.88

1.100
106.7
0924
64.0

0.958
110.8
0.958

14.9780
6-1-69
20,309.0
0.00

990
56.30
0.0

0.00
2.00 Ar
On

9.57E10
0.720
1.23E11
1.017
1.60E11
1.039
4.58E9
0.840
1.39E11
0.747
7.54E10
1.234
5.80E10
1.030
1.13E11
0.904
—2.00E9
—0.022
1.45E10
0.116
4.29E8
0.009
2.03E7
0.003
2.66E7
0.040

2.17E8
0.025
4.99E9
0.043
6.68E9
0.082

6.58
0.984
6.75
0.836
98.6
0.854
70.6
1.057
111.7
0.965
34

Table 6.8 (continued)

Sample number
Sample weight, g
Date
Megawatt-hours
Power, MW

Rpm

Pump bowl level, %
Overflow rate, Ib/hr
Voids, %

Flow rate of gas, std liters/min

Sample line purge

Half-life ©ossion

(days) yield
Isotope (%)
Sr-89 5200 546
Y91 58.80 5.57
Ba-140 12.80 540
Cs-137  10,958.00 6.58
Ce-141 33.00 7.09
Ce-144 284.00 4.61
Nd-147 11.10 1.98
Zr-95 65.00 6.05
Nb-95 35.00 6.05
Mo-99 2.79 480
Ru-103 "39.60 1.99
Ru-106 367.00 043
Ag-111 7.50 0.02
Te-129m 3400 033
Te-132 325 440
[-131 8.05 290
Constituent
U-233
U-total
Li
Be
Zr

19-55-FVS
4.2936
10-14-69

0.53
3.30 He
Off

6.82E10
0.770
8.08E10
0.996
1.54E11
1.116

4.01E9
0.689

1.11E11
0.902

6.70E10
1.216

7.16E10
1.341

8.31E10
0.954

—-6.40E9
—0.092

1.09E11

© 0.626

5.84E9
0.180
3.24E8.
0.058
2.96E7
0.041
7.24E8
0.128
6.60E9
0.042
3.84E10
0.444

7.60
1.137
6.84
0.848
1232.0
10.667
68.2
1.021
133.5
1.154

19-57-FVS 19-58-FVS 19-59-FVS
3.3515 12.3829 8.4980
10-169  10-1769  10-17-69
27,1630 27,1770 27,177.0

8.00 0.01 0.01
1186 1188 1189
63.00 67.40 65.20
3.7 9.0 39
0.53 0.53 0.53
3.30 He 3.30 He 3.30 He
Off Off Off

Fission product isotopes?

691E10  6.7SE10  $.96E10

0.747 0.730 0.645
9.54E10 9.87E10 1.89E11
1.128 1.165 2.229
1.58E11 1.49E11 1.57E11
1.097 1.035 1.098
4.97E9 5.06ES 4.67E9
0.848 0.863 0.797
1.12E11 1.05E11 1.02E11
0.868 0.814 0.791
6.50E10 6.55E10- 5.93El10
1.165 1.174 1.063
6.78E10 7.52E10 6.36E10
1.224 1.360 1.161
8.42E10 8.33E10 8.11E10
0.930 0.920 0.897
3.55E9 -2.37E9  -9.50E8
0.050 -0.034 -0.013
1.26E11 3.11E9 2.98E10
0.754 0.019 0.186
9.48E9 1.25E8 1.11E9
0.279 0.004 0.033
6.05E8 1.17E7 8.37E7
0.107 0.002 0.015
1.76E9 1.38EB
0.295 0.023
1.22E10 2.28E8 1.00E9
0.080 0.002 0.007
4.78E10 5.65E10 1.29E10
0.539 0.640 0.148

Salt Constituents?

7.14 6.84 6.68

1.068 1.023 0.999
741 5.78 5.94

0.918 0.716 0.736
138.9 99.3 107.1
1.203 0.860 0.927
67.5 64.7 65.8

1.010 0.969 0.985
160.5 127.7 131.0
1.387 1.104 1.132

19-76 FVS  20-1-FVS

134771

8.04E10
0.731
7.57E10
0.760
1.71E11
1.043
5.94E9
0.983
1.32E11
0.846
6.82E10
1.156
7.28E10
1.170
7.89E10
0.744
2.30E9
0.030
2.33E9
0.014
7.98E7
0.002

4.67E7
0.000
3.85E10
0.402

7.28
1.089
6.38
0.791
108.4
0.939
624
0.934
75.8
0.655

3.1065
11-26-69
30,315.0
8.00
1190
64.00
6.8

0.53
3.30 He
Off

4.80E10
0.625
9.42E10
1.292

4 .82E10
1.071

4 51E9
0.749
6.12E10
0.673
6.19E10
1.128

7.26E10
0919
149E10
0.186
4.71E10
3.340
7.48E9
0.289
7.81E8
0.140
5.07E7
0.478
9.44E8
0.222
6.40E9
0.547
9.49E9
0.678

7.13

1.067
7.44

0.922
100.6
0.871
68.7

1.028
110.7
0.957

20-19-FVS§
54784
12-569
31,9270
8.00
1200
63.50
5.6

0.53
3.30 He
Off

4.79E10
0.532
8.48E10
1.017
8.80E10
0912
4.86E9
0.793
8.04E10
0.693
6.03E10
1.058

7.64E10
0.848

4 36E9
0.054
1.14E11
0.781
3.50E9
0.110
2.90E8
0.051
1.16E8
0.234
1.49E9
0.278
2:64E10
0.206
2.37E10
0.407

6.67
0.998
7.32
0.907
87.7
0.759
65.3
0.978
170.5
1.474
35

Table 6.8 (continued)

Sample number
Sample weight, g
Date
Megawatt-hours
Power, MW

Rpm

Pump bowl level, %
Overflow rate, lb/hr
Voids, %

Flow rate of gas. std liters/min

Sample line purge

Halrdife 50"
tsotope VY ")
Sr-89 52.00 546
Y91 5880 5.57
Ba-140 12.80 540
Cs-137  10,958.00 6.58
Ce-141 33.00 7.09
Ce-144 284.00 4.61
Nd-147 11.10 198
Zr-95 65.00 6.05
Nb-95 35.00 6.05
Mo-99 279 480
Ru-103 39.60 199
Ru-106 367.00 043
Ag-111 7.50 0.02
Te-129m 3400 033
Te-132 3.25 440
1-131 3.05 290
Constituent
U-233
U-total
Li
Be
Zr

19-1-FVS
11.2623
8-11-69
20,309.0

4.22E8
0.009
1.94E8
0.004
1.12E7
0.004
1.35E8
0.025
6.92E7
0.002
4.94E8
0.010

1.60E8
0.003
8.79E9
0.111

9.53E8
0.071
1.23E8
0.023

1.97E8
0.104

4.70E7
0.281

0.0520
0.008
0.0760
0.009
118.1
1.023
90.7
1.358
5.79
0.050

19-6-FVS
0.2130
8-1569
20,309.0
0.00
1170
62.00
0.7

0.00
3.30 He
On

7.80E8
0.017
7.69E8
0.016
5.35E7
0.022
5.36E8
0.100
5.49E9
0.152
1.07E9
0.021
7.02E6
0.014
5.44E8
0.010
~8.33E7
~0.001

3.88E7
0.003
1.02E7
0.002

2.08E7
0.012

0.1454
0.022
0.1454
0.018
246
0.021

110.80
0.958

19-9-FVS
10.2040
8-18-69
20,309.0
0.01
1189
65.00
2.4

0.60
3.30 He
On

19-24-FVS
1.7933
3-10-69
22,255.0
.01
1165
62.10
4.7

0.70
2.90 Ar
Off

Fission product isotopes”

3.13E10
0.700
1.81E10
0.393
1.79E9
0.848
2.77E9
0.516
2.04E10
0.600
4.51E10
0913
7.50E7
0.184
3.75E10
0.723
2.60E10
0.343

8.48E6
0.001

5.05E10
0.886
7.58E10
1.366
1.22E10
0.190
4.67E9
0.848
4.78E10
0.748
5.05E10
1.000
3.00E9
0.119
5.24E10
0.859
2.25E10
0.330
1.85E10
0.220
3.83E9
0.209
1.24E9
0.233

2.21E9
0.742
1.83E10
0.236
2.55E10
0.596

Salt constituents?

6.54
0978
4.13
0.512
71.4
0.670
521
0.780
64.40
0.557

7.06
1.056
5.26
0.652
954
0.826
71.6
1.072
131.1
1.133

19-36-FVS
14.0175
9-29-69
23,838.0
5.50
608
58.20
0.9

0.00
3.35 He
Off

5.24E10
0.820
6.04E10
0.998
7.10E10
0.938
6.19E9
1.101
5.95E10
0.764
549E10
1.068
2.71E10
0.944
5.70E10
0.858
--5.76E9
-0.086
1.51E9
0.019
4.72E7
0.002

4.23E7
0.115
1.21E7
0.003
4.36E8
0.006
2.54E10
0.576

6.28
0.940
6.42
0.796
86.6
0.750
61.2
0.916
115.7
1.000.

1942-FVS
2.1662
10-3-69
24,391.0
7.00
1176
68.00
6.3

0.53
3.30 He
Off

5.87E10
0.859
7.19E10
1.123
8.58E10
0.975
7.67E9
1.355
7.26E10
0.841
5.84E10
1.123
4.08E10
1.214
6.53E10
0.932
—4.00E9
~0.060
4.78E10
0.431
3.14E9
0.133
4.80E8
0.089
9.05E7
0.202
7.24E8
0.181
9.59E9
0.098
3.47E10
0.652

7.28
1.089
3.73
0.462
933
0.808
70.5
1.055
149.9
1.296

19-44-FVS§
49789
10-6-69
24,9490
8.00
1188
64.00
7.4

0.53
3.30 He
Off

5.73E10
0.780
8.24E10
1.208
1.02E11
0.990
4.64E8
0.081
7.85E10
0.819
5.78E10
1.095
4.67E10
1.176
7.52E10
1.011
—3.54E9
—0.053
5.85E10
0412
1.16E9
0.045
7.15E7
0.013
3.23E8
0.596

2.26E9
0.018
5.72E10
0.891

6.94
1.038
4.93
0.611
97.3
0.842
68.4
1.024
157.6
1.362

1947-FVS
2.1144
10-7-69
25,190.0
8.00
1175
61.40
1.8

0.53
3.35 He
Ooff

6.74E10
0.889
8.72E10
1.244
1.07E11
0.982
4.55E9
0.795
8.60E10
0.860
6.09E10
1.147
5.06E10
1.202
7.55E10
0.990
1.77E9
0.026
1.26E11
0.829
4.10E9
0.153
2.92E8
0.053
5.88E7
0.102
8.99E8
0.194
8.52E9
0.064
7.36E9
0.108

7.24

1.083
5.24

0.649
111.3
0.964
61.3

0918
129.2
1.117
36

Table 6.8 (continued)
Sample number 17-2-FV§  17-7-FVS  17-10-FV§S 17-22-FVS 17-29-FVS 17-31-FVS 17-32-NFV
Sample weight, g 53.3293 6.0976 13.2173 38.3200 46.2666 57.3671 5.6200
Date 1-14-69 12369 1-28-69 2-28-69 3-2669 4-169 4-369°
Megawatt-hours 13.5 752.0 1244.0 5757.0 9158.0 10,168.0 10,456.0
Power, MW 0.45 4.60 4.70 7.00 6.90 7.20 7.20
Rpm 1180 1180 1180 942 1050 1050 1050
Pump bowl level, % 59.20 57.20 56.80 64.50 57.80 61.50 60.40
Overflow rate, lb/hr 38 1.8 1.6 0.7 1.3 30 4.5
Voids, % 0.60 0.60 0.60 0.00 0.05 0.05 0.05
Flow rate of gas, std liters/min 3.30 He 3.30 He 3.30 He 3.30 He 3.30 He 3.30 He 3.30 He
Sample line purge On On On ~ On On On On
' Fission product isotopes?
Halflife ©Ssion
(days) yield
isotope (%)
Sr-89 5200 546 1.38E9 5.58E9 1.06E10 3.52E10 6.56E10 5.86E10 7.21E10
0.687 0477 0.596 0.554 6.765 0.630 0.758
S$r90 10,264.00 586 2.87E9 4.59E8 3.12E9 1.20E10
0.709 0.112 0.755 2.526
Y91 5880 557 243E9 1.06E10 1.97E10 7.26E10 9.35E10 5.22E10 791E10
_ 0.657 0.964 1.224 1.310 1.221 0.636 0.942
Ba-140 1280 540 1.37E8 2.64E10  3.33E10  4.76E10  1.38El11  146E1l  1.44E1l
1.421 0.805 0.688 0.384 1.078 1.050 1.014
Cs-137 10,958.00 6.58 3.00E9 2.79E9 3.45E9 4.37E8 4.30E9 3.89E9 3.69E9
0.741 0.680 0.833 0.098 0.911 0.812 0.766
Ce-141 33.00 7.09 1.02E10 2.83E10 B.34E10 1.12E11 1.09E11 1.16E11
0.531 0.943 0.802 0.855 0.768 0.800
Ce-144 284.00 4.61 4.23E10 27SE10  246E10° 4.89E10  6.33E10  S.64E10  S97E10
1.096 0.702 0.618 1.072 1.276 1.108 1.164
Nd-147 11,10 1.98 8.87E9 1.93E10 4 65E10 5.75E10 5.60E10 6.11E10
0.657 "0.985 0.985 1.216 1.083 1.157
Zr95 65.00 6.05 4.28E9 947E9 1.62E10 © S.10E10 7.71E10 7.84E10 8.37E10
0.790 0.717 0.890 ~ 0.887 0.983 1 0.921 0.962
Nb-95 3500 605 4.18E9 2.34E9 -2.00E8 -1.19E9 2.15E10 1.49E10 1.48E10
0.519 0.289 -0.023 ~0.050 0.513 0.322 0.311
Mo-99 279 480 5.11E8 9.09E9 4.60E10 1.34E9 4.63E8 1.89E9 4 92E8
2.481 0.117 0.525 0.010 0.004 0.014 0.003
Ru-103 39.60 199 2.13E7 1.59E8 8.34E7 1.98E7 7.44E8 6.46E7 8.28E8
0.050 0.034 0.011 0.001 0.022 0.002 0.022
Ru-106 367.00 043 146E7 7.99E6 2.69E6 1.40E6
0.003 0.002 0.001 0.000 .
Ag-111 ’ 7.50 002 1.15E6 2.01E7 2.63E6 8.59E7 5.35E7
2.573 0.091 0.004 0.133 0.081
Te-129m 3400 0.33 3.69E6 1.76E8 2.66E7 1.10E7 6.09E7 1.25E8
0.257 0.208 0.020 0.002 0.010 0.019
Te-132 325 440 S599EB 2.06E10 2.35E9 2.55E9 1.15E9 4.78E9 4.52E8
3675 0.307 0.030 0.020 0.012 0.038 0.003
1-131 8.05 290 B8.78E7 1.02E10 1.30E10 3.37E10 1.98E10 2.37E10 3.22E10
1.672 0.403 0.369 0.456 0.284 0.305 0.404
Salt constituents®
Constituent .
U-233 6.39 4.72 6.116 598 7.01 7.88
0.956 0.706 0915 0.895 1.049 1.179
U-total 741 5.570 3.19 4 .88 5916 7.892
0918 0.690 0.395 0.605 0.733 0.978
Li 100.9 9204 117.68 91.0 110.2 107.0 112.1
0.874 0.783 1.019 0.788 0.954 0.926 0971
Be 629 61.78 575 6147 71.23 63.7
0.942 0.925 0.861 0.920 1.066 0.954
Zr 98.90 162.14 92.7 110.9 75.25 134.7
0.855 0.883 0.801 0.959 0.650 1.164

“Each entry for the fission product isotopes consists of two numbers. The first number is the radioactivity of the isotope in the
sample expressed in disintegrations per minute per gram of salt. The second number is the ratio of the isotope to the amount
calculated for 1 g of inventory salt at time of sampling.

bEach entry for the salt constituents consists of two numbers. The first number is the amount of the constituent in the sample
expressed in milligrams per gram of salt. The second number is the ratio to the amount calculated for 1 g of inventory salt at time of

sampling.
37

7. SURFACE DEPOSITION OF FISSION PRODUCTS BY PUMP BOWL EXPOSURE

7.1. Cable

It was evident from the earliest samples taken from
the MSRE pump bowl that certain fission product
elements, notably the noble metals, were concentrated
on surfaces exposed to the gas within the mist shield,
and possibly also the liquid below. The first such tests
looked at.segments of the latch cable for capsule
samples, or of coils of wire of various metals (Hastelloy
N, stainless steel, and silver). Data from such tests
extending over the period of 23% U operation are shown
in Table 7.3 at the end of this chapter. Both liquid and
gas exposures were obtained. In order to facilitate
comparisons, it is desirable where possible to express
the deposition in terms of activity per unit area.
However, the determined values were almost always in
terms of the total sample. Thereby appropriate areas for
any individual sample were used as given in the
tabulations. These were obtained by calculation from
measurement when possible or by estimate (marked ~)
where necessary.

It was soon evident that the values for salt-seeking
isotopes (least), daughters of noble-gas nuclides, and the
so-called noble metals (greatest) fell into distinctly
different categories. The basis of comparison for these
numbers depends on the average time between produc-
tion of the nuclide element by fission and deposition on
the surface. The activities of elements having the same
behavior should be proportional for brief holdup times
to fission rate yield and decay constant and for long
holdup times to yield and to the power-averaged
saturation factor, becoming independent of decay
constant and approaching proportionality to inventory.
The ratios of activities of two isotopes of certain
elements (cerium, ruthenium, tellurium) appear to
reflect appreciable holdup, and so the values given in
the tables in this section are compared with inventory
salt values, the units “‘equivalent milligrams of inven-
tory salt per square centimeter” being used to indicate a
convenient order of magnitude.

7.2 Capsule Surfaces

During the 233U operation a variety of capsules were
dipped into the pump bowl for purposes other than
removing salt samples. The capsules and attached
materials were dissolved, and radiochemical analyses
were obtained. As these constituted a type of dipped
sample, the values obtained and ratios to inventory are

shown in Table 7.4 at the end of this chapter. Again,
salt-seeking nuclides form a lowest group, nuclides with
noble-gas precursors a somewhat higher group, and the
noble-metal group the highest, by about two or more
orders of magnitude. Significantly greater amounts of
noble metals were found when the capsule involved a
reductant (beryllium, zirconium, chromium) than when
the added material was oxidizing (FeF,;) or not
reducing (nickel, copper).

After the introduction of the double-walled sample
capsule (about the end of run 17), the exterior capsule
of both gas and salt samples became available for
dissolution and radiochemical analysis. Data from these
sources are presented in Tables 7.5 and 7.6 (end of
chapter) as disintegrations per minute per square
centimeter and equivalent milligrams of inventory salt
per square centimeter. Again, the values generally are
lowest for salt-seeking nuclides, higher for nuclides with
noble-gas precufsors, and orders of magnitude higher
for the noble metals.

It is of particular interest to note the values found
when the system was at low or zero power. Since
mostly (except for nuclides with noble-gas precursors)
the values did not go down much, it would appear that
the holdup period (in particular for noble metals)
appreciably exceeded the period of low or zero power
preceding the sample.

7.3 Exposure Experiments

Five experiments were conducted in which both
graphite and metal speciments were exposed for varied
periods of time below the surface of the salt. Data from
the first of these experiments are shown in Table 7.1.
These data are of interest because they permit compari-
son of deposition rates on metal and graphite, and
between liquid and gas phases. In addition, some
protection against contamination during passage, by
contact with substances deposited in the sample trans-
fer tube, was inherent in the design of the capsule cage.

The first experiment, 11-50, used three graphite
specimens and one Hastelloy N specimen held by end
caps to be out of contact with the transfer tube at all
times (Fig. 7.1.). Two of these assemblies were con-
tained in a perforated nickel container or capsule which
was lowered into the pump bowl. Contamination from
the areas above the pump bowl during removal, etc.,
though not likely, was not precluded, however.
Table 7.1. Data for graphite and metal specimens immersed in pump bowl
Test 11-50 exposed on April 26, 1967, for 8 hr at a reactor power of 8MW

Deposition (equivalent milligrams of inventory salt? per square centimeter of specimen surface)

Material Phase (:;\rrneza) U-235 Sr-89 Ba-140 Ce-143 Z1-95 Nb-95 Mo-99 Ru-103 Ru-106 T§~1 32 I-131
CGB#11 Liquid 1 0.09 _ 31 2.2 0.09 1.1 113 11 8 102 9
CGB#22 Liquid 1 0.03 0.5 0.04 0.01 1.1 162 18 14 90 7
Pyrolytic Liquid 1 0.03 49 1.1 0.06 2.0 643 117 10 206 11
Hastelloy N Liquid 1 0.04 0.69 0.13 0.3 ‘ 0.08 27 81 4 3 510 29
Wire Liquid ~1 1.2 9 950 45 33 5000 650
Wire Interface  ~1 7 12 7000 180 90 3600 4800
CGB#61 Gas 1 0.10 19 2.3 0.10 1.4 92 4.7 3.6 680 10
CGB#92 Gas 1 0.03 0.7 0.03 2.5 80 7.6 54 860 5
Pyrolytic Gas 1 0.02 2.4 2.1 0.03 2.0 118 10.7 7.2 100 5
Hastelloy N Gas 1 0.2 3.8 2.3 0.20 1.0 105 2200 67
Wire Gas ~1 2.4 ' 0.6 55 8400 180

Inventory for 1g of salt
Value for U-235 in micrograms per gram of salt; values for fission products in disintegrations per minute per gram of salt
14,250 1.16E11 1.91E11 1.87E11 1.38E11 9.6E10 1.88E11 7.8E10 3.0E9 1.40E11 9.5E10

2.79E11

8¢
39

SAMPLES,
3 in. DIAM X Y5 in —fm

ORNL - DWG 67 - 12239

MATERIAL:, Ni OR NICKEL PLATED

MILD STEEL

/35~ in. HOLES FOR CABLE
(SPACE BETWEEN SAMPLES)

SET SCREW

Ni WIRE RING THROUGH HOLE IN
CENTRAL ROD. MAY BE OMITTED
IF SET SCREW CONSIDERED SAFE
AGAINST LOOSENING

INOR-8

CGB GRAPHITE CGB GRAPHITE

PYROLYTIC GRAPHITE

Fig. 7.1. Specimen holder designed to prevent contamination by contact with transfer tube.

For the specimens, we note that generally the
differences between metal and graphite and between
liquid and gas exposure are not great. It does appear
that the !32Te activity was significantly higher on
metal than on graphite. We again find the general
patterns previously observed, with salt-seeking elements
lowest, noble-gas daughter nuclides higher, and appreci-
ably higher levels of noble metals.

Clearly, the wire to which the capsule was attached
received considerably more activity than the specimens.
This increased activity on the wire might be a result of
easier contamination or easier access to its surface while
in the pump bowl, than to the specimens which were
within a perforated container.

Data for four subsequent experiments, 18-26, 19-66,
19-67, and 19-68, which were liquid-phase exposures
for various periods conducted during the operation with
233U, are shown in Table 7.2. In order to avoid
contamination problems and problems of contact with
the gas in the pump bowl, the specimens were con-
tained in a windowed capsule, below a bulb which
would float in salt, and thereby open the windows, but
when above the salt would drop to close off the
windows. This device is shown in Fig. 7.2.

Data for the various exposures are shown as milli-
grams of inventory salt equivalent to the activity
deposited on one square centimeter. Data for isotope
pairs(193-106 Ry, 132:129MTe) apnear more consistent
40

Table 7.2. Deposition of fission products on graphite and metal specimens
in float-window capsule immersed for various periods in MSRE pump bowl liquid salt

All specimens exposed at reactor power of 8 MW

Nuclide Sr-89 Ba-140 Nd-147 Ce-141 Ce-144 Z1-95 Nb-95 Mo-99 Ru-103 Ru-106 Ag-111 Te-132 - Te-129 I-13
Fission yield, % 5.86 54 1.98 7.09 4.61 6.05 6.05 4.8 1.99 0.24 0.024 44 0.33 2.9
Half-life, days 52 12.8 11.1 33 284 65 35 2.79 39.6 367 7.5 3.25 34 8.05
inventory
Disintegrations per minute per milligram of salt
Sample
No. Date
19-68 10-27-69 1.06E8 1.60E8 6.09E7 1.50E8 5.72E7 1.02E8 17.50E7 1.65E8 394E7 3.26E6 7.88E5 1.51E8 1.51E7 9.44E7
18-26 5-19-69 1.28E8  1.62EB S599E7 1.B3E8 S590E7 1.20E8 7.94E7 1.43E8 4.85E7 3.39E6 7.34E5 1.32EB8 1.85E7 8.B4E7
19-67 10-24-69 1.02E8 1.56E8 595E7 144E8 5.76E7 9.89E7 7.36E7 1.64E8 3.79E7 3.23E6 7.77E5 1.50E8 145E7 9.29E7
19-66 10-24-69 1.02E8  1.56E8 593E7 144EB 5.75E7 9.85E7 7.34E7 1.64EB 3.77E7 3.23E6 7.76E5 1.50EB 144E7 5.2EB
Disintegrations per minute produced by MSRE in 1 hr per square centimeter of MSRE surface
1.62E8 6.1EB 2.6EB 3.1E8 2.3E7 1.3E8 1.2E8* 2.5E9 7.2E7 9.4ES5 4.6E6 1.9E9 3.0E7 5.2E8
Deposit activity
Expressed as ratio of activity per square centimeter to amount calculated for I mg of inventory salt at time of sampling
Sample Duration Material
No.
19-68 10 min  Graphite 0.12 0.005 0.001 0.005 0.013 1.1 8 1.7 0.9 0.9 09
0.12 0.005 0.002 0.013 0.011 1.5 8 21 1.0 1.0 1.0
0.07 0.003 0.011 0.013 0.011 1.0 4 0.8 0.4 0.5 0.8
Metal 0.11 0.019 0.015 0.024 0.024 1.9 7 1.1 1.3° 1.0 1.6
0.09 0.004 0.005 0.013 0.009 2.0 3.2 0.4 0.4 0.4 04
0.07 0.006 0.005 0.009 0.009 19 4.1 0.6 0.5 0.5 0.3
18-26 1 hr Graphite 0.06 0.004 0.004 <0.0001 0.006 0.9 4 0.7 0.2 1.5 1.0 05
0.08 0.030 0.011 0.015 0.017 1.7 11 1.4 0.5 2.8 2.0 1.4
0.03 0.006 0.0002 <0.0001 0.007 1.0 3 0.6 0.3 1.2 0.7 0.5
Metal 0.03 0.015 0.001 0.001 0.006 2.1 4 0.4 0.2 1.1 0.7 0.5
0.02 0.003 0.007 0.007 0.003 2.0 3 0.3 0.1 1.0 0.5 04
0.03 0.002 0.0001 0.001 0.002 1.8 9 0.8 0.2 1.9 1.3 0.6
19-67 3hr Graphite (.10 0.001 0.0003 0.001 0.014 2.5 39 7 4 3 4
0.17 0.0003 0.002 0.023 34 5.5 9 5 5 6
0.09 0.001 0.003 0.020 0.008 3.0 5.2 9 5 4 8
Metal 0.10 0.021 0.012 0.020 0.026 4.0 58 9 5 4 "2
0.06 0.002 0.002 0.007 0.004 2.6 0.5 3 2 1 04
0.09 0.003 0.002 0.005 0.009 4.2 54 9 22 5 0.7
19-66 10 hr Graphite 0.10 0.0004 0.002 0.017 5.2 4.0 5 4 2 3
0.08 0.004 0.015 5.3 34 4.5 5 4 2
0.08 0.0011 0.009 32 4 3 2
Metal 0.13 0.002 0.010 57 10 7 5
0.13 0.016 0.011 0.029 0.040 24 2.3 3 5 3 2
0.10 0.004 0.004 0.018 0.016 7.6 28 3.2 4. 4 3 3

with inventory than

with amounts directly produced

during the exposure. An overview of the table indicates
the following similar deposit intensities, for essentially
all metal-graphite pairs:

1.

When examined pair by pair, over all nuclides and
all exposure time, the deposit intensity on the
metal is about the same as on the graphite
member. No preference for either is indicated by
these data.

The salt-seeking elements for the lowest group, of
the order of 0.01 mg of inventory salt per square
centimeter being indicated. No time trends are
evident. This is regarded as a minute amount of

some form of adhering salt (film; mist droplets?)
which remained after withdrawal. '

. 89Sr runs about an order of magnitude higher.

This could have been deposited during withdrawal
operations by the ®°Kr which was present in the
gas drawn in as the capsule was removed from the
salt. (The speciments had about 1 X 10? dis min™*
cm® of #%Sr; if 8°Kr contained in the salt
entering the pump bowl, the pump bowl gas
should contain, in addition to any actual ®°Sr,
about 1 X 1073 atoms of 8°Kr per cubic cent-
imeter of helium, equivalent after decay to about
1 X 108 dis min™* cm™® of 3°Sr.)
PHOTO 97167

&, ¥ 3

L

|

avs

|
1
i
L

0.25

INCHES

Fig. 7.2. Sample holder for short-term deposition test (fits
into outer capsule with windows).

41

4. The noble metals run appreciably higher — two
orders of magnitude or so — than the other
elements.

Salt samples taken at about the same times were
generally relatively depleted in noble metals, though
they contained amounts which appeared to vary simi-
larly from sample to sample. Thus it appears that the
surfaces were capable of preferentially removing some
noble-metal-bearing material borne by the salt moving
through the sample shield.

5. The noble metals increased somewhat with time,
but considerably less than proportionately, as if an
initial rapid uptake were followed by a much
slower rise.

Most importantly, the increase in activity level with
time implies that the activity came from salt exposure
rather than any explanation related to passage through
the pump bowl gas, coupled with the further assump-
tion that the window improperly and inexplicably was
open.

6. As mentioned earlier, the activity ratios on the
inventory basis are about equal for the 123:196Ry
and '32:129MTe isotope pairs. But the longer-
lived isotope is generally somewhat lower. This is
consistent with the deposition of material which
has been held up in the system for periods that are
appreciable but not as long as the inventory
accumulation period.

Thus for noble metals the data of Table 7.2 imply the
accumulation from salt of colloidal noble-metal mate-
rial, which has been in the system for an appreciable
period but is carried by the salt in amounts below
inventory, but with the surface retaining much more (in
proportion to inventory) of the noble-metal nuclides
than it does the salt-seeking nuclides. For the periods of
exposure used, deposit intensities of given nuclides on
metal and graphite did not appear to differ appreciably.

7.4 Mist

One factor to be considered in the interpretation of
pump bowl sample data is the presence of mist in the
gas space’ within the mist shield in the pump bowl.
There is much evidence for this, none clearer than Fig.
7.3, which is a photograph of a '/, -in. strip of stainless
steel (holding electron microscope sample screens) that
was exposed in the sampler cage for 12 hr. The lower
end of the 4-in. strip was at the salt surface.

Clearly, the photograph shows that larger droplets
accumulated at lower levels, doubtless as a consequence
42

PHOTO 1855-74

SALT SURFACE

Fig. 7.3. Salt droplets on a metal strip exposed in MSRE pump
bowl gas space for 10 hr.

of a greater mist density nearer the bottom of the gas
space, at least the larger drops representing the accumu-
lation of numerous mist particles.

The mist must have been generated in one of two
ways. The first is outside the sampler shield by the
vigorous spray into the pump bowl liquid, which also
generated spray by the rising of large and small
entrained bubbles. This would be followed by drifting

of some of this spray mist around the spiral 1/8-in.-wide
aperture of the shield.

The second way in which mist could develop within
the sampler shield is by the rise of entrained bubbles
too fine to resist the undertow of the pump bowl
liquid. As salt entered the bottom of the sampler shield,
these bubbles would rise within the more quiescent
liquid and would generate a fine mist as they broke the
surface. In particular,® the liquid rushing to fill the
bubble space would create a “‘jet” droplet (as well as
corona droplets) which might be impelled to a consider-
able height — in the case of champagne, tickling the
nose several inches away.

The jet droplets can accumulate or concentrate
surface contaminants®»* on the droplet surface, being
referred to as a surface microtome by Maclntyre. Jet
droplets are likely to be about 10% of bubble diameter,
thereby a few tens of microns in diameter. It is not
certain how much of the mist within the sampler shield
was produced from outside and how much from within;
however, at least some of the mist must have been
produced within the shield, and this explanation ap-
pears sufficient to account for the phenomena ob-
served.

Kohn? shows that fine graphite dust was carried from
the surface of molten LiF-BeF, (66-34 mole %) by jet
droplets, with the implication that nonwetted colloidal
material on the fuel surface could similarly become
gas-borne.

Thus a plausible mechanism for the transport of
noble-metal fission products by mists could involve the
transport of unwetted colloidal material in the salt. This
could be transported to the surface of the liquid within
the sampler shield, by rising bubbles, and should
accumulate there to some extent if surface outflow
around the spiral was impeded. A jet droplet would
concentrate this material significantly on its surface,
and any receiver in the gas phase would indicate such
concentration. Because most droplets most of the time
would settle back into the surface, the accumulation
of noble-metal-bearing colloidal material on the surface
would not strongly be altered by this, and accumulation
would continue until by various mechanisms inflow and
outflow quantities became balanced.

References

1. J. R. Engel, P. N. Haubenreich, and A. Houtzeel,
Spray, Mist, Bubbles, and Foam in the Molten Salt
43

Reactor Experiment, ORNL-TM-3027, p. 10 (June
1970).

2. H. W.Kohn, Bubbles, Drops, and Entrainment in
Molten Salts, ORNL-TM-2373 (December 1968).

3. D. C. Blanchard, in Progress in Oceanography,
vol. 1, ed. M. Sears, Pergamon Press, Inc., New York,

1963, p. 71; “Sea-to-Air Transport of Surface-Active
Material,” Science 146, 396—97 (1964).

4. Ferren MacIntyre, “Bubbles: A Boundary-Layer
‘Microtome’ for Micron-Thick Samples of a Liquid
Surface,” J. Phys. Chem. 72,589-92 (1968).
44

Table 7.3. Data for wire coils and cables exposed in MSRE pump bowl

Each entry consists of two numbers. The first number is the total amount on the specimen, expressed in micrograms for U-235 and
in disintegrations per minute for the fission products. The second number is the ratio of the amount on 1 cm? of specimen
to the amount calculated for 1 mg of inventory salt at time of sampling (equivalent milligrams of inventory salt per square centimeter).

Power Duration of Area Total amounts on specimen
Test No. Date ) .
(MW) exposure (em™) U-235 Sr-89 Ba-140  Ce-141 Zr95 Nb-95 Mo-99 Ru-103 Ru-106 Te-132 Te-129 I-131
Hastelloy N coils
7-12 7-13-66 8 1-6 min? 6 2.63 34E10 2.1E9 7E7 3.7E11 2.7E10
-54 10-7-66 0 10 min 6 71.5E6 3 9E9 1.5E8 4.9E6 2.4E8
0.00012 0.20 0.0036 0.00535 0.14
10-12b 12-28-66 8 10 min 6 0.64 1.8E8 <4E6 1E7 4E10 3.6E8  ~TE6 8.8E8 1.8E10
After Be 0.00004 0.0017 <0.00006 <0.0006 0.22 ~0.011 ~0.0023 0.63 0.27
10-20 1-9-67 8 10 min 6 0.91 2.7E8 <4E6 14El11 1.0E9 3.6E7 34EN 2.5E9 1E10
' 2-13-67 0.00006 0.0018 0.00005 0.72 0023 0.023 24 1.3 0.12
11-8¢ 2-13-67 8 10 min 6 0.42 1.7E8 <4.5E6 56E10  4.7E8 1.5E7 1.2E11 5.8E9
0.00003 0.0012 <0.00005 0.30 0.0095 0.0081 0.86 0.07
11-12 2-21-67 8 10 min 6 1.9 1.4E8 <3Eé6 1.5E11 2.3E9 7.1E7 1.5E11 7.6E9
0.00013 0.0008 <0.00003 0.78 0.040 0.035 1.1 0.085
11-224 3-9-67 0 10 min 6 1.8 7.4E7 <6.5E6 2.5E10 1.1E9 2.7E10 9.7E8
0.00012 0.0005 <0.00006 0.18 0.019 0.26 0.012
Stainless steel cables
11-45 4-17-67 8 1-6 min Lig~2 21 3.5E5 6.5E10  2.0E9 1.5E7 9.0E10 4.6E9
0.0015 0.000003 0.35 0.027 0.026 0.66 0.047
Int.~2 20 1.0E8 18E11 14E10 4.5E8 1.3E11 1.0E10
0.0014 0.00076 0.98 0.19 0.16 0.96 0.11
Gas~2 5 8.0E7 1.1E11 2.0E9 7.5E7 3.5E10 1.1E10
0.0004 0.000006 0.60 0.027 0.026 0.26 0.12
11-50 4-26-67 8 8 hr Lig~2 35 2.5E9 3.5E11 1.E9 2E8 1.4E12 1.2E11
0.0024 0.018 19 0.090 0.066 10 1.3
Int.~2 190 3.2E9 26E12 28E1L0 5.5E8 1.0E13 9.0E11
0.013 0.023 14 0.36 0.18 72 9.5
Gas~2 60 1.7E8 2.0EIL0 2.3E12 9.2E10
: 0.0048 0.0012 0.11 16 0.97
11-51 4-28-67 7 1-6 min Lig.~2 3.5 1E7 3E10 1.3E9 4.7E7 1.4E10 1E9
0.00024 0.00007 0.17 0.017 0.015 ¢.10 0.011
Int.~2 19 1.5E7 85E10  SE9 1.5E8 5.2E10 8.5E9
0.0013 0.00011 0.47 0.064 0.049 0.38 0.091
Gas~2 6.0 3E7 4E10 25E9 9.5E7 3.0E10 6E9
) 0.0004 0.00022 0.22 0.032 0.031 0.22 0.064
11-54 5-5-67 8 1-6 min Lig.~2.5 6.4 1.6E7 4.2E10  2.5E9 1.5E7 1.7E10 2.2E9
0.00045 0.00011 023 0.031 0.024 0.12 0.023
Int.~2.5 5.6 3.5E7 7.2E10 1.0E10 3.0E8 6.1E10 7.0E9
0.00039 0.00025 0.39 0.13 0.095 0.44 0.074
Gas~2.5 2.0 3.5E7 42E11 4E9 " 1.3E8 2.8E10 S.5E10
0.0004 0.00025 0.23 0.050 0.041 0.20 0.058
11-58¢ 5-10-67 0 1-6 min Lig.~25 8.4 3E6 SE9 1.3E9 4.5E7 1.7E10 1.1E10
0.00058 0.00002 0.044 0.017 0.014 0.19 0.014
Int. 3.3 6Ee6 1.5E10 1.5E9 4.5E9 7E9 2.8E10
0.00023 0.00004 0.13 0.019 0.014 0.078 0.035
Gas 2.5 7Ee6 58EI0  2.3E9 7.5E7 4E9 6.0E10
0.0017 0.00005 0.51 0.030 0.024 0.044 0.075
12-6 6--20-67 0 1-6 min Liq. 22 8.2E6 1.1E9 3.8E7
0.0015 0.00006 0014 0.012
Int. 46 6.0E6 1.7E9 6E7
0.0032 0.00004 0.022 0.019
Gas 60 . 3.0E6 3.8E9 1.4E8
0.0042 0.00002 0.049 0.044
12:27 7-17-67 8 1-6 min Liq. 8.2 9ES 6E9 SE8 2E7 9E11 1.4E9
0.00057 0.000009 0.033 0.010 0.004 0.060 ) 0.018
Int. 5.6 1.0E7 2E10 4E8 4E8 1.8E10 2.4E9
0.00039 0.00010 0.11 0.008 0.078 0.12 0.030
Gas 6.6 1.3E7 23E10 23Ell 9E7 4.0E10 3.7E9
0.00046 0.00013 0.13 0.043 0.018 0.28 0.046

4Before startup. Corresponds to run 7 shutdown after beryllium addmon

b after beryllium addition.

€ After addition of 11.66 g of beryllium,

d3) days after shutdown.

€Vs shutdown May, before drain, 2.3 hr after shutdown, 2 hr after stopping fuel pump
fRefitled. Before power (vs shutdown May 8).

45

Table 7.4. Data for Miscellaneous capsules from MSRE pump bowl

No. Date Capsule Basis Y-91 Cs-137 Sr-89  Ba-140 Nd-147 Ce-141 C(Ce-144  Zr95 Nb-95 Mo-99 Ru-103 Ru-106 Ag-111 Te-132 Te-129m [-131 Sb-125 Li Be U-233
17-8 1-24-69 Be addition capsule Total 2.6E11 3.5E10 23E13 16E14 24E12 14El1l 4 9E13 5.2E12
vs inv? 7.2 24 2800 1990 450 28 700 185
17-11 1-29-69 Cr addition capsule  Total 9.9E10 3.1E10 1.2E12 6.2E13 3.7E11 1.7E10 1.4E14 5.0E11 1.0E1l3
vs inv 1.8 1.5 140 630 45 36 1580 340 260
18-3 4-17-69 Zr addition capsule Total 4.5E9 4.1El11 62E10 1.0E11 6.9E13 9.4E13 9.1E11 49E10 7.6EI10 3.2E13 9.8E11 1.9E12
vs inv 0.9 4.3 1.2 1.1 1300 920 24 9 140 340 160 28
18-7 4-25-69 Zr addition capsule total 5.6E10 1.5E12 TE7 86E11 1.3E12 3.3E13 1.2E14 28E12 14E1l1 2.1E1l 36E13 1.1E12 1.1E12
vs inv 11 14 0.0004 16 13 560 760 69 25 300 250 160 13
18-11 5-1-69 Ni capsule Total 2.0E9 7.3E10 4.5E9 1.1IE10 1.5E12 16E13 9.2E11 3.0E10 3.2FE11 6.8E13 1.1E12 ' 8.1E12
vs inv 0.40 0.63 0.08 0.10 23 94 21 5 410 440. 140 88
18-17 5-8-69 FeF, addition Total 1.1E10 8E9 3.3E11 1.4E11 2.4E10 2.5E11 5.8E12 1.0E11 4.4E9 1.2E10 1.6E13 1.6E11 | 8.6El11
capsule vs inv 2.1 0.07 2.1 2.4 0.21 3.6 40 23 73 16 120 20 10
1820 5-12-69 Cu 8 hr exposure Total 2.8E11 1.0E10 24E11 4.3E11 1.5E11 84El11 3.4E11 2.1E11 4.6El11 2.7E13 4.3El1l1 16E10 5.5E11 9.6E13 1.1E12 7.6E12
2.5 2.0 2.0 2.6 2.4 4.7 58 1.9 6.4 170 9 2.7 720 670 140 - 84
1823 5-15-69 Be addition capsule Total 1.1E11 1.7E12 5.7E11 26E12 1.0E13 6.3E14 1.1E13 S§.2E11 7.5El1l1 1.7E14 3.7El12 ' 2.5E13
vs inv 20 13 10 22 135 4300 220 89 1000 1300 450 280
18-28 5-20-69 Be surface Total 7.0E10 1.3E12 90E11 26E12 9.1E11 39E13 79E11 36E10 76E10 1.8E12 24E11 1.8El12
tension effects vs inv 13 10 15 21 11 270 16 6 100 62 29 20
19-12 8-19--69 Enrichment capsule Total 2.0E9 5.2E8 9.7E9 2.2E8 9.0E8 2.0E9 1.3E9 8.8E10 9.2E9 1.3E6 ' S5E9 1.7E8 1.5E8
vsinv  0.045 0.10 0.22 0.11 0.028 0.041 0.025 1.2 0.88 04 44 20 2
19-52  10-12—-69 “DRG”-Cudummy Total 7.6E8§ 46E8 44E10 2.1E9 5.0E8 70E8 S.1E8 6.1E8 4.1E10 1.3E12 44E10 2.3E9 3.5E9 S5.0E7 0.58
Leach vsinv 001 0.08 0.51 0.016 0.01 0.006 0.009 0.007 0.60 7.8 1.6 0.43 1.0 0.19 0.005
19-61 10-21-69 Nb strip and Ni 3.0E9 4.2E9 S5.2E11 2.8E10 2.0E10 1.6 345u
capsule, Harold 0.56 6.0 34 24 0.24 0.013 0.05
Kohn )
Nb strip Total 1.4E10 6ES8 8.8E9 2E10 8E9 1.5E10 7E9 1.0E10 6.4E9 1.6E12 2.0E10 9ES 3.6E9 1.2E11 3.3E9 ! 2.7E10 120 6.2
vsinv (.16 0.10 0.09 0.13 0.16 0.10 0.12 0.10 0.09 10 0.6 0.17 5 0.8 0.2 ] 0.3 1.0 0.09
Ni cap Total 2.1E9 5ES8 74E10 4E9 1.4E9 14E9 9E8 - 1.5E9 2.8E11 R8.1E12 1.5E11 9E9 1.5E11 3.3E12 1.2E11 2.1E11 2.1 0.96
vsinv  0.024 0.08 0.8 0.03 0.02 0.01 0016 0.016 4 50 4.6 3 21 22 8 23 0.018 0.014

ICalculated inventory per gram of salt assuming no losses.
46

Table 7.5. Data for salt samples from double-walled capsules immersed in salt in the MSRE pump bowl during uranium-233 operation

Each entry in the table consists of two numbers. The first number is the radioactivity of the isotope on the outside surface of the capsule expressed
in disintegrations per minute per square centimeter. The second number is the number of equivalent milligrams of inventory salt per square centimeter

of capsule surface, defined as the amount of saft that would contain the analyzed quantity assuming uniform distribution in the fuel salt.

Power

Minutes in

Sample Date Sr-89 Y-91 Ba-140 Cs-137 Ce-141 Ce-144 Nd-147 Zr95 Nb-95 Mo-99 Ru-103  Ru-106 Ag-111 Sb-125  Te-129m Te-132 1-131
(MW)  pump bowl i
18-12 5-2-69 8.0 2.3E7 1.0E7 1.2E8 7.7E7 1.2E10: 2.9E9 14E8 1.6E8 7.1E8 3.1E10 3.3E8
0.13 0.18 1.13 1.18 70.59 63.93 24.54 202.02 89.64 198.57 3.54
18-19 5-9-69 8.0 5.3E7 1.6E7 2.1E7 6.9E8 1.4E10 1.8E8 6.9E6 9.8E7 8.6E8 6.6E10
0.33 0.28 0.19 9.63 90.37 3.95 1.16 132.28 106.06 485.00
18-44 5-29-69 6.9 1.3E9 8.6E7 7.5E6 5.9E7 3.5E7 7.0E7 3.8E7 2.6E8 4.1E10 1.5E9 7.3E7 1.2E8 4.3E6 1.3E9 5.3E10 1.0E9
9.76 0.72 1.39 0.32 0.57 1.22 0.31 3.00 338.80 31.25 11.90 172.84 14.91 156.86 468.81 12.63
1845 6—1-69 0.0 29E8 4 3E9 2.7E7 4.5E7 1.2E10 2.2E11 2.4E9 1.0E8 2.0E8 9.7E6 4.7E9 1.6E11 3.6E9
2.18 795.11 ¢ 044 0.36 133.58 1691.92 48.13 16.45 288.37 33.18 545.45 1294 .77 44.07
18-46 6—1-69 0.0 6.4E8 2.2E9 4 9E8 2.6E7 65E8  2.8E8 2.3E8 4.1E8 2.6E9 3.2E10 5.6E8 2.6E7 7.0E7 2.1E6 2.3E9 6.8E10 3.6E9
4.82 17.81 3.17 4.79 3.52 4.60 4.14 3.25 28.77 259.56 11.19 421 104.14 7.12 265.76 589.08 44 88
19-9 8-18-69 0.01 60 6.7E6 3.5ES 1.6ES 3.7ES 8.2E5 1.5E6 3.1E9 1.3E6 1.9E7 4.3E6 24E7 6.6E6 3.6E6
0.15 0.0076 0.0302 0.0109 0.0166 0.0295 40.60 31.48 1.56 0.81 14.61 201.35 35.20
19-24 9-10-69 0.01 31 7.4E7 7.2E6 T.0E6 5.8E6 3.4E6 3.9E6 2.3E6 7.2E8 2.3ES 6.6E9 1.7E8 1.2E7 34E7 2.0E8 5.5E9 2.7E8
1.31 0.13 0.11 1.05 0.0534 0.0777 0.0913 11.79 33.24 78.61 9.51 2.28 93.66 6751 70.49 6.20
19-36 9-29-69 5.5 43ES 3.2E6 7.2E6 2.8E6 8.3ES 9.5ES 2.6E6 7.0E9 3.0E10 2.6E8 1.5E7 4,2E7 7.4E8 1.3E10 7.8E8
6.75 0.0536 0.0948 0.49 0.0106 0.0185 0.0387 104.13 382.12 12.11 2.72 115.65 203.71 181.53 17.59
1942 10-3-69 7.0 4.1E7 5.4E6 3.8E6 4.6E6 3.7E6 3.2E6 2.9E6 7.0E7 3.2E9 29E8 1.7E7 3.6E7 1.9E8 8.1E9 1.3E9
0.60 0.0844 0.67 0.0538 0.0720 0.0956 0.0418 1.05 29.13 12.41 3.14 80.54 48.15 82.82 25.06
19-44 10-6-69 8.0 180 8.9E7 3.1Ee6 2.5E7 8.9E6 5.8E6 3.7E6 1.1E7 1.7E9 8.5E10 8.8E8 4.2E7 2.3E8 2.0E9 9.6E10 1.5E10
1.22 0.0461 0.24 1.56 0.0605 0.0703 0.15 24 .64 595.46 3393 7.75 428.45 451.26 768.00 237.11
1947 10-7-69 8.0 17 3.1E8 2.4E7 2.0E7 1.0E7 9.2E6 6.4E6 4 9E6 8.4E6 1.3E9 3.5E10 1.3E9 7.9E7 1.5E8 9.1E8 7.1E9 1.1E9
. 4.08 0.34 0.18 1.78 0.0916 0.12 0.12 0.11 19.48 228.80 49.22 14.43 267.20 195.62 53.23 15.56
19-55 10-14-69 8.0 20 2.1E8 2.5E7 2.4E6 1.6E7 9.5E6 1.1E7 1.1E7 9.1ES§ 5.8E10 6.4E8 3.2E7 3.7E8 1.4E9 5.2E10 2.8E9
2.41 0.18 0.42 0.13 0.17 0.21 0.13 13.13 332.69 19.79 5.73 509 .89 248.34 332.63 32.24
19-58 10-17-69 0.01 30 1.2E8 3.7E6 1.3E7 2.0E6 8.2E6 4 4E6 3.1E6 3.8E6 9.8E8 2.1E10 3.7E8 2.1E7 1.0E8 4.5E8 8.2E9 2.7E8
1.26 0.0436  0.0887 0.34 0.0634 0.0789 0.0555 0.0420 1395 127.27 10.93 371 134.44 76.08 54.30 3.03
19-59 10-17-69 0.01 30 6.3E8 9.1E6 3.0E7 3.1E6 7.6E6 5.4E6 6.2E6 6.7E6 . 1.4E9 4.8E10 1.0E9 59E7 8.2E7 1.6E9 3.3E10 4 3E8
6.85 0.11 0.21 053 0.0587 0.0970 0.11 0.0737 2052 302.08 30.67 10.48 111.72 266.85 225.25 4.88
19-76 10-30-69 8.0 5.6E8 7.9E6 3.2E7 2.5E7 1.1E7 5.8E6 S4E6 9 9E6 3.2E8 2.6E9 7.2E8 4.1E7 6.3E7 6.7E8 1.8E10 4 4E8
5.07 0.0795 0.20 4.10 0.0979 0.0866 0.0936 4.21 15.56 17.55 7.01 79.23 92.49 121.41 4.64

0.0689

*Equivalent mg inventory salt means the amount of salt which would contain the analyzed quantity assuming uniform distribution in the fuel salt.

47

Table 7.6. Data for gas samples from double-walled capsules exposed to gas in the MSRE pump bowl during uranium-233 openation

Each entry in the table consists of two numbers. The first number is the radioactivity of the isotope on the outside surface

of the capsule expressed in disintegrations per minute per square centimeter. The second number is the number of equivalent milligrams
of inventory salt per square centimeter of capsule surface, defined as the amount of salt that would contain the analyzed quantity assuming uniform distripution in the fuel salt.

Sample  Date  ower Minutesin g 0 Y91  Bald0 Cs137  Cel4l  Ce-l44 Nd-147 295  Nb95  Mo-99  Ru-103  Ru-106 Agill  Sb-125 Tel29m  Te-132 1131
(MW) pump bow!
1821 5-12-69 19 60 3768 13E7  49E7  LIEB  SOE6  4.4E7 1IE6"  6.5E8  32EI0  69E8  42E7  49E7 44E8 34E10 1.5E9
298 012 030 2039 00279 0.7 00098  8.86 20524  14.57 707 65.00 5360 23932 16.55
1825  5-17-69 7.0 60 29E7 2.2E6 61E6 - 11E8  12E10 14E8  72E6  33E7 6.0E8 24E10
0.18 0.0381 00516 136  79.37 2.95 120 44.44 7140 18025
1829  5-21-69 7.8 60 2088 1.2E7 65E6  36E6  2.0E6 18E6  I.IE8  -1IEI0  1SE8  26E6  10E7 3.0E8 1.4E10 11E9
_ LS1 0099 122 00194 0.0337 00152 139 7007 301 043 1386 3532 10219 12.39
1842  5-28-69 6.6 60 24E8 17E6 1JE7  87E6 3.7E6  23E6  6JE6  36E6  63E7  STE9  84E7  44E6  2.2E7 6.5E8 23F10 1.2k9
183 00139 0.1 161 00198 00383 012 00296 073 4849 1.64 071 3281 7612 21121 14.38
19-13  8-21-69  0.008 60 4SE7  10E6 14E6  27E6  28ES  7.ES 17E6  18E8  69E6  7SE7  19E7 1.5E6 1.9€7 2.8E6 166
105 00229 078 051 00088 0.0145 00345 242 7459 6.56 3.64 5.70 12.44 36.37 16.34
1914 8-21-69  0.008 53 31E7  18E6 22E6  17E6 87ES  2.0E6 I4E6  60E7  46E6  13E7  13.2E6 1.9E7 1.9E7 2.5K6
073 00414 120 031 00274 00413 00272 080  48.42 115 0.62 1223 23188 25.34
19-15  8-21-69  0.008 60 12E7  LIE6 B8.8ES  43ES  9.2ES 66ES  46E7  13E6  83E6  2.4E6 98E4  3.IE7 3.5E7 1.5E8
028 0.0239 0.16 00136 00188 00132 062 1365 0.74 0.45 0.37 2048 41851  1538.46
1916  8-21-69  0.008 60 7857  26E6 B86ES 4286  LIE6  2.5E6 19E6  40E8  20E7  10E8  2.6E7 2.3ES 18E7  6.6F7 456
182 00597 049 078 00335 00501 00384 540 19362 8.84 4.93 0.85 5100 77034 45.70
19-19  9-4-69 5.5 20 23E8  43E6 11E7  LIET 69ES  18E6 6.1E5  26E8  41E9  30E8  19E7  3IE7 1.3E8 6.8E9 6 8E8
439 00855 024 200 00131 00367 00107 375 5183 1877 359 1ILI 52.78 97.00 20.51
1920 9-4-69 5.5 20 22E8  25E6 1.2E7  83E6  64ES  1.1E6 68E7  22E8  18EI0  34E8  16E7  44E7 1.3E8 7.7E9 7.9E8
430 00495 024 152 00120 00223 1.20 324 21811 2114 299 15379 5176 106.77 23.54
1923 9-10-69 0.0l 20 99E7 35E6 29E7  1.0E7 2.1E6 24F6  12E6 97E7  3.1E8  S6E9  22ES  14E7T  21E7  1.7E7 1.0E8 28E9 2.2E9
174 00633 044 184 00322 00477 00469 159 448 6285 1204 267 5676 6121 34.15 3511 50.57
1928 9-23-69 5.5 20 23E8  65E6 14E7  63E7T  36E6  3.6E6 24E6  74E8  11EI0  8.1E8  4.0E7  14E8 4.7E8 1.6E10 1.8F9
365 011 018 1123 00472  0.0697 00368 1111 11654 3807 750 33935 13101 18898 37.96
1929 9-23-69 5.5 2 22E8  26E7 38E7T  S52E6 18E7 16E7 93E6 1SE7  3.1E8  95E9  96E8  48E7  1SE7 4.4E8 LIE10 L4E9
345 044 048 094 - 023 032 031 023 463 10178 4497 8.85 184.10 12307 12560 28.58
1937 9-30-69 0.1 20 4SE7 9SES 87ES  14E6 13E4  19E4 38ES  19E8  32E9  12E8  78E6  SAE6 9.4E7 1.5E9 6.4E8
070 00156 00I14 025 00002 0.0004 00057 283  40.18 5.53 145 1457 25.55 20.10 1421
1938 10-1-69 5.5 2 66E7 S552ES 42E8  19E6 11ES  2.7E4 3.0E4  SOE7T  43E8  37E8  24ET  12E7 7.5E7 1.2E9 1.4E9
© 101 00084 532 0.33 00013 0.0005 0.0005 0.5 507 1655 455 3232 20.03 16.29 30.61
1941 10-3-69 1.0 20 14E8  7.2E6 19E7  28E6  36ES  29ES S2ES  1SER  43E9  16E8  1LI1E7  2.7E7 3.2E8 12E10 1.6E9
203 011 022 049 00042 0.0056 00075 229 4012 6.87 201 6111 8099  126.55 30.32
1946  10-7-69 8.0 200 33E8  2.1E7 76E7  4.1E6  12E6  10E6 47ES  26E7  63E8  18E8  12E7  SOE7 26E8 12E10 1.9E9
435 030 071 0.71 00122 00193 00062  0.39 421 6.64 221 8745 56.73 92.73 27.90
19-54  10-14-69 8.0 2 67E8 - 46E6 18E7  28E6 11E6 12E6 3SES 1786  1.1E9  B86E9  16E9  11E8  23E7 3.SE8 1.1E9 3.2E8
762 00565 0.3 048 00092 00218 00071 00201 1556 4955 4870 1973 3164 61.89 6.70 3.68
19-56  10-15-69 0.0l 20 10E8 3.7E6 7.8E6  SOE6 38E6 24E6  24E6 2.1E6  14E8  73E9  27E8  15E7  22E7 3.4E8 9.8E9 8.1E8
114 00450 00559 086 00305 00427 00436 00233 206 4268 8.23 274 3028 59.35 6385 9.35
1962  10-22-69 8.0 2 34EB 3286 3.0E7  9.6E6 15E6 14E6 12E6 19E5  23E8  20EI0  67ES8  33ET  36E7 S7E8 2SEI0 249
345 00355 0.20 162 00110 00251 00214 00020 313 12338 1819 570 4706 89.18 16480 26.41
1964  10-22-69 8.0 40 29E8  13E7  36E7  4.1B6  64E6 3SE6  43E6  S52B6  16E8  9.5E9  3SE8  1.8E7  19E7 32E8 1IEI0  8.5E8
291 014 024 069 00451 00607 00731 00535 223  58.20 9.39 305 2503 49.00 76.30 9.23
1965  10-23-69 8.0 40 47E8  1IE7 47E7  15E6  10E6  74ES 13E6  30E8  16E10  78ES  43ET  46E7 6.9E8 23E10  6.8E8
465 012 030 1.26 00071 00129 00129 412 10027 2094 749 5966 10528 154.07 7.33
1970 10-28-69 8.0 40 IIE8  8.5E6 49E6  74E6  4.4E6 98E6 . 85E7  66EI0 17ES  7.1E7  T.IE7 15E10 54E9 2.9E9
098  0.0879 082 00487 00746 00949 113 40202 00043 1210  89.09 2107.35 35.69 31.03
1973 10-29-69 8.0 40 48E8 11E7 61E7  6.1E6 7.AE6 38E6 44E6  24E6  12E8  23EI0  14E9  T2E7  6JE7 1.8E8 5.3E9 1.0E9
a43 011 037 101 00464 00650 00708 00226 153  13670. 3460 1221 8445 25.34 35.10 10.87
1977 10-31-69 8.0 a1 38E8  2.5B6 4.6E7  80ES 26E6 13E6 1266 3.1E6  23E8  LIEI0  82E8  75E7  4.1E7 11E9 39E10 2.8F9
346 00250 028 0.3 00167 00221 0019 00293 303 6894 1975 1276 5140 14652 256.58 29.45
1978 10-31-69 8.0 a0 63E8  12E7 78E7  68E6 10E7 62E6 69E6 86E6  SSE8  39EI0 19E8  6.1E8  82E7 13E9  43E10  29E9
571 012 047 113 00655 0.0 011 00803 745 23360 466 10392 10252 17381 279.97 30.49
1979 11-2-69 08 40 S4E7  7.4ES  18E6  14ES  I4ES  LIES 20ES  42E7  19E9  70E7  50E6  LJET7 1.0E8 2.2E9 3.6E9
048 00072 00109 00237 00009 00019 00020 053 1233 165 085 1609 13.96 15.23 37.82
209 12-1-69 80 40 2.2E8 41E7  34E6  36E6  34E6 47E6  1.8E8 35E7  3.9E6 6.0E8 3.2F9
_ 261 0.54 056 00345 00612 00556 967 121 0.70 122.12 74.07
2012 12-2-69 80 40 2767 6586  LIET 16E6  47E8  1SEN0  43E8  2JET  S.JE7 1.3E8 39E9 2458
031 00808 0.13 00179 581 11709  14.32 412 14067 26.45 35.24 4.96
2027 12-10-69 80 40 LIE7  49E6 76E6  2.8ES  3.7E6 266 33E6  16E8  3IE9  90E7  3.6E6  4.5E7 LIE8  38F9 3.0F8
0.1  0.0549 00641  0.0447 00283 00444 00340 196  19.37 2.58 063 728 18.46 2.57 4.09
032 12-12-69 0 40 75£6  1IES 23ES  LIES 26E4  2.8E4 10ES  34E7  3E8  25E7  18E6  9.8ES 45E6  48E7 1.1E8
00748 00012 0.0018  0.0178 00002 0.0005 00011 041 193 0.69 031 150 0.74 0.33 1.46

48

8. GAS SAMPLES

Fuel salt was continuously sprayed through the pump
bowl purge gas to permit removal of all possible xenon
and krypton fission products. The purge gas passed into
the off-gas lines and thence to charcoal beds.

The examination! of surfaces exposed in the gaseous
region above the liquid within the sampler shield spiral
in the pump bowl indicated the presence of appreciable
concentrations of noble metals, raising a question as to
what might actually be in the pump bowl gas. (The data
on surfaces exposed in the pump bowl are in a separate
section.)

8.1 Freeze-Valve Capsille

The transfer tube and spray shield region above the
pump bowl liquid level were not designed as a facility
for sampling the pump bowl gas. However, the inven-
tion of a freeze-valve capsule sampling device? (men-
tioned earlier for salt samples) made it possible to
obtain useful samples from the gas region within the
spray shield. It was required that the sampling device be
small enough to pass freely through the bends of the
1%4-in. diam. sampling pipe and that it should operate
automatically when it reached the pump bowl.

The device, the original form of which is shown in
Fig. 8.1, operated satisfactorily to furnish 20-cc samples
of gas. The capsule was evacuated and heated to 600°C,
then cooled under vacuum to allow the Li, BeF, in the
seal to freeze. The double seal prevented loss of
Li;BeF; from the capsule during sampling. The
weighed, evacuated capsule was lowered into the pump
bowl through the salt sampling pipe and positioned
with the bottom of the capsule 1 in. above the fuel salt
level for 10 min. The freeze seal melted at the 600°C
pump bowl temperature, and gas filled the 20-cc
volume. The capsule was then withdrawn to 2 ft above
the pump bowl and allowed to cool.

The cooled resealed capsule was withdrawn from the
sampling pipe and transported in a carrier to the
analytical hot cells. A Teflon plug was placed over the
protruding capillary, and the exterior of the capsule was
thoroughly leached free of fission product activities.
The top of the capsule was cut off, and the interior
metal surface was leached with basic and acid solutions
for 1 hr. The bottom of the capsule was then cut at two
levels to expose the bottom chamber of the capsule.
The four capsule pieces were placed in a beaker and
thoroughly leached with 8 N HNOj; until the remaining
activity was less than 0.1% of the origina! activity. The
three leach solutions were analyzed radiochemically.

ORNL-DWG 67-4784A

STAINLESS STEEL
CABLE

i | e ——— 3/4-in. OD NICKEL
VOLUME f f
ZOCC\*\ f
! i
1 H
1 /
f !
LipBeFy / NICKEL CAPILLARY

Fig. 8.1. Freeze valve capsule.

Using this device the first gas sample was obtained in
late December 1966, during run 10. In all, eight such
samples were taken during operation with 235U fuel.
Data for these samples are shown in Table 8.2 (at the
end of this chapter).

8.2 Validity of Gas Samples

There are at least two particular questions that should
be addressed to the data obtained on gas samples.

First, do the data indicate that a valid sample of
pump bowl gas was obtained?/Second, what fraction of
the MSRE production of any given fission product
chain is represented by flow to off-gas of a gas of the
indicated composition?

49

The first question was approached by estimating the
activity of 50-day ®°Sr resulting from the stripping of
3.2-min ®*?Kr into the pump bowl purge gas. For
example, the activity of full-power samples 11-46,
11-53, and 12-26 averaged 3.8 X 10° dis/min for 3°Sr,
indicating that the gas contained 2.0 X 10'? atoms of
89Sr per cubic centimeter of capsule volume. If the
only losses of 3°Kr were by decay or 100% efficient
stripping in the pump bowl, then

®°Kr stripped/min _  F

89Kr produced/min  F + A

where A is the decay constant of 3.2-min 3°Kr
(02177 min '), F is the fraction of fuel vol-
ume that passes through the pump bowl per min-
ute; for 50 gpm spray flow and 15 gpm foun-

tain flow and a salt volume of 72 cu ft, F =0.121
Thereby, F/(F+ X)=0.357. At nominal full power of 8

MW and a fission yield of 4.79%, the rate of production
of ®°Kr is 7.25 X 107 atoms/min. Thus 2.59 X 1017
atoms/min enter the pump bowl. Helium purge flow at
the pump bowl temperature and pressure is 8370
cc/min, so that the gases mix for a concentration of
3.09 X 103 atoms of ¥°Kr per cubic centimeter. With
"agas space-in the pump bowl of 54,400 cc, the average
concentration in the well-mixed pump bowl is 1.28 X
10'* atoms of %% Kr per cubic centimeter. The
difference between this and the entrant concentration
represents the ®°Rb and ®°Sr produced in the pump
bowl gas phase. Doubtless most of these atoms return
to the salt.

The observed concentration, 2.0 X 10'® atoms of
89Sr per cubic centimeter of capsule volume, is
somewhat greater than the calculated average ®*° Kr
concentration in the pump bowl gas (maximum, 1.3 X
10" atoms of *? Kr per cubic centimeter of pump bowl
gas). Possibly some of the 2°Rb and 3°Sr atoms from
89Kr decay in the pump bowl remained gas-borne,
entering with the sample. The concentration of 8%Sr
in the fuel salt was about 1.2 X 10'® atoms/g.
Thus all of the observed ®°Sr activity in these
gas samples corresponds to about 2 mg of fuel

salt per cubic centimeter of gas. However, the 23U

contents of these samples were 9, 23,and 25 ug
per 20-cc sample, and the salt contained about 15,000
pg of 235U per gram, implying entrained fuel salt
amounts of 0.03, 0.08, and 0.08 mg/cc. Very possibly,
mists containing fuel salt could remain stable within
the relatively tranquil sampler shield for a time suffi-
cient to accumulate on their surfaces much of the gas-
borne 3°Rb and ®°Sr and to enter the capsule with the
sample gas.

Two other gas samples (11-42 and 12-7) were taken
while the system had been at negligible power for several
hours. Concentrations |of ®°Srwere about an order of
magnitude lower (0.06 and 0.23) in accord with the
above viewpoint; of course, purely gas samples should
go to zero, and purely salt samples should be unaffected
within such a period.

Thus we conclude that the samples are quite possibly-
valid gas samples, probably with some mist involve-
ment. :

The saltseeking elements, including zirconium, cer-
ium, and uranium, do not involve volatilization as a
means of entering the gas phase. It is shown in Table
8.2 that an individual gas sample contains quantities of
these elements equivalent to a comnion magnitude of
inventory salt. Thus the conjectured presence of salt
mist in the gas samples appears verified.

The noble metals appeared in the gas samples in
quantities that are orders of magnitude higher (in
proportion to inventory salt) than was found for the
salt-seeking elements.

Thus appreciable quantities of noble metals are
involved in the gas-phase samples. If they are rapidly
stripped as a result of volatility, then the observed
concentrations represent the stripping of a major part
of the fission products in this way. However, the
volatile compounds of these elements thermo-
dynamically are not stable in the fuel salt. If the noble
metals are not volatile, then some form of aerosol mist
or spray is indicated. The relationship in this case
between aerosol concentrations within the sampler
shield, in the gas space above the violently agitated
pump bowl liquid surface, and in the shielded
approaches to the off-gas exit from the pump bowl have
not been established. However,> in the pump bowl
proper, “the spray produced a mist of salt droplets,
some of which drifted into the off-gas line at a rate of a
few grams per month” (3.6 g/month is equal to 108 g
of salt per cubic centimeter of off-gas flow). As we shall
see, in the samples taken during 233U operation, the
quantity of gas-borne salt mist in our samples, though
low, was higher than this.

8.3 Double-Wall Freeze Valve Capsule

A number of gas samples were taken during 233U
operation using the freeze-valve capsule, with capsule
volume of 30 cc; results are shown in Table 8.3. These
extended across run 17. However, many aggressive acid
leaches of the capsule were required to reduce external
activities to wvalues assuredly below the contained
sample, and sometimes leakage resulted. To relieve this
and also to provide a more certain fusible vacuum seal,
the double-walled capsule sampler shown in Fig. 8.2
was employed. This device contained an evacuated
copper vial with an!internal nozzle sealed by a soldered
ball. The nozzle tip was inserted through and welded at
the end to an outer capsule tip. A cap was welded to
the top of the outer capsule, completely protecting the
inner capsule from contamination. In practice, after a
sample obtained using this device was transferred to the
High Radiation Level Analytical Laboratory, the tip was
abraded to free the nozzle, and the upper cap was cut
off, permitting the inner capsule to slide directly out
into a clean container for dissolution without touching
any contaminated objects. Usually the part of the
nozzle projecting from the internal capsule was cut off
and analyzed separately.

Samples 19-77, 19-79, 209, 20-12, 20-27, and 20-32,
in addition, employed capsules in which a cap con-
taining a metal felt filter (capable of retaining 100% of
4-u particles) covered the nozzle tip. This served to
reduce the amounts of the larger mist particles carried
into the capsule.

Data from all the gas samples taken during the 233U
operation are shown in Table 8.3. Reactor operating
conditions which might affect samples are also shown.
The data of Table 8.3 show the activity of the various
nuclides in the entire capsule (including nozzle for

ORNL-DWG 70-6758

CUT FOR

SAMPLE REMOVAL-= ~=——NICKEL OUTER TUBE

4fljj> VACUUM

COPPER INNER CAPSULE

|| —BALL RETAINER

BALL WITH SOFT SCLDER SEAL

L/

e COPPER NOZZLE TUBE
ABRADE FOR J
SAMPLE REMOVAL

Fig. 8.2. Double-wall sample capsule.

50

double-walled capsules), divided by the capsule volume.
Some attributes of a number of the samples are of
interest,

Samples 19-23, 19-37, and 19-56 were taken after the
power had been lowered for several hours.

Samples 19-79 and 20-32 were taken after reactor
shutdown and drain. Some salt constituents and noble
metals still remain reasonably strong, implying that the
salt mist is fairly persistent.

Sample 70 was taken “upside down”; strangely, it
appeared to accumulate more salt-seeking elements.
Samples 19-29, 19-64, and 19-73 were ‘“‘control”
samples: the internal nozzle seal, normally soldered,
was instead a bored copper bar which did not open. So
data are only from the nozzle tube, as no gas could
enter the capsule.

8.4 Effect of Mist

As it was evident that all samples tended to have salt
mist, daughters of noble gases, and relatively high
proportions of noble metals in them, a variety of ways
were examined to separate these and to determine
which materials, if any, were truly gas-borne as opposed
to being components of the mist. It was concluded that
the lower part of the nozzle tube (external to the gas
capsule proper, but within the containment capsule)
would carry mostly mist-borne materials; some of these
would continue to the part of the nozzle tube that
extended into the internal capsule, and of course was
included with it when dissolved for analysis. The
amounts of salt in nozzle and capsule segments were
estimated for each sample by calculating and averaging
the amounts of “inventory” salt indicated by the
various salt- seeking nuclides.

For all nuclides a gross value was obtained by
summing nozzle and capsule total and dividing by
capsule volume. The “net” value for a given nuclide was
obtained by subtracting from the observed capsule
value an amount of nozzle mist measured by the
salt-seeking elements and nuclides contained in the
capsule; the amount remaining was then divided by
capsule volume.

This was done for all gas samples taken at power
during runs 19 and 20. The results are shown in Table
8.1, expressed as fractions of MSRE production indi-
cated by the samples to have been gas-borne, with gross
values including, and net values excluding, mist. Median
values, which do not give undue importance to occa-
sional high values, should represent the data best,
though means are also shown.

The median values indicate that only very slight net
amounts of noble metals (the table indicates '°Ru as a
51

Table 8.1. Gas-borne percentage of MSRE /production rate
Double-wall capsules, runs 19 and 20 (sampled during power operation)

Gross? Net

Isotope Stripping®
Number Range Median Mean Number Range Median Mean (caled)
Isotopes with Gaseous Precursors
89r 13 0.3-17 5.2 6.5%1 11 - 0.06—15 3 57%1.2 14
137¢cs 11 6-98 22 33+6 9 ~1.6-91 23 25 +6 18
oy 13 0.005-3 0.08 0.36 £0.17 11 ~0.11-0.08 0.003  0.006 £0.010 0.07
140g, 13 0.005-0.4  0.08 0.10 £0.02 11 —0.004-0.18 0.027  0.056 £0.013 0.16
Salt-Seeking Isotopes
95 7r 13 0.002-0.3  0.04  0.057+0.014 11 ~0.007-0.05 0.006  0.012+0.004
Hlc 13 0.002-0.2  0.009 0.025%0.011 10 -0.03-0.009 —0.0003 —0.003 *0.003
144 13 0.01-1.7  0.22 0.32 £0.09 11 ~0.12-0.42 0.007 0.05 +0.03
147Nd 9 0.0001-0.1  0.012  0.021 £0.007 9 ~-0.01-0.01  -0.001  0.002 £0.002
“Noble” Metal Isotopes
*SNb 13 0.07-7 0.7 1.9 £0.5 11 -0.2-3.6 0.4 0.9%0.2
Mo 13 0.16-16 1.0 2.7%0.9 11 -0.6-7.3 0.3 1.5 £0.5
M1ag 13 0.2-20 LS5 40%1.1 11 -0.9-4.1 0.3 0.710.3
193Ru 13 0.31-20 L8 43%1.2 11 0.05-10 1.1 2.3£0.7
106Ru 13 3.8-67 13 224 11 -2-36 6 11%3
Tellutium-lodine Isotopes
1297 13 0.3-27 1.8 51%1.5 11 —0.11-4 0.1 1108
1327e 13 0.03-23 1.0 3.5%1.2 11 ~12-2 ~0.4 ~1+0.8
13 13 0.04-6 0.8 1.6 £0.4 11 ~0.1-2 0.2 - 0.5 0.1

9Gross includes capsule plus nozzle isotopes.

bNet includes capsule isotopes only, less proportional quantity of material of nozzle composition, for the given sample.

“This is the percentage of MSRE production of the chain that is present, as the noble-gas precursor of the indicated nuclide, in the
average gas in (and leaving) the pump bowl, if complete stripping occurs in the pump bowl. Daughters of the noble gas resulting from

its decay while in the pump bowl are not included.

possible exception at 3%) are to be found as actually
gas-borne, and little or no tellurium and iodine. The
quantities,of *°Sr (3%) and !*7Cs (23%) are undoubt-
edly real and do indicate gas-borne material. Com-
parison with the values indicated by calculations assum-
ing complete stripping of noble gases on passage of salt
into the pump bowl (14% for 3°Sr, 18% for '37Cs,
0.07% for °'Y, and 0.16% for '*°Ba) show that
observed values in the gas phase are below fully stripped
values by moderate amounts except in the case of
137Cs. The low values could result from some addit-
ional holdup of the gases in the sampler spiral, and high
1375 values could result from the greater volatility of
cesium, which might permit this, as a product of decay
in the pump bowl, to remain uncondensed. Though we
doubt that these arguments could stretch enough for
the data to fit perfectly, the magnitudes are right, and
we conclude that the net values for *°Sr and '*7Csin
the gas are real and are reasonably correct.

“The gas samples thereby indicate that, except for
nuclides having noble-gas precursors, only small frac-
tions of any fission product chain should be carried out
of the pump bowl with the off-gas, with mist account-
ing for the major part of the activity in samples. The

- amount of salt carried out with off-gas as mist has been

estimated® as “at most a few grams a month.” Far
lower mist concentrations than appeared in our sam-
ples, which were taken within the sampler shield, are
indicated for the off-gas. We conclude that the “net”
median column, which discounts the mist, is the best
measure furnished by our gas samples of the fraction of
the various chains leaving the system with the off-gas.

REFERENCES

1. S. 8. Kirslis and F. F. Blankenship, “Pump Bowl
Volatilization and Plating Tests,”” MSR Program Semi-
annu. Progr. Rep. Aug. 31, 1966, ORNL-4037, pp.
169-71.
52

2. S. S. Kirslis and F. F. Blankenship, “Freeze Valve 3. J. R. Engel, P. N. Haubenreich, and A. Houtzeel,

Capsule Experiments,” MSR Program Semiannu. Progr. Spray, Mist, Bubbles, and Foam in the Molten Salt
Rep. Feb. 28, 1967, ORNL4119, pp. 139-41. Reactor Experiment, ORNL-TM-3027 (June 1970).

53

Table 8.2, Gas samples,23 U operation

For the fission products, each entry in the table consists of three numbers. The first number is the observed activity of the isotope in disintegrations per minute per cubic centimeter of capsule
volume, corrected to time of sampling; the second number is the ratio of the activity to the production in disintegrations per minute per cubic centimeter of purge; the third number is the ratio
to the activity in 1 mg of inventory salt. For U-235, the first number is the observed amount in micrograms per square centimeter of capsule surface; the second number is the ratio of this amount to
the amount in 1 mg of inventory salt (14 ug).

Sr-89 Ba-140 - U-235% Ce-141  Ce-144 Z1-95 Nb-95 Mo-99  Ag-111  Ru-103  Ru-106 Te-132  Te-129m 1-131

S*:]‘:)ple Date I&‘;‘;‘; H;;ifie iys 479 6.5l 6.3 5.6 6.2 62 6.06  0.019 3.0 0.38 4.24 0.133 3.1
' ; 504 12.8 33 285 65 35 275 76 40 367 321 37 8.05
10-11  12-27-66 8 14E7 020 <1.5E5 <17E6 1.0E10 19E8  3.4E6 2.9E9 4.9E8
0.0031 <0.0002 <0.006  0.52 0.29 0.37 0.23 0.15
1.3 0.014 <0026 <010 55 6.0 26 24 75
10-22°  1-1167 8 1.8E7  0.028 <0.1E6  1.1E7  7.0E9 13E8  3.9E6 2.6E9 1.0E8
0.0039 <0.0013  0.037  0.36 0.20 0.43 0.21 0.029
11 0.002 <0015 036 36 2.8 2.5 19 1.2
11429 41167 0 1.0E7 3.0 <22E6  33E7  5.5E9 13E8  4.1E6 6.0E9 2.9E8
(87) 011  0.26 0.19 0.45 0.49 0.085
0.10 0.21 <0015 039 30 1.7 1.5 47 3.1
1146 41867 8 20E8 3.1E7 046  _, ~1E6 65E7  12E10 23E8  4.8E6 1.7E10  4.0E7 4.9E7
025  0.0070 00012 023  0.60 0.35 0.52 135 1.27 0.015
18 16 0.03 ~0008 075 60 3.1 1.7 125 12 0.55
11-53¢ 5267 8 1.8E8 1.2 9.E6 SE8 8EF9  28E7 55E8  2.0E7 1.0E10  1.8E8 4.3E8
0.22 ' 0011 175 041 126 083 2.2 078 5.6 0.13
16 0.18 0065 50 43 48 70 6.5 70 50 45
12277 62167 0 4.1T7 24E7  2.1E6 43E6  1.2E8 Q1E6¥F 1.1E7
0.014 0004 0.005 040 (0.0002) 0.35
, 0.36 0.15 0.035 0030 12 ©.020) 32
1226" 71767 8 1.8E8 1.3 18E8  15E6 14E10 65E6 20ES  8.5E6 1.6E9  3.3E7 8.5E8
0.22 0.21 0.005 070 030 030 0.94 0.13  1.05 0.25
2.3 0.09 1.7 0025 175 14 40 27 13 15 1055
1467 3668 5 8.5E7 20E7 14 9.0E6  8.5E6 1.1E7  1.1E8 16E10 6.0ES  2.5E7 6.0E9 1.6E9
0.16  0.0072 0.009  0.024 0020 039 128 145  _4.4 078 0.74
0.850 0.160  0.10 0.021  0.110 0.85 10 78 9.50 55 125 28

?Corrected to time of sampling or to prior shutdown where nécessary.

blnventory: 14 pg of U-235 per milligram of salt.
€ After addition of 5.6 g of beryllium.

4 pfter addition of 8.4 g of beryllium.
®Helium bubbles.

fAfter 42 days down.

£ Approximate.

PR - e
2994

54

Table 8.3. Data for gas samples from MSRE pump bowl during uranium-233 operation

Sample number FP15-29 FP1543 FP15-52 FP15-58 FP15-71 FP17-6- FP17-17 FP17-25
. Capsule volume, cc 30.00 30.00 30.00 30.00 30.00 30.00 30.00 30.00
Date 10-13-68 10-29-68 11-6-68 11-12-68 11-27-68 1-22-69 2-10-69 3-14-69
Megawatt-hours 0.1 0.1 0.1 0.1 1.8 647.0 3408.0 7284.0
Power, MW 0.00 0.00 0.00 0.00 0.03 4.60 4.60 7.20
Rpm 1180 1180 1180 1180 1180 1180 1180 942
Pump bowi level, % 62.00 69.50 66.00 62.50 63.20 60.00 55.30 54.60
Overflow rate, Ib/hr 1.4 74 1.0 0.8 3.7 2.2 1.6 1.0
Voids, % 0.60 0.00 0.00 0.00 0.60 0.60 0.60 0.00
Flow rate of gas, std liters/min 3.30 He 3.30 He 3.30 He 3.30 He 3.30 He 3.30 He 3.30 He 3.30 He
Sample line purge On On On On On - On On . On
Fission product isotopes®
Half-life Fission yield
Isotope (days) (%)
Sr-89 52.00 5.46 1.01E7 4.70E5 4.10E4 5.67E4 3.29E6 1.44E8 4.63E7
1531 88.679 8.613 12.908 870 13,846 638
Sr90 10264.00 5.86 ‘4.87E6 2.47E6
1196 609
Y91 58.80 5.57 2.24E8
22,626
Ba-140 12.80 5.40 6.83E8 1.18E6
23,727 11,346
Cs-137 10958.00 6.58 2.21E7 3.63E7 1.34E4 8.07ES5 2.67E7 9.17E7 8.50E8
5430 8927 3.284 198 6576 22,358 185,590
Ce-141 33.00 7.09 1.57E8 5.37E4
9345 0.475
Ce-144 284.00 4.61 8.63E7 5.50E5 1.16E4 6.93E3 3.19E7 7.47E8 7.04E5
’ 1784 11.828 0.254 0.154 736 19,096 15.678
Nd-147 11.10 1.98 : 2.28E7
_ 544
Zr-95 65.00 6.05 2.20E7 8.43E4 2.93E3 1.48E3 5.83E6 3.14E8 1.23E8
1528 7.028 0.264 0.143 647 25,923 3138
Nb-95 35.00 6.05 1.23E8 1.51E7 1.68E6 1.99E6 4.67E8 1.18E8 6.63E8 6.10E7
_ 14,171 1524 165 195 46,205 14,713 48,775 1848
Mo-99 2.79 4.80 7.57E8 2.38E9 3.50E9
315,278 18,740 41,766
Ru-103  39.60 1.99 3.50E7 1.07E8 1.35E8 8.67E7
35,611 25,660 7519 2992
Ru-106  367.00 0.43 1.99E7 5.10E6 1.98E6 3.32E7 4 87E6 7.60E6 5.13Ee6
. 3471 918 362 6303 1010 1504 968
Ag-111 7.50 0.02 1.21E7
' 24,346
Sb-125 986.00 0.08 8.03E5 1.70ES5
7109 1518
Te-129m 34.00 0.33 6.47E7 1.24E8 9.97E7
87,634 37,268 19,244
Te-132 3.25 4.40 3.24E8 5.87E9 8.57E9
’ 169,808 49,718 109,408
I-131 8.05 2.90 1.81E9 1.36E8 4.73E8 1.02E9
3,389,619 6003_ 6940 16,860
Salt constituents?
Constituent
U-233 0.0046 0.1183 0.0231 0.0010
682 17,704 3461 149
Li 2.9667
25,685
Be 0.2000
55

Table 8.3. (continued) Data for gas samples from MSRE pump bowl during uranium-233 operation

Sample number FP17-33 FP18-14 FP18-15 FP18-21 .FP18-25 FP18-29 FP1842
Capsule volume, cc 30.00 7.80 30.00 7.80 7.80 7.80 7.80
Date 4-4-69 5-6-69 5669 5-12-69 5-17-69 5-21-69 5-28-69
Megawatt-hours 10695.0 16009.0 16052.0 17163.0 18002.0 18665.0 19679.0
Power, MW 8.00 7.80 7.70 7.90 7.00 7.80 6.60
Rpm 1180 1180 1180 1180 990 1180 990
Pump bowl level, % 60.00 63.40 60.10 60.50 55.90 60.10 53.20
Overflow rate, Ib/hr 4.8 4.6 2.2 1.2 0.1 1.3 0.0
Voids, % 0.60 0.60 0.60 0.60 0.00 0.60 0.00
Flow rate of gas, std liters/min 3.30 He 3.30 He 3.30 He 3.30 He 3.30 He 3.30 He 2.30 He
Sample line purge On On On On On On On

Fission product isotopes?

Half-life Fission yield

Isotope (days) (%)
Sr-89 52.00 546 5.70E8 2.24E8 2.33E7 1.54E7 3.23E8 4.35E8
590 1918 199 124 2485 3293
Y-91 58.80 5.57 1.85E4 2.56E6 1.03E6 3.88E6 4.12E6
0.217 24 .420 9.371 33.488 34.583
Ba-140 12.80 5.40 B.30E6 1.591:7 3.03E6 3.09E6 2.33E7 2.05E6
57.639 102 19423 18.840 142 13.234
Cs-137 10958.00 6.58 1.19E7 4.99k6 2.42E6 1.06E6 4.06L6 3.00E7
2457 . 974 471 203 762 5556
Ce-141 33.00 7.09 5.57ES8 9.6015 3.67LE5 1.16E6 2.59E6
3.787 5.649 2.144 6.272 13.999
Cc-144 284.00 4.6l 5.33LE58 5.7615 1.09ES 5.24L5 71315 1.58E6
10.336 10,261 1.934 9.103 12.000 26.108
Z1-95 65.00 6.05 1.8615 49015 8.29i1:5 6.74L5
1.721 4495 6.855 5483
Nb-95 35.00 6.05 2.051°8 1.551:7 1.9518 2.06E7 1.991:7 44717
: 4218 226 2834 281 246 519
Mo-99 2.79 4.80 5.63E8 1.14E9 5.60E9 1.16E9 1.51E9 2.10E9
) 3832 8748 42,424 7407 10,085 17.818
Ru-103  39.60 1.99 1.37E8 1.08E8 1.73E8 6.86E7 - 4.47E7 8.73E7
3604 2402 3847 1453 912 1771
Ru-106 367.00 0.43 8.43E6 4.92E6 8.83E6 2.95E6 1.95E6 4.15E6
1509 838 1503 493 320 674
Ag-111 7.50 0.02 1.84E7 8.27E6 . 2.01E7 1.81E6 4.28E6 4.27E6
27,170 11,630 28,191 2382 5763 6334
Sb-125 986.00 0.08 1.58ES
628
Te-129m 34.00 0.33 3.70E8 8.90E6 1.27E8 2.32E7 9.22E6 4.717E7
54,917 1135 16,152 2813 1081 5594
Te-132 3.25 4.40 1.63E]10 3.35E11 6.63E9 1.91E9 4.32E8 1.53E9
122,556 2,720,450 53,495 13,358 3154 13,745
[-131 8.05 2.90 2.55E8 2.23E8 8.77E8 1.97E8 3.99E7 1.67E8 1.55E8
3148 2600 10,182 2162 446 1858 1894
Salt constituents?
Constituent
U-233 0.0006 0.0001 0.0002 0.0001 0.0001 0.0006
95.402 12.276 34.560 17.646 17.646 87.849
Li 1.1853 0.0067 1.6333 0.0042 0.3077 0.0962

10,262 57.720 14,141 36.630 2664 833
56

Table 8.3 (continued) Data for gas samples from MS.RE pump bowl during uranium-233 operation

Sample number
Capsule volume, cc
Date
Megawatt-hours
Power, MW

Rpm

Pump bowl level, %
Overflow rate, Ib/hr
Voids, %

Flow rate of gas, std liters/min
Sample line purge

Half-life Fission yield

Isotope (days) (%)

St-89 52.00 5.46
Y91 58.80 5.57
Ba-140 12.80 5.40

Cs-137 10958.00 6.58

Ce-141 33.00 7.09
Ce-144  284.00 4.61
Nd-147 11.10 1098
Zr-95 65.00 6.05
Nb-95 35.00 6.05
Mo-99 2.79 4.80
Ru-103  39.60 1.99
Ru-106  367.00 0.43
Ag-111 750 0.02
Sb-125  986.00 0.08
Te-129m 34.00 0.33
Te-132  3.25 4.40
k131 8.05 2.90
Constituents

U-233

U-total

Li

Be .

Zr

FP19-13
7.80
8-21-69
20310.0

- 001

1165
63.25
0.1
0.38
3.30 He
On

5.49E6
128

5.63E5
12.676

7.47ES
139
2.12ES
6.631
6.47E5
13.213
1.16E4
33.378
4.33ES
8.615
2.22E7
298

1.20E7
1053
3.22E6
613

1.92E6
1238
1.05E6
13462
1.62E5
1650

0.0012
182

0.2218
27,484
0.0015
13.320

FP19-14
15.00
8-21-69
20310.0
0.01
1165
64.10
0.1

0.38
3.30 He
On

5.65E6
132

8.93E5
20.166
9.40E4
52.222
3.50E6
652

1.37E6
43.187
4.59E6
93.605

3.09E6
61.487
4.27E7
575
6.19E5
6472
1.26E7
1107
3.60E6
688

2.57E6
1662
2.20E6
27329
2.07ES
2113

0.0004
61.041
0.0011
132
0.0108
93.506
00114
171
0.5253
4540

FP19-15
7.80

3.30 He
On

FP19-16
15.00
8-21-69
20310.0
0.01

3.30 He
On

Fission product isotopes®

2.68E7
628
3.12E6
70.644
1.55E5
87.151
2.00E6
372
1.63E6
51.525
3.49E6
71.167
4.21E4
124
2.27E6
45.385
2.42E8
3270
1.82E6
18482
3.28E7
2910
9.65E6
1845

7.05E6
26311

7.92E6
5160

4 49E6
53997

4.83ES5
4957

7.87E6
185
4 48ES
10.159
3.35E5
189
1.63E6
304
3.48E6
110
8.13E6
166
1.20ES
356
5.13E6
103
3.18E7
429
8.40Es5
8317
5.96Ee6
529
1.47E6
280

8.07E4
301
3.07E6
2008
3.70E6
43478
4.98ES
5108

Salt constituents?

0.0007
98.398

0.0064
55.500
0.0011
16.889

0.0011
166
0.0010
124
0.0147
127
0.0096
144

FP19-19
7.80
9469
21557.0
5.50
1100
62.00
1.5
0.50
2.40 Ar
Off

2.94E7
569
1.27E6
25.158
1.58E6
33.409
3.04E7
5565
1.45ES
2.739
5.22E5
10457
3.35ES
17.799
4.64ES5
8.171
1.33E7
192
1.64E8
2075
9.74E6
616
8.21ES5
155
7.46ES
2637
2.04E4
73.063
291E6
1172
1.06E8
1518
3.59E7
1088

0.0002
25.894
0.0008
95.320

0.0009
13.435

FP19-20
15.00
94-69
21587.0
5.50

240 Ar
Off

2.28E6
43.931
6.61ES
13.044
4.99E5
10.346
3.65E6
668
1.19ES
2.222
6.12E5
12.265
1.39E5
7.222
3.43E5
6.012
1.73E6
24916
1.16E8
1432
4.87E6
304
6.67ES
126
1.76ES
609

1.41E6
563
9.93E7
1385
5.89E7
1749

0.0001
13.166
0.0006
74.349

0.0018
26.946

FP19-23
15.00
9-10-69
22255.0
0.01

2.40 Ar
Off

4.65E6
81.378
343E6
61.751
8.20E6
126
1.37E7
2480
3.00E6
46.802
4.59E6
90.646
1.19E6
46.615
2.09E6
34.261
8.53E6
125
1.68E8
1900
5.95E7
3221
5.44E6
1022
1.99E5
3537

2.29E6
766
6.31E7
782
1.01E8
2314

0.0005
71.215
0.0013
165

0.0035
30.014
0.0025
36.926

FP19-28
15.00
9-23-69
23437.0

2.40 He
Off

9.07E7
1430
7.67ES
12.757
3.28E6
41.677
1.01E6
180
2.16ES
2.809
2.36ES
4.600
2.11ES
6.975
1.38ES
2.088
4.07E7
608
2.08E9
22584
6.87E7
3218
3.29E6
613
247E6
6092

1 .49E7
4183
6.65E8
8018
1.21E8
2498

0.0000
6.284

0.0030
25.974
0.0012
17.964
57

Table 8.3. (continued) Data for gas samples from MSRE pump bowl during uranium-233 operation

Sample number FP19-29 FP19-37 FP19-38 FP1941 FP1946 FP19-54 FP19-56 FP19-62
Capsule volume, cc¢ 7.80 15.00 15.00 15.00 15.00 15.00 15.00 15.00
Date 9-23-69 9-30-69 10-1-69 10-3-69 10-7-69 10-14-69 10-15-69 10-22-69
Megawatt-hours 23458.0 23936.0 24025.0 243520 251530 26601.0 26821.0 28073.0
Power, MW 5.50 0.01 5.50 7.00 8.00 8.00 0.01 8.00
Rpm 1188 611 610 1175 1176 1185 1185 1186
Pump bowl level, % 64.00 64.60 60.80 64.00 63.80 67.00 66.00 63.60
Overflow rate. Ib/ht 39 0.5 09 4.3 5.8 8.3 7.0 39
Voids, % 0.53 0.00 0.00 0.53 0.53 0.53 0.53 0.53
Flow rate of gas, std liters/min 2.40 He 2.45 He 2.40 He 2.40 He 3.30 He 3.30 He 3.35 He 3.30 He
Sample line purge Off Off Off Off Off - Off Off Off

Fission product isotopes?

Half-life Fission yield

Isotope (days) (%)
Sr-89 52.00 5.46 7.23E6 " 1.67E6 2.40E7 6.29E6 1.51E8 2.69E6 1.11Eé6 2.56E7
114 119 369 92.451 2007 30.537 12.310 259
Y91 58.80 5.57 8.71ES5 4.68ES5 3.65E4 3.55E5 4.67E6 1.65E5 2.99ES 4.61E5
14.436 " 7.635 0.591 5.568 66.858 2.038 3.628 5.107
Ba-140 12.80 5.40 1.83E6 5.19E5 6.58E5 1.21E6 1.33E7 1.64E5 5.69ES5 191E6
23.177 6.745 8.361 13.870 123 - 1.188 4.067 12.544
Cs-137 10958.00 6.58 7.13E5 6.03E6 3.04E5 1.69E6 2.03E6 5.59E5 4.35E5 4.73E6
128 1072 53.996 299 354 95.991 74.429 798
Ce-141 33.00 7.09 5.29ES5 1.90ES 291E3 1.85ES 3.53E6 1.19ES 3.81ES 2.55ES
6.859 2.402 0.036 2.165 35.582 0973 3.045 1.837
Ce-144 284.00 461 9.36E5 9.07ES 1.31E4 2.64E5 2.25E6 2.79E5 2.47E5 1.54E5
18.244 17.639 0.254 5.077 42.310 5.057 4.465 2.702
Nd-147 11.10 1.98 2.85E5 8.00E4 9.27E4 1.41E6 7.00E4 2.35ES5 1.93E5
9.362 2.749 2.791 33.733 1.318 4.346 3.328
2195 65.00 6.05 3.63ES 1.63E6 2.51F4 347ES5 1.91E6 . 1.37Es 1.69E5 9.07E4
5.481 24.315 0.372 4976 25.175 1.575 1.910 0.943
Nd-95 35.00 6.05 5.95E6 4.13E7 4.17E5 3.29E7 247E7 2.83E6 3.74E6 340E6
88.787 619 6.238 491 367 40.848 53.736 46.897
Mo-99 2.79 4.80 3.17E8 5.88E8 2.19E7 6.93E8 9.53E8 2.81E7 1.81E8 1.51E8
3409 7332 257 6420 6356 162 . 1063 928
Ru-103  39.60 1.99 1.77E7 3.95E7 2.97E6 8.07E6 5.79E7 5.56E6 7.13E6 6.87E6
825 1803 134 344 2165 173 217 188
Ru-106  367.00 0.43 1.15E6 3.55E6 1.97Es 245E6 3.25E6 1.15E6 4.35E5 - 4.08ES5
213 660 36.589 453 592 204 77.017 70.941
Ag-111 7.50 0.02 7.51E5 7.20E5 3.22E5 7.13E5 5.01E6 8.47E4 1.02ES 1.57E5
1841 - 1930 836 1621 - 8749 116 139 205
Te-129M  34.00 0.33 1.05E7 1.26E7 1.01E6 1.24E7 463E7 9.07E5 2.77E6 3.63E6
2922 3436 272 3128 10109 161 483 566
Te-132 3.25 4.40 5.51E8 1.47E8 2.71E7 4 .46E8 2.48E9 3.61E6 8.53E7 1.77E8
6594 2026 355 4660 18647 23.015 554 1190
I-131 8.05 2.90 1.56E8 2.77E8 2.33E7 7.60E7 1.75E8 1.21E6 8.93Eé6 1.77E7
3218 6190 506 1450 2580 14.092 03 194
Salt constituents?
Constituent
U-233 0.0002 0.0002 0.0000 0.0001 0.0004 0.0000 0.0000 0.0000
28.388 35.907 2.693 10.074 54 857 6.882 4987 5.187
U-total 0.0009 0.0039
116 484
Li 0.0005 ' . 0.0046 0.0079
4440 39.758 68.687.
Be 0.0010 0.0008 1 0.0017 0.0014 0.0017 0.0006 0.0028 0.0011

15.354 11.976 25948 20958 25948 8.982 41916 15.968
Sample number
Capsule volume, cc
Date
Megawatt-hours
Power, MW

Rpm -
Pump bowl level, %
Overflow rate, Ib/hr
Voids, %

Flow rate of gas, std liters/min

Sample line purge

Half-life Fission yield

Isotope (days)

Sr-89 52.00
Y91 58.80
Ba-140 12.80
Cs-137 10958.00
Ce-141 33.00
Ce-144 284.00
Nd-147  11.10
Zr95 65.00
Nb95 35.00
Mo-99 2.79
Ru-103  39.60
Ru-106  367.00
Ag-111 7.50
Te-129m  34.00
Te-132 3.25
1-131 8.05
Constituent
U-233

U-total

Li

Be

58

Table 8.3 (continued) Data for gas samples from MSRE pump bowl during uranium-233 operation.

(%)
5.46

5.57
5.40

6.58

4.40

290

FP19-64
7.80
10-22-69
28182.0
8.00
1188
63.00
3.7

0.53
3.30 He
Off

4.27E6
42.821
1.92E5
2.116
1.38E6
9.050
2.82E6
475
2.71ES
1.919
1.6415
2.874
1.90ES
3.243
2.2415
2.318
1.67E6
22,925
8.46E7
516
2.96E6
80.203
2.00ES
34,775
3.46ES
451
1.28E7
1984
7.65E8
5103
6.23E7
679

0.0001
10.358

0.0154
133

0.0023
34.546

FP19-65 FP19-70 FP19-73
15.00 7.80 7.80

10-23-69 10-28-69 10-29-69
28299.0 29219.0 29450.0

8.00 8.00 8.00
1185 1188 1185
63.40 63.50 63.60
4.0 53 3.2
0.53 0.53 0.53
3.25 He 3.30 He 3.25 He

Ooff Off Off

Fission product isotopes’

4.86E7 8.41E8 1.03E7
481 7860 95.560
6.80ES 1.35E9 5.62ES
7.424 13849 5.701
4.86E6 1.73E9 3.24E6
31.558 10684 19.899
1.53E6 1.08E8 1.21E6
258 17940 200
1.3714 1.17E9 3.73ES
0.097 7675 2423
1.47E4 6.2218 2.90ES
0.256 10629 4928
2.15E8 1.861:6
3514 30.080
1.811:4 8.28L8 33108
0.186 8041 3.180
6.5TES 1.37K9 47116
9.005 18169 61.747
4.12E7 3.31E9 4.10E8
251 20047 2486
1.15E6 1.26E8 1.63E7
30.956 3160 402
4.17E5 6.15E6 1.01E6
72,223 1054 172
6.67E4 4.45E6 9.42ES
86.468 5617 1184
4.89E6 8.73E7 1.56E7
746 12450 2202
2.43E8 2.09E9 8.27E8
1622 13839 5476
2.10E7 7.60E07 8.13E7
228 801 852

Salt constituents?

0.0003
42.198
0.0667 0.1890
8261 23417
0.0029
25.530
0.0005
1.677

FP19-77
15.00
10-31-69

FP19-78
15.00
10-31-69

1.02E6
9.027
5.09E04
0.494
947E4
0570
2.33E5
38.205
6.18E4
0.384
6.43L4
1.076

6.11014
0.560
1.47L6

18,613
3.41E7
220
3.66E6
86.664
2.74E5
46.202
1.29E5
163
2.12E6
286
3.94E7
276
361E7
382

0.0002
26.731

0.0034
29.437
0.0013
19.960
59

Table 8.3 (continued) Data for gas samples from MSRE pump bowl during uranium-233 operation

Sample number : : FP20-9 FP20-12 FP20-27 FP20-32
Capsule volume, cc 13.80 13.80 13.80 15.00
Date 12-1-69 12-2-69 12-10-69 12-12-69
Megawatt-hours 31212.0 31429.0 329250 33297.0
Power, MW 8.00 = 8.00 8.00 0.00
Rpm 1188 1188 1190 1180
Pump bowl level, % 62.40 59.80 61.50 0.00
Overflow rate, Ib/hr 4.6 1.3 2.0 0.0
Voids, % 0.53 0.53 0.50 0.00
Flow rate of gas, std liters/min 3.30 He 3.30 He 3.30 He 3.30 He
Sample line purge Off Off oft Off

iy .
Half-life Fission yield Fission product isotopes”

Isotope (days) (%)
Sr-89 52.00 5.46 1.41E8 6.43E7 7.54E7 1.41E6
1674 741 173 14.067
Y-91 58.80 5.57 5.34ES 2.26E7 3.55E5
6.676 251 3.843
Ba-140 12.80 5.40 6.96E6 2.65E6 4.33E6 4.65ES5
90.816 31.915 36.662 3.717
Cs-137 10958.00 6.58 7.68E5 5.39ES 4.35E5 9.33ES
126 88.382 70.013 150
Ce-141 33.00 7.09 4,75E4 8.48E4 2.55E5 3.29ES
0.452 0.785 1.977 2471
Ce-144 284.00 4.61 2.71ES 3.43E5 2.84E5 2.49E5
4.831 6.088 4.872 4.236
Zr-95 65.00 6.05 1.09ES 3.05E5 4.78ES 3.16E5
1.274 3.515 4.949 3.202
Nb-95 35.00 6.05 1.19E6 5.83E6 6.73E6 4 87E6
14,781 72.374 81.996 58.928
Mo-99 2.79 4.80 1.22E8 1.57E8 591E7
942 978 374
Ru-103 39.60 1.99 6.40ES 5.49E6 6.29E6 3.99E6
21.923 183 180 111
Ru-106 367.00 0.43 8.62ES 3.86ES 3.21E$ 2.63E5
153 68.219 55.300 45085
Ag-111 7.50 0.02 2.49E6 6.12ES5 1.79E4
6122 986 27487
Te-129M 34,00 0.33 2.98ES 5.61ES 2.15E7 8.33E5
60.729 111 3612 136
Te-132 3.25 4.40 1.51E7 5.64E8 8.93E6
136 3948 62471
I-131 8.05 2.90 4.25E5 2.74E6 2.50E7 1.69E7
Salt constituents?
Constituent
U-233 0.0000 0.0000 0.0001 0.0000
6.071 5.421 7.589 6.982
U-total 0.0009 0.0307
_ 117 3800
Li 0.0007 0.0043
6.274 37.518
Be 0.0008
11.933

2Each entry for the fission product isotopes consists of two numbers. The first number is the radioactivity of the isotope in the entire capsule (in
disintegrations per minute) divided by the capsule volume (in cubic centimeters). The second number is the ratio of the isotope to the amount
calculated for 1 ug of inventory salt at time of sampling.

bEach entry for the salt constituents consists of two numbers. The first number is the amount of the constituent in the capsule (in micrograms)
divided by the capsulevolume (in cubic centimeters). The second number is the ratio to the amount calculated for 1 ug of inventory salt at time of
sampling.
60

9. SURVEILLANCE SPECIMENS

9.1 Assemblies 1-4

9.1.1 Preface. Considerable information about the
interaction of fission products generated in fissioning
fuel salt and the surfaces of materials of construction
such as were used in MSRE was obtained from an array
of surveillance specimens which was inserted in a
central graphite bar position in the MSRE core, being
removed periodically to obtain certain specimens for
examination and replace them with others. A control
specimen rig' was also prepared in order to subject
materials to fluoride salts with essentially the same
temperature-time profile and temperature and pressure
fluctuations as the reactor in the absence of radiation;
it, of course, was not a source of fission product data.

A photograph of typical graphite shapes used in a
stringer is shown in Fig. 9.1, their assembly into
stringers in Fig. 9.2, and the containment of stringers in
a perforated container basket in Fig. 9.3.

A B

Assemblies of this design were used in exposures
during runs 1 to 18; during runs 19 and 20 a different
design, described later, was used.

The graphite pieces were generally rectangular slabs
(with notched ends) arranged longitudinally along a
stringer. The bars were 5 to 9 in. long, 0.66 in. wide,
and 0.47 in. thick and were generally fabricated from
pieces of MSRE graphite (CGB) selected to be crack
free by radiographic examination. Bars of half this
thickness were also employed. The bars were assembled
into long stringers by clamping together with a pair of
Hastelloy tensile specimen assemblies and an associated
flux monitor tube. Three such stringers were clamped
together as shown in Fig. 9.4 and placed within the
perforated 2-in. cylindrical container basket (0.03-in.
wall). The basket was inserted in a 2.6-in.-diam channel
occupying a central bar position in the MSRE core. This
central region, with no lattice bars below it, had flows
around the basket that were in the low turbulent range;

Y-64822

C

Fig. 9.1. Typical graphite shapes used in a stringer of surveillance specimens.

2

61

PHOTO 80761

L
S
T R
Fig. 9.2. Surveillance specimen stringer.
PHOTO BI671
SPACER AND GUIDE PIN
UPPER GUIDE BASKET LOCK.\ ASSEMBLY

INOR-8 ROD OF TENSILE SPECIMENS
GRAPHITE (CGB) SPECIMENS
BINDING STRAP

BASKET

PIN
FLUX MONITORS TUBE

(a) BASKET. BALL LOCK ASSEMBLY

~CONTROL ROD

GUIDE TUBE
GUIDE BAR

TYPICAL

FUEL CHANNEL

() SURVEILLANCE SPECIMEN 2 in. TYPICAL

R=REMOVABLE STRINGER

Fig. 9.3. Stringer containment.

ORNL_ - DWG 65-4184

———GRAPHITE SPECIMEN

INOR-8 SPECIMEN

—~FLUX MONITOR

Fig. 9.4. Stringer assembly.

within the basket along the specimens, no detailed flow
analysis has been presented, but quite likely the flow
may have been barely turbulent, because of entrance
effects which persisted or recurred along the flow
channel.

The first two times that the surveillance assemblies
were removed from the MSRE (after runs 7 and 11) for
fission product examinations, metal samples were ob-
tained by cutting the perforated cylindrical Hastelloy N
basket that held the assembly. The next two times
(after rums 14 and 18), '%-in. tubing that had held
dosimeter wires was cut up to provide samples. When
the tubing was first used, lower values were obtained
(after run 11) for the deposition of fission products.
After run 18 the basket was no longer of use and was
cut up to see if differences in deposition were due to
differences in type of sample.

The distribution of temperatures and of neutron
fluxes along the graphite sample assembly were esti-
mated? for 223U operation (Fig. 9.5). The temperature
of the graphite was generally 8 to 10°F greater than
that of the adjacent fuel, normally which entered the
channel at about 1180°F and emerged at 1210°F. The
thermal-neutron flux at the center was about 4.5 X
10'3 with a fast flux of 11 X10'3; these values
declined to about one-third or one-fourth of the peak at
the ends of the rig.

It was found on removal of the assembly after run 7
that mechanical distortion of the stringer bundle with
specimen breakage had occurred, apparently because
tolerance to thermal expansion had been reduced by
salt frozen between ends of consecutive graphite speci-
mens. The entire assembly was thereby replaced, with
slight modifications to design; this design was used
without further difficulty until the end of run 18.

After the termination of a particular period of reactor
operation, draining of fuel salt, and circulation and
draining of flush salt (except after run 18, when no
flush was used), the core access port at the top of the
reactor vessel was opened, and the cylindrical basket
containing the stringer assemblies was removed, placed
in a sealed shielded carrier, and transported to the

segmenting cell of the High Radiation Level Examina-

tion facilities. Here the stringer assembly was removed,
and one or more stringers were disassembled, being
replaced by a fresh stringer assembly. Graphite bars
from different regions of the stringer were marked on
one face and set aside for fission product examination.

Stringers replaced after runs 11 and 14 contained
graphite bars made from modified and experimental
grades of graphite in addition to that obtained directly
from MSRE core bars (type CGB). After runs 7 and 11,
'/, 6-in. rings of the cylindrical 2-in. containment basket
were cut from top, middle, and bottom regions for
fission product analysis. Samples of the perforated
Hastelloy N rings were weighed and dissolved. A similar
approach was used after run 18.

After runs 14 and 18, seven sections of the !%-in.-
diam (0.020-in.-wall) Hastelloy N tube containing the
flux monitor wire, which was attached along one
stringer, were obtained, extending from the top to the
bottom of the core. These were dissolved for fission
product analysis. This tube was necessarily subjected to
about the same flow conditions as the adjacent graphite
specimens. The flow was doubtless less turbulent than
that existing on the outside of the cylindrical contain-
ment basket.

For the graphite specimens removed at the end of
runs 7, 11, 14, and 18, the bars were first sectioned
transversely with a thin carborundum saw to provide
specimens for photographic, metallographic, autoradio-
graphic, x-radiographic, and surface x-ray examination.
The remainders of the bars — 7 in. long for the middle
specimen, 2%, to 3 in. for the end specimens — were
used for milling off successive surface layers for fission
product deposition studies.

A “planer” was constructed by the Hot Cell Opera-
tions group for milling thin layers from the four long
surfaces of each of the graphite bars. The cutter and
collection system were so designed that the major part
of the graphite dust removed was collected. By com-

63

paring the collected weights of samples with the initial
and final dinensions of the bars and their known
densities, sampling losses of 18.5%, 4.5%, and 9.1%
were indicated for the top, middle, and bottom bars of
the first specimen array so examined.

The pattern of sampling graphite layers shown in Fig.
9.6 was designed to minimize cross contamination
between cuts. The identifying groove was made on the
graphite face pressed against graphite from another
stringer in the bundle and not exposed to flowing salt.
After each cut the surfaces were vacuumed to minimize
cross contamination between samples. The powdered
samples were placed in capped plastic vials and weighed.
The depth of cut mostly was obtained from sample
weights, though checked (satisfactorily) on the middle
bar by micrometer measurements. In this way it was
possible to obtain both a *“‘profile” of the activity of a
nuclide at various depths within the graphite and also,
by appropriate summation, to determine the total
deposit activity related to one square centimeter of
(superficial) graphite sample surface. The profiles and
the total deposit intensity values, though originating in
the same measurements, are most conveniently dis-
cussed separately.

9.1.2 Relative deposit intensity. In order to compare
the intensity of fission product deposition under
various circumstances, we generally have first obtained

ORNL- DWG 65-12049

(x10'3) '
“IMN FAST FLUX

10 1220
T \ GRAPHITE TEMPERATURE
2 \ L _{APPROX)
"] o — —

- |
Y g < 1210
g / L | ——=—FUEL
J P TEMPERATURE _
P . V4 - w
=] 1 o
s e 47 L } 1200 w
=~ =
= // ~ =
3 ” e i)
g 4 P R 190 %
=z b ' ~ w
o e e -
@ ”~ e
s |- -
e L / //
Z 2 THERMAL FLUX A #80
-
-
0 H70
o) 10 20 30 40 50 60 70 80

DISTANCE ABOVE BOTTOM OF HORIZONTAL GRAPHITE GRID (in.)

Fig. 9.5. Neutron flux and temperature profiles for core surveillance assembly.
64

ORNL-DWG 66-11378

— 27
32
54
61
63
QM P8 TOP GRAPHITE
G2
&0
58
29
25
/IDENTIFYING GROOVE
| 50
mlo|o|0[L| = 01O Naoloey
c
o MIDDLE GRAPHITE
5
(@]
73
17
14
11
J
4
1
t———————————— (0660 in, =—|
36
40
65
67
68
. olo| BOTTOM GRAPHITE
<t {0
&9
66
64
38
34

Fig. 9.6. Scheme for milling graphite samples.
the observed activity of the nuclide in question per unit
of specimen surface (observed disintegrations per
minute per square centimeter). For the same time, the
“MSRE inventory activity was obtained and dividied by
the total MSRE area (metal, 0.79 X 10® c¢cm?;graphite
channels, edges, ends, and lattice bars, 2.25 X 10% cm?,
for a total of 3.04 X 10° ¢m?), giving an inventory
value (disintegrations per minute per square centimeter)
that would result if the nuclide deposited evenly on all
surfaces. The ratio of observed to inventory disintegra-
tions per minute per square centimeter then yields a
relative deposit intensity which may be tabulated along
with the inventory disintegrations per minute per
square centimeter.

The relative deposit intensities, of course, will average
between O and 1.0 when summed over "all areas,
indicating the average fraction of inventory deposited
on the reactor surface. The values may be compared
freely between runs, nuclides, regions, kinds of surfaces,
and qualities of flow. The fraction of the total
inventory estimated to be deposited on a particular
surface domain is the product of the relative deposit
intensity attributed to it and the fraction of the total
area it represents. In particular, all the metal of the
system represents 26% of the total area, graphite edges
a similar amount, and flow channels (and lattice bars)
about 48%.

Results of the examination of metal and graphite
samples from surveillance arrays removed after runs 7,
11, 14, and 18 are shown, respectively, in Tables 9.1,
9.2, 9.3, and 9.4, expressed in each -case as relative
deposit intensities.

For each sample set the metal specimens are listed
from the reactor top to the bottom, followed by
graphite specimens. The nuclides with noble-gas pre-
cursors are followed by the salt-seeking isotopes,
followed by noble metals and ending with tellurium and
iodine.

A number of generalizations of interest may be noted.
As a base line, in the absence of minute salt particles
which might have come with a sample, recoils from
fission in adjacent salt should give a relative deposit
intensity of 0.001 to 0.003. The salt-seeking elements
(Ce, Nb, Zr) on both metal and graphite and the
elements with noble-gas precursors (Sr, Y, Ba, Cs) on
metal specimens do show such levels of values, even
generally being higher near the core midplane con-
sistent with the higher flux.

For graphite, the elements with noble-gas daughters
have some tendency to diffuse into the graphite in the
form of their short-lived noble-gas precursor. Thus the
values for ®°Sr are mostly in the 0.10—0.20 range,

65

indicating that appreciable entry has indeed taken
place. Values for '*®Ba and °!Y are an order of
magnitude lower, consistent with noble-gas precursors
of much shorter half-life; their values are still about an
order of magnitude above those of the salt-seeking
elements. On specimens of pyrolytic graphite, which
had appreciably less internal porosity, the entry perpen-
dicular to the graphite planes was less than that where
the ‘edges of the planes faced the salt; both directions
were lower than CGB graphite. The data for ! 37Cs were
more than an order of magnitude lower than for
strontium, though the half-lives of the noble-gas pre-
cursors were similar (3.18 min for °Kr and 3.9 min for
137Xe). The inventory data used in the tabulation were
for all material built up since first power operation;
such an inventory is severalfold too high for the later
runs in cases such as this where transport rather than
salt accumulation is important. However, inventory
adjustment is not sufficiént to account for the low
levels of '37Cs values. Appreciable diffusion of cesium
from the graphite, as discussed later, appears to account
for the low values.

The noble metals (Nb, Mo, Ru, Ag, Sb, Te) as a group
exhibited deposits relatively more intense than other
groups on graphite and even more intense deposits on
metal, where the relative intensities on various samples
approached or exceeded 1 and were in practically all
cases above 0.1. Significant percentages of the noble
metals appear to have been deposited on system
surfaces. '

Generally the deposit intensities on the 2-in.-OD
perforated container cylinder (after runs 7, 11, and 18)
were higher than on the '“-in.-OD flux monitor tube,

which was attached to a stringer within the container.

Flow was turbulent around the container cylinders but
was less turbulent and may have been laminar along the
internal tube.

The most intensely deposited elements appear to have
been tellurium, antimony, and silver; on the container
cylinder the relative deposit intensities were mostly
above 1. The deposit intensities for flux monitor tubes
were at least severalfold greater than for graphite, which
should have had similar flow, so that we can conclude
that these elements had definitely more tendency to
deposit on metal than on graphite under comparable
conditions.

To a degree only slightly less intense, molybdenum
exhibited similar behavior to -that described above.
Niobium appeared to have deposited with roughly
similar intensities on graphite and metal; deposition
became somewhat less intense and varied proportion-
Table 9.1. Surveillance specimen data: first array, removed after run 7

Ratio [(obs dis min ! ¢cm™2)/(inventory dis min~! cm™2)]

Specimen

89Sl. 91Y 140_Ba 137CS ldlCe 144Ce l47Nd QSZl. 95Nb 99M0 103Ru 132Te 131[
Hastelloy N
Top 0.0004 0.0004 0.019 0.83 0.56 2.8 0.068
Middle 0.0011 0.0014 0.027 0.79 0.37 24 0.044
~ Bottom 0.0016 0.0036 0.019 1.07 0.40 1.9 0.033
Graphite
Top
Wide, salt 0.16 0.015 0.022 0.005 0.0019 0.0006 <0.0004 0.0004 0.16 0.18 0.19 0.17 0.0012
Side 1, salt ' 0.004 0.0020 0.0005 0.10 0.10 0.12 0.13 0.0011
Side 2, salt 0.026 0.0031 0.0008 0.14 0.19 0.10 0.26 0.0023
Wide, graphite - 0.010 0.012 0.004 0.0016 0.0005 0.00014 0.0007 1.00 0.08 0.04 0.09
Middle
Wide, salt 0.14 0.0001 0.019 0.003 0.00{17 0.0016 <0.0003 0.0018 0.80 0.15 0.09 0.19 0.0027
Side 1, salt 0.002 0.016 0.0050 0.0017 0.00005 0.45 0.13 0.08 0.17 0.0037
Side 2, salt 0.023 0.0043 0.0017 0.74 0.12 0.06 0.15 0.0027
Wide, graphite 0.003 0.020 0.0032 0016 <0.0003 0.0016 0.61 0.02 0.07 0.17
Bottom
Wide, salt 0.25 0.008 0.051 0.092 0.0043 0.0038 0.00024 0.0035 0.67 0.21 0.12 0.19 0.0036
Side 1, salt 0.010 0.046 0.001 0.0079 0.0037 0.00017 1.03 0.21 0.13 0.23 0.0043
Side 2, salt 0.005 0.042 0.002 0.0091 0.0031 0.00009 0.66 0.21 0.13 0.21 0.0044
Wide, graphite 0.024 0.018 0.0060 0.0037 0.0003 0.15 0.05 0.05 0.15 0.0048
Inventory? (dis min™! cm™2)
8.8E10 8.8E10 2.2E11 7.6E8 14E11 2.5E10 8.2E10 9.6E10 . 3.0E10 26E11 64E10 1.8E4 1.2E11

2MSRE inventory activity divided by the total MSRE area.

99
Table 9.2. Survey 2, removed after run 11, inserted after run 7

_ Ratio [(obs dis min~' cm™)/(inventory dis min™" cm 2)]
Specimen 89g, 95, 95 Nb 9910 1035, 106, 132, 131]
Hastelloy N perforated 0.0010 0.0011 0.7 1.7 0.09 5.2 0.013
cylindrical container
Near top 0.0054 0.0007 1.5 2.3 0.91 4.4 0.033
Middle 0.0263 0.0011 1.5 1.7 0.13 1.6 0.027
Bottom 0.0005 0.0002 0.11 04 1.5
Graphite (CGB) specimens
Top
Wide, salt 0.0015 0.60 0.37 0.136 0.064 0.28
Side 1, salt 0.0011 0.33 0.12 0.039 0.049 0.11
Side 2, salt 0.0016 0.40 021 0.076 0.074 0.014
Wide, graphite 0.0013 0.31 0.16 0.060 0.057 0.14
Middle
Wide 0.0018 0.87 0.32 0.074 0.078 0.17
Side 0.0015 0.76 0.13 0.059 0.065 0.14
Side 0.0038 2.0 0.31 0.159 0.167 0.34
Wide 0.0017 0.93 0.36 0.083 0.075 0.17
Bottom
Wide 0.0033 1.0 0.20 0.136 0.145 0.36
Side 0.0017 0.13 0.19 0.059 0.073 0.20
Side 0.6030 1.2 0.47 0.110 0.131 0.28
Wide 0.0038 0.17 0.165 0.26
Inventory? (dis min~! cm™2)
1.8E11 2.1E11 1.5E11 1.7E11 1.2E11 4.8E9 1.3E11 1.2E11

4MSRE inventory activity divided by the total MSRE area,

ately more widely during the ?33U operation period
ending with run 18.

Ruthenium appeared to be about half (or less) as
intensely deposited as molybdenum and niobium, on
both graphite and metals. Like the other noble metals,
it was more intensely deposited on the metal cylinder
container.

The 39.6-day '°?Ru and the 367-day '°®Ru offer
the possibility of some insight into deposition processes
by comparison of their deposit intensity ratios. In
particular, as will be developed in detail later, if the
longer-lived isotope is present in higher relative in-
tensity and if it is assumed that reasonably similar and
steady deposition conditions have prevailed, then some
kind of retention or holdup must have occurred prior to
formation of the observed deposit.

By and large this appears to have been the case with
ruthenium, and also to some extent with tellurium.

In these tables, with respect to '®¢Ru values, two
factors which, though significant, appear at least
roughly to offset each other have not been entered.
First, the inventory used is that accumulated from first

power (January 1966) rather than during the interval of
exposure. Perhaps the lower exposure period value
could be used. Second, for '°®Ru, in particular during
235U operation (runs 4 to 14), the fission yield
increased as 239Pu grew into the fuel; about 5% of the
fissions at the end of run 14 were from this source, and
the 1°6Ru yield of the fuel was roughly doubled. This
would lead to increasing inventories. The effects are
approximately compensatory, and we did not correct
for them in this table, though they were included in the
“compartment’ model described later.

We have noted that less intense deposits were ob-
served on the flux monitor tube than on the perforated
cylinder container, where the flow was doubtless more
turbulent. The flow conditions at graphite surfaces and
flux monitor tube surfaces would appear to be more
nearly similar, but differences are difficult to assess.

Mechanisms for fission product deposition from
flowing salt must certainly involve a mass transfer step,
as well as a statement as to the areas to which it applies
and an assumption as to the properties and species
(atomic, colloidal) undergoing transport. Usually a
68

Table 9.3. Third surveillance specimen survey, removed after run 14

Numbers given represent relative deposit intensity assuming uniform deposition on all surfaces
Specimens inserted after run 11 unless otherwise noted

1.’osmon Ratio [(obs dis min ! cm ?){(inventory dis min * cm %))
Specimen ;Td:lrailrg) Face 89, Sly  140p, 1370, 18lg, 1440, 187Ny 957, 95N, 99y, 103p, 106p, 110m,, 129My, 1327, 1314
Hastel- +30b 0.0010 0.0010 0.0008 0.00002 0.0035 0.0003 0.0007 0.0006 0.20 0.39 0.104 0.098 0.28 2.0 0.72 0.012
loy N? +23 0.0021 0.0020 0.0009 0.0009 0.0014 0.44 0.57 0.093 0.109 1.8 0.64 0.015
19 0.039 0.0035 0.0019 0.0020 0.0033 0.55 0.71 0.114 0.150 2.1 0.91 0.017
0°¢ 0.0044 0.0075 0.0016 0.0020 0.0024 0.0013 0.0019 0.0034 0.63 1.0 0.104 0.137 1.14 2.2 0.89 0.015
-9 0.0036 0.0030 0.0020 0.0011 0.0032 0.53 0.93 0.114 0.147 1.9 0.72 0.010
-19 0.0027 0.0024 0.0016 0.0010 0.0028 0.55 1.04 0.082 0.115 1.8 0.66 0.003
—28d 0.0028 0.0029 0.0017 0.0007 0.0010 0.0010 0.0010 0.0015 0.51 0.67 0.092 1.26 0.57 2.3 0.91 0.010
Graphite
CGB* +27 Wide 0.14 0.022 0.0023 0.0006 0.0008 0.16 0.085 0.035 0.053 0.47 0.12 0.0014
Narrow 0.11 0.019 0.0017 0.0004 0.0005 0.04 0.034 0.013 0.021 0.28 0.08 0.0005
CGBf +27 Wide 0.12 0.023 0.0019 0.0004 0.0007 0.19 0.12 0.051 0.057 0.19 0.09 0.0011
Narrow 0.13 0.023 0.0020 0.0004 0.0006 0.24 0.11 0.042 0.051 0.17 0.08 0.0011
Pyrolytic  +20.8 1§ 0.003 0.002 0.0008 0.0007 0.0010 0.34 0.13 0.094 0.078 0.10 0.05 0.0007
llh 0.019 0.10 0.0012 <0.0012 0.0012 0.20 0.08 0,035 <0.043 0.18 0.10 0.0008
Poco +12.5 Wide 0.10 0.024 0.0075 0.0018 0.0028 0.50 0.025 0063 <0.070 0.71 0.26 0.0033
Narrow 0.08 0.20 0.0044 0.0019 0.0020 0.44 0.096 0066 <0.068 1.71 0.23 <0.0040
CGB +4.5 Wide 0.21 0.40 0.0075 0.0017 0.0033 1.0 0.15 0.071 0.089 0.14 0.06 0.0009 )
CGBf +4.5 Wide 0.10 0.038 0.0071 0.0024 0.0029 09 0.0078 0.094 0.15
Na\rrow 0.12 0.034 0.0089 0.0023 0.0032 1.2 0.093 0.110 0.15
CGB° -4.5 Wide 0.14 0.035 0.0084 0.0024 0.0040 0.54 0.16 0.050 0.076 0.51 0.13 0.0034
Narrow 0.10 0.034 0.0062 0.0023 0.0028 0.59 0.085 0.041 0.051 0.14 0.06 0.0029 .
CGB —-4.5 Wide 0.16 0.051 0.0102 0.0024 0.0044 1.6 0.16 0.090 0.105 0.20 0.08 0.0053
CGB -12.5 Wide 0.15 0.038 0.0066 0.0018 0.0036 1.2 0.64 0.041 0.043 0.08 0.20 0.0019
CGBf -12.5 Wide 0.16 0.047 0.0062 0.0015 0.0026 1.0 0.059 0.065 0.13
Narrow 0.10 - 0.043 0.0066 0.0018 0.0032 1.2 0.094 0.114 0.17
CGBf —30d Wide 0.18 0.022 0.0049 0.0017 0.0019 0.17 0.117 0.065 0.078 0.16 0.07 0.0018
Narrow 0.12 0.017 0.0032 0.0010 0.0020 0.75 0.096 0.054 0.048 0.12 0.05 0.0017
CGB® —30° Wide 0.15 0.018 0.0035s 0.0015 0.0016 0.29 0.081 0.039 0.061 0.29 0.11 0.0015
Narrow 0.15 0.017 0.0031 0.0013 0.0016 0.27 0.067 0.037 0.058 0.21 0.08 0.0013

lnventoryi (dis min ! cm %)

1.7E11 2.0E4 2.5E1l1 6.4E9 2.3E11 1.3E11 8.7E10 2.1E11 1.7E11 2.8E11 1.1E11 7.5E9 (5E7)'I 5/8E9 2.0E11 13EIll

@1.-in.-OD flux monitor tube attached to stringer.

brop.

“Midplane.

dBottom.

€Inserted after run 7.

by Impregnated.

&Deposition perpendicular to graphite planes.
’_’Deposition parallel to graphite planes.

IMSRE inventory activity divided by the total MSRE area.
JApproxjmate.

69

Table 9.4. Fourth surveillance specimen survey, removed after run L8

No flush salt after drain
U-233 operation began with run 15
Specimens inserted after run 14 unless otherwise noted

_ E_’osition Ratio [(obs dis min ' em™?)/(inventory dis min ! cm )]
Specimen xf:irlcipfir:nrrel) Face 89¢, 91y  140p. 137 141, 1840, 14744 957 95\ 99y 103p, 106p. 1Il,, 125gy (29Mp 132, 131)
Hastelloy N

Perforated cylindrical  Top 0.0038 0.0038 0.0027 0.0023 0.0003 0.0015 0.16 0.76 0.26 0.94 2.6 31 1.9 0.072

container? Bottom 0.0075 0.0035 0.0035 0.00l6 0.0010 0.0029 0.73 1.1 0.28 1.05 3.6 2.6 2.8 0.093
Flux monitor tube? +30° 0.0011 0.0004 0.0043 0.0004 0.0003 0.0007 0.08 0.05 0.11 0.14 0.10 0.59 0.67 0.53 0.030

+23 0.0025 0.0007 0.0028 0.0006 0.0006 0.0010 0.09 0.94 0.12 0.17 0.38 0.09 0.43 0.38 0.044

+9 0.0031 0.0025 0.0027 0.0011 0.0010 0.0019 0.10 0.66 0.09 0.17 0.21 0.31 0.78 0.49 0.041

Od 0.0030 0.0035 0.0070 0.0012 0.0014 0.0024 0.14 1.07 0.14 0.29 0.42 0.15 0.81 0.52 0.048

-9 0.0030 0.0036 0.0027 0.0010 0.0013 0.0024 0.19 1.27 0.24 0.45 0.62 0.15 3.7 0.50 0.050

-19 0.0030 0.0028 0.0030 0.0010 0.0013 0.0020 0.12 098 0.11 0.19 0.51 0.81 0.87 0.47 0.033

—-29¢ 0.0380 0.0086 0.0111 0.0267 0.0019 0.0031 0.05 0.47 0.19 0.35 0.72 0.57 1.6 0.91 0.068

Graphitef
CGBf +27¢ Wide 0.12 0.015 0.011 0.009 0.0006 0.0036 0.036 0.034 0.018 0.033 0.69 0.057 0.15 0.083 0.0033
Narrow (.16 0.025 0.013 0.013 0.0009 0.0010 0.046 0.050 0.013 0.032 0.70 0.0d6 0.14 0.114 0.0064
Od Wide 0.21 0.033 0.014 0.035 0.0016 0.0013 0.180 0.035 0.024 0.054 0.87 0.119 0.09 0.08 0.0014
Narrow 0.30 0.015 0.026 0.038 0.0024 0.0016 0.095 0.030 0.024 0.059 1.31 0.082 0.17 0.13 0.0097
—27¢ Wide 0.18 0.023 0.017 0.018 0.0028 0.0027 0.33 0.11 0.053 0.096 0.97 0.21 0.18 0.15 0.0085
Narrow 0.31 0.039 0.028 0.018 0.0028 0.0028 0.38 0.11 0.069 0.131 1.03 0.15 0.26 0.17 0.0090
CGB +27¢ Wide 0.14 0.017 0.013 0.002 0.0005 0.0006 0.036 0.044 0.10 0.006 0.63 0.011 0.065 0.061 0.0044
Narrow 0.13 0.050 0.014 0.0008 0.0007 0.042 0.092 0.040 0.035 0.82 0.016 0.083 0.089 0.0037
CGBh +27¢ Wide 0.11 0.018 0.010 0.006 0.0004 0.0005 0.020 0.064 0.026 0.013 0.47 0.008 0.048 0.052 0.0031
CGB +24 Wide 0.14 0.023 0.0009 0.044 0.097 0.026 0.022 0.60 0.010 0.061 0.058 0.0052
Namow 0.18 0.020 0.010 0.001 0.0006 0.024 0.040 0014 0.015 0.55 0.005 0.041 0.051 0.0042
Poco” +19 Wide 0.14 0.031 0.016 0.002 0.0003 0.005 0.010 0.004 0.004 0.12 0.001 0.059 0.040 0.0019
Narrow 0.15 0.033 0.019 0.001 0.0021 0.069 0.091 0.044 0.043 1.60 0.015 0.083 0.090 0.0067
CGB 0?  wide 0.6 0030 0022 0.005 0.0012 0.0037 0.14 0067 0037 0043 067 0016 0.109 0.045 0.0046
Narrow 0.12 0.028 0.016 0.001 0.0010 0.0015 0.17 0.069 0.026 0.023 0.55 0.004 0.039 0.059 0.0015
CGB oY  wide 024 0031 0024 0004 0.0020 00036 0.14 0083 0055 0071 094 0025 0072 0.052 0.0054
Narrow 0.24  0.032 0.024 0.004 0.0016 00027 021 0063 0052 0054 124 0031 0057 0067 0.0042
CGB -27¢ Wide 0.43 0.048 0.036 0.009 0.0108 0.102 0.052 0.036 0.033 1.17 0.016 0.078 0.0022
Narrow (.28 0.026 0.022 0.003 0.0079 0.064 0.046 (0.003Y 0.030 1.16 0.026 0.074 0.078 0.0042
cGB" ~27°  Wide 0.5 0014 0014 0.004 0.0008 00015 0.057 0041 0023 0025 079 0013 0037 0032 0.0018
Narow 030  0.002 0.024 0.006 0.0001 0.0011 0.093 0047 0027 0029 064 0013 0035 0032 0.0022
Inventory' (dis min ' cm™2)

2.0E11 1.8Ell 23Ell 8.3E9 9.3E10 1.9E11 14E11 19E11 6.8E10 5.2E9 1.0E7 44E8 1.3E10 1.8EI1 1.2Ell

“Inserted after run 11. Salt flow in the low turbulent range.

blé-in.-OD flux monitor tube attached to stringer. Salt flow barely turbulent.
“Top.

9Midplane.

€Bottom.

fall samples taken from faces exposed to flowing fuel salt.

&Inserted after run 7.

’.’Impregnated.

IMSRE inventory activity divided by the total MSRE area.

IApproximate.

factor related to the permanent adherence of deposited
material to a given surface, the so-called *‘sticking
factor,” is included, usually followed by the assumption
that for lack of data it will be assumed that all metal
and graphite surfaces have equal values of unity —
whatever hits, sticks and stays.

9.2 Final Assembly

9.2.1 Design. The final surveillance specimen array,
inserted after run 18 and removed after run 20, was of a
different design®** from those previously used, in order
to include capsules containing substantial quantities of
233J and other isotopes to determine accurately their
neutron capture characteristics in the MSRE spectrum.

The cylindrical geometry permitted the inclusion of a
surveillance array consisting of sets of paired metal and
graphite specimens with differing axial positions, sur-
face roughness, and adjacent flow velocities. Because
flow conditions were the same or essentially so for
metal-graphite pairs, the hydrodynamically controlled
mass transport effects, if simple, should cancel in
comparisons, and differences can be attributed to
differences in what is commonly called sticking factor.
The sample pairs are discussed below in order of
increasing turbulence.

A photograph of the final surveillance specimen
assembly is shown in Fig. 9.7. The individual specimens
and the flow associated with them will be considered
next.

9.2.2 Specimens and flow. In the noncentral regions
of the core, the flow to a fuel channel had to pass

HASTELLOY N BASKET

through the grid of lattice bars, and according to
measurements reported®’® on models, the velocity in
the channels was 0.7 fps with a Reynolds number of
1000. However, the flow varied with the square root of
head loss, implying that nonlaminar entrance conditions
extended over much of these channels.

The lattice bars did not extend across the central
region. The flow through central fuel channels was
indicated by model studies to be 3.7 gpm, equivalent to
2.66 fps, or a Reynolds number of 3700; the associated
head loss due to turbulent flow can thus be calculated
as 0.45 ft. In this region were also the circular channels
for rod thimbles and surveillance specimens; the same
driving force across the 2.6- to 2.0-in. annulus yields a
velocity of 2.6 fps and a Reynolds number of 3460.
These flows are clearly turbulent.

Flow in the circular annulus around the surveillance
specimen basket essentially controlled the pressure
drops driving the more restricted flows around and
through various specimens within the basket.

At the bottom of the basket cage was a hollow
graphite cylinder (No. 7-3, Table 9.5) with a 1'%-in.
outside diameter and a °-in. inside diameter, contain-
ing a '4-in.-OD Hastelloy N closed cylinder (No. 7-1,
Table 9.6). The velocity in the annulus was estimated as
0.27 fps, with an associated Reynolds number of
DVp/u =0.0104 X 0.27 X 141/0.00528 = 75; this flow
was, therefore, clearly laminar. This value was obtained
by considering flow through three resistances in series —
respectively, 20 holes in parallel, ' in. in diameter, %
in. long; then 6 holes in parallel, ', in. in diameter, ¥,

PHOTO 96504

. FLOW TUBE ASSEMBLY
. U CAPSULE (233-238)
. U CAPSULE (234-238)
. U CAPSULE (233-238)
. U CAPSULE (234-238)

DDHBUN -
QVwm~NO

. PYROLYTIC GRAPHITE

. FISSION PRODUCT DEPOSITION TEST SPECIMEN (GRAPHITE)

. FISSION PRODUCT DEPOSITION TEST SPECIMEN (HASTELLOY N)
. GAS TRAP SPECIMEN (HASTELLOY N)

. SINKER (HASTELLOY N)

Fig. 9.7. Final surveillance specimen assembly.

71

Table 9.5, Relative deposition intensity of fission products on graphite surveillance specimens from final core specimen array

Observed dpm/cmzl(MSRE inventory as isotope dpm/MSRE total metal and graphite area, cm?)
Activity and inventory data are as of reactor shutdown 12/12/69
Numbers in parentheses are (MSRE inventory/MSRE total area), dpm/cm?
B, M, and T in sample numbers signify bottom, middle, and top regions of the core

. Roughness Cenftri(r::ters Sambie No 89¢, 137~ 140p, 1416 144 957, 9SNb 99Mo 103, 106R, 125y 132, 129m, 131]
ype (uin.) core center P " (1.37E11%  (8.53E9%) (1.73E11) (1.83Ell) (8.05E10% (1.35E11%) (1.14E11) (2.26E11) (4.48E10%) (4.47E9%) (5.63E8) (1.85E10) (1.06E11)

Outside 5 -29 7-3-1-B outer 4.2 0.023 0.17 0.0039 0.0036 0.0021 0.21 0.083 0.069 0.011 0.0805
(transition flow) 25 -27 7-3-1-M outer 1.6 0.022 0.13 0.0019 0.21 0.033 0.033 0.046 0.0059
125 -25 7-3-1-T outer 2.4 0.08 0.0031 0.0008 0.0018 0.18 0.035 0.035 0.029 0.0035
Outside with wire 5 +8 12-1-B outer 1.9 0.17 0.0069 0.0013 0.0025 0.25 0.040 0.0001 0.012 0.0033
(turbulent flow) 25 +10 12-1-M outer 2.0 0.17 0.0090 0.0015 0.0023 0.15 0.039 0.039 0.025 0.0047
125 +12 12-1-T outer 1.8 0.16 0.0037 0.0005 0.0021 0.15 0.050 0.0059
Inside annulus 5 -28 7-3-1-B innex 0.31 0.0058 0.0l6 0.0012 0.0006 0.0014 0.04 0.003 0.022 0.019 0.065 0.0033
(laminar flow) 5 =26 7-3-1-M inner 0.26 0.0016 0.009 0.0012 0.0006 0.0016 0.25 0.22 0.150 0.108 0.054 0.0026
125 24 7-3-1-T inner 0.18 0.0015 0.009 0.0012 0.0011 0.0016 0.24 0.19 0.064 0.049 0.579 0.0010
Inside tube 5 +9 12-1-B inner 0.49 0.0029 0.046 0.0031 0.0009 0.0027 0.25 0.056 0.047 0.061 0.0031
(transition flow) 5 +11 12-1-M inner 0.34 0.0010 0.028 0.0018 0.0009 0.0011 0.20 0.035 0.029 0.057 0.0019
125 +13 12-1-T inner 0.39 0.0032 0.033 0.0010 0.0002 0.0018 0.09 0.084 0.065 0.050 0.0017

127

991 Te
(Inv = 2.9E9)
Postmortem: MSRE 0.049 0.23 0.56 0.44

core bar segment

®Inventories shown accrue from all operation beginning with original startup. To correct inventories to show the material produced during current period (runs 19 and 20) only, multiply by factors given below. To obtain similarly corrected

deposition intensity ratios, divide the value in the table by the factor for the isotope. Factors are: 52-day 89gr = 0.90; 59-day

0.14. For isotopes with shorter half-lives, corrections are trivial.

9

Y = 0.86; 40-day 1°3Ru = 0.95; 65-day °5Zr = 0.84; 284-day ' **Ce = 0.36; 1-year 1°®Ru = 0.32; 30-year 137Cs =
72

Table 9,6. Relative deposition intensity of fission products on Hastelloy N surveillance specimens from final core specimen array

Observed dpm/cmzl(MSRE inventory as isotope dpm/MSRE total metal and graphite area, cm?)
Activity and inventory data are as of reactor shutdown 12/12/69
Numbers in parentheses are (MSRE inventory/MSRE total area), dpm,‘cm2
B, M, and T in sample numbers signify bottom, middle, and top regions of the core

Type Roughness Cenftrl;neters Sample 89¢r 137¢s 140p, 141 e Wlee 957¢ 75Nb Mo 103pu 106 Ry 125gh 13271, 129mr 3y
yp (uin.) core cemnter No. (1.37E119) (8.53E9%) (1.73El1l) (1.83El11) (8.05E10%) (1.35E11%) (1.14E11) (2.26El11) (4.48E10% (4.47E9%) (5.63E8) (2.01E11) (1.06E11)
Outside (transition flow) 5 +24 14-3-B 0.0016 0.0015 0.0012 0.0009 0.0004 0.0004 0.13 34 0.094 0.127 6.4 0.060
25 +25 14-3-M 0.0013 0.0056 0.0009 0.0008 0.0003 0.0003 0.12 1.4 0.059 0.078 1.9 0.051
125 +27 14-3-T 0.0010 0.0006 0.0007 0.0006 0.0003 0.0003 0.14 1.7 0.056 0.079 2.2 0.046
Qutside with wire (turbulent flow) 5 +16 13-2-B 0.0013 0.0009 0.0013 0.0009 0.0004 0.0003 0.26 0.46 0.10 0.09 1.1 0.22
25 +18 13-2-M 0.0020 0.0253 0.0018 0.0011 0.0005 0.0007 0.34 2.2 0.20 0.17 2.3 0.37
125 +20 13-2-T 0.0039 0.0022 0.0030 0.0012 0.0005 0.6004 0.49 0.32 0.10 0.08 3.0 0.60
Wire +18 13-3 wire 0.0019 0.0006 0.0015 0.0010 0.0001 0.0005 0.21 0.85 0.14 0.23 0.17 0.09
Wire +11 12-2 wire 0.0030 0.0011 0.0027 0.0016 0.0008 0.0008 0.88 1.7 0.25 0.19 0.79 0.09
Inside annulus (laminar flow) 5 -28 7-1-B 0.0018 0.0006 0.0015 0.0011 0.0005 0.0006 0.21 1.2 0.09 0.06 0.87 0.13
125 —-26 7-1-T 0.0020 0.0007 0.0017 0.0013 0.0006 0.0006 0.39 1.4 0.15 0.23 0.93 0.19
Inside tube [transition (?) flow] 5 +16 13-1-B 0.0021 0.0005 0.0020 0.0014 0.0006 0.0007 0.37 3.7 0.34 0.32 1.2 0.18
5 +16 13-1-M 0.0021 0.0012 0.0018 0.0013 0.0006 0.41 4.1 0.19 0.19 1.1 0.18
125 +20 13-1-T 0.0019 0.0007 0.0015 0.0011 0.0005 0.0004 0.71 3.6 0.23 0.17 3.0 0.38
Stagnant (inside, liquid region) +26 14-2-L2 0.0030 0.0020 0.0002 0.00003 0.00002 0.00001 0.0051 6.0 0.005 0.007 0.03 0.002
+24 14-2-L1 0.0010 0.0002 0.0002 0.00011 0.00006 0.00006 0.025 3.0 0.044 0.043 0.13 0.010
Stagnant (inside, gas region) +29 14-2-G1 0.14 0.0027 0.0057 0.00001 0.000005 0.00003 0.0010 4.5 0.0012 0.0010 0.14 0.0008
+28 14-2-G2 0.47 0.0022 0.0077 0.00002 0.00001 0.00002 0.0012 5.3 0.0006 0.0006 0.03 0.0028
+27 14-2-G3 0.41 0.0014 0.0069 0.00004 0.00001 0.00086 0.0012 6.6 0.0009 0.0009 0.03 0.0005
127
99Tc Te
{Inv = 2.9E9)
Postmortem
MSRE heat exchanger segment 0.20 0.55 0.13 1.4 1.0
MSRE rod thimble segment (core) 0.83 0.66 0.32 1.6 1.0

?Inventories shown accrue from all operation beginning with original startup. To correct inventories to show the material produced during current period (runs 19 and 20) only, multiply by factors given below. To obtain similarly corrected deposition
intensity ratios, divide the value in table by the factor for isotope. Factors are: 52-day 8%Sr = 0.90; 59-day °' Y = 0.86;40-day '®3Ru = 0.95; 65-day ?5Zr = 0.84; 284-day 14%Ce = 0.36; 1-year 126 Ru = 0.32; 30-year ! 37Cs = 0.14. For isotopes with shorter

haif-lives, corrections are trivial.

in. long;.then an annulus” ;4 in. wide, 5 in. long. A

flow head loss of 0.057 ft, which should develop along
the outer part of the basket, was assumed.

In the annulus between the outside of the graphite
cylinder and the basket, the velocity was estimated to
be about 1.5 fps; the associated Reynolds number is
2200, and the flow was either laminar or in the
transition region. Some distance above, at the top of
the assembly, was a Hastelloy N cylinder (No. 14, Table
9.6; No. 1 in Fig. 9.7) of similar external dimension,
which presumably experienced similar flow conditions
on the outside. This specimen was closed at the top and
had a double wall. Inside was a bar containing electron
microscope screens. The liquid around the bar within
the cylinder was stagnant, and gas was trapped in the
upper part of the specimen.

Below this, above the midplane of the specimen cage
were located respectively graphite (Table 9.5, No. 12)
and double-walled Hastelloy N (Table 9.6, No. 13)

73

cylinders, with connecting ';-in.-diam bores. Flow

through this tube is believed to have been transition or
possibly turbulent flow, though doubtless less turbulent
than -around the specimen exterior. The exterior of the
1-in.-OD cylinders was wrapped with Y 4-in. Hastelloy
N wire on %-in. pitch as a flow disturbance. Flow in
the annulus between the specimen exterior and the
basket was undoubtedly the most turbulent of any
affecting the set of specimens.

The data from the various specimens are presented in
Table 9.5 for graphite and Table 9.6 for Hastelloy N in
terms of relative deposit intensity.

We saw no effect of surface roughness, which ranged
from 5 to 125 uin. rms, on either metal or graphite, so
this will not be further considered here.

9.2.3 Fission recoil. Because the specimens were
adjacent to fissioning salt in the core, some fission
products should recoil into the surface.® We calculate
that where the fission density equals the average for the
core, the relative impingement intensity of recoiling
fission fragments, (recoil atoms per square centimeter)/
(reactor production/surface area), ranges from 0.0036
for light fragments to 0.0027 for heavy fragments. The
ratio will be higher (around 0.005) where the fission
density is highest.

9.2.4 Salt-seeking nuclides. Relative deposition in-
tensities for salt-seeking nuclides (°°Zr, '*!Ce) are of
the order of the calculated impingement intensities or
less: for ®5Zr, 0.001 to 0.0027 on graphite and 0.0003
to 0.0008 on metal; for '*!Ce, 0.0010 to 0.0090 on
graphite and 0.0006 to 0.0016 on metal. The deposi-
tion of 284-day '#*Ce is consistent with this, on a
current basis after adjustment for prior inventory as
shown in the table footnotes.

There also appears to be some dependence on axial
location, with higher values nearer the center of the
core. Thus all of the salt-seeking nuclides observed on
surfaces could have arrived there by fission recoil; the
fact that remaining deposition intensities on metal
surfaces are consistently less than impingement densi-
ties indicates that many atoms that impinge on the
surface may sooner or later return to the salt. ‘

9.2.5 Nuclides with noble-gas precursors. The nu-
clides with noble-gas precursors (39Sr, !37Cs, 14%Ba,
and, to a slight extent, ' ! Ce) are, after formation, also
salt seekers. They are found to be deposited on metal to
about the same extent as isotopes of salt-seeking
elements, doubtless by fission recoil. However, noble-
gas precursors can diffuse into graphite before decay,
providing an additional and major path into graphite. It
may be seen that values for °Sr, '3 7Cs, and !4 °Ba for
the graphite samples are generally an order of magni-
tude or more greater than for the salt-seeking elements.

[t appears evident that the deposit intensity on
graphite of the isotopes with noble-gas precursors was
higher on the outside than on the inside, both of
specimen 7 and specimen 12. Flow was also more
turbulent outside than inside, and atomic mass transfer
coefficients should be higher. [Flows are not well
enough known to accurately compare the outside of the
lower specimen (No. 7-3) with the inside of the upper
graphite tube specimen (No. 12-1)].

Appreciably more 3.1-min 8°Kr and 3.9-min '37Xe
should enter the graphite than 16-sec '?%Xe or 2-sec
141 Xe but the '37Cs values are considerably lower
than for 82Sr. Only about 14% of the ! ®7Cs inventory
was formed during runs 19 and 20. With this correction,
however, '37Cs deposition intensities still are less than
those observed for strontium. As will be discussed
later® the major part of the cesium formed in graphite
will diffuse back into the salt much more strongly than

“the less volatile strontium; this presumably accounts for

the lower ! ® 7Cs intensity.

At first glance the *‘fast flow” values for ®Sr on
graphite appear somewhat high even though 8?Kr entry
to graphite from salt was facilitated by the more rapid
flow. Similar intensity on all flow-channel graphite
would account for the major part of the ®Sr inven-
tory, while salt analysis for the period showed that the
salt contained about 82% of the 3°Sr. But the
discrepancy is not unacceptable, since most core fuel
channels had lower velocity and less turbulent, possibly
laminar flow.

On the inside of the closed tube the deposition of
39St and other salt-seeking daughters of noble gases was
much higher in the gas space than in the salt-filled
region. This is consistent with collection of ?Kr in the
gas space and the relative immobility of strontium
deposits on surfaces not washed by salt.

9.2.6 Noble metals: niobium and molybdenum.
Turning to the noble-metal fission products, we note
that ® SNb deposited fairly strongly and fairly evenly on
all surfaces. The data are not inconsistent with post-
mortem examination of reactor components, to be
described later. Molybdenum (°?Mo) deposited con-
siderably more strongly on metal than on graphite
(limited graphite data). Because the relative deposit
intensity of molybdenum on metal is similar to that of
8981 on graphite, which is attributed to atomic krypton
diffusion through the salt boundary layer, it may be
that molybdenum could also have been transported in
appreciable part by an atomic mechanism, and pre-
sumably had a high sticking factor on metal (about 1?).
Under similar flow conditions, the deposit intensity of
molybdenum on graphite is much less; hence the
sticking factor on graphite is doubtless much below
unity. Postmortem component examination found that
the ®° Tc daughter also was more intensely deposited on
metal than on graphite.

The widely varying °?Mo values for salt samples
taken during this period, however, imply that a signifi-
cant amount of °?Mo occurred along with other
noble-metal isotopes in pump bowl salt samples as
particulates. Since molybdenum was relatively high in
the present surveillance samples also, it may be that an
appreciable part of the deposition involved material
from this pool.

Because molybdenum deposited more strongly than
its precursor niobium, an appreciable part of the
molybdenum found must have deposited independently
of niobium deposition, and niobium behavior may only
roughly indicate molybdenum behavior at best. This
may well be due to the relation of niobium behavior to
the redox potential of the salt, while molybdenum may
not be affected in the same way.

9.2.7 Ruthenium. The ruthenium isotopes, '°3Ru
and '°¢Ru, showed quite similar behavior as would be
expected, and did not exhibit any marked response to
flow or flux variations. The ruthenium isotopes appear

. to deposit severalfold more intensely on metal than on

graphite. The correction of inventories to material
formed only during the exposure period will increase
the '°6Ru intensity ratios about threefold but will
change the '°3Ru intensity ratio very little. On such a
inventory basis the '°3Ru deposition will then be
appreciably lower than that for !°®Ru; this indicates
that an appreciable net time lag may occur before
deposition and argues against a dominant direct atomic
deposition mechanism for this element.

74

9.2.8 Tellurium. The tellurium isotopes '32Te (on
metals) and '2°™Te (on graphite) show an appreciably
stronger (almost 40 times) relative deposit intensity on
metal than on graphite, indicative of real differences in
sticking factor. Deposit intensities of tellurium were
moderately higher in faster flow regions than in low
flow regions (2 times or more), possibly indicative of
response to mass transfer effects. Flux effects are not
significant.

Postmortem examination of MSRE components
showed '238b and '27Te deposition intensities which
were consistent with this, except that the deposition
intensity of tellurium on the core fuel-channel surfaces
was higher than we observe here on surveillance
specimens.

On balance, it appears that the sticking factor of
tellurium on metal is relatively high. This might result
from direct atomic deposition andfor deposition on
particulate material which then deposits selectively but

. securely. Such strong intensity of tellurium in the

deposits could be the result of direct deposition of
tellurium, or of similar prompt deposition of precursor
antimony with retention of the tellurium daughter, or
both. The data do not tell.

9.2.9 lodine. lodine exhibits deposit- intensities
which appear to be at least an order of magnitude lower
than tellurium, both for graphite specimens and, at
considerably higher levels for both tellurium and iodine,
for the metal specimens. The data for the metal surface
mostly vary with tellurium, suggesting that the iodine
found is a result of tellurium deposition and decay, but
with most of the iodine formed having returned to the
salt.

9.2.10 Sticking factors. In general the data of the
final surveillance assembly are consistent with those of
the earlier assemblies. The pairing of the metal and
graphite specimens in the final assembly permits some
conclusions about relative sticking factors that were
implied but less certain in the earlier assemblies.

It appears evident, because the deposition intensity
differs for different isotopes under the same flow
cconditions, that the sticking factor must be below unity
for may of the noble-metal isotopes, either on metal or
on graphite. The deposition intensity appears rather
generally to be higher on metals than on graphite and
could approach unity. In terms of mechanism, low
values of sticking factor could result if only part of the
area was active or if material was returned to the liquid,
either as atoms via chemical equilibrium processes, or
by pickup of deposited particulate material from the
surface. '
The values would be low if the inventory should be
distributed over a larger area than just the sum of
system metal and graphite areas. Such areas might
include the surface of bubbles or colloidal particles in
the circulating salt.

[t is adequately clear, however, that under similar
flow conditions, the intensities of deposits on metal and
graphite surfaces differ significantly for most noble-
metal elements, with more intense deposits of a given
nuclide generally occurring on the metal surface of a
similarly exposed metal-graphite pair.

9.3 Profile Data

9.3.1 Procedure. As described earlier in this section,
for most graphite surveillance specimens a succession of
thin cuts were made inward from each face to depths
frequently of about 50 mils (0.050 in.). These cuts were
individually bottled, weighed, and analyzed for each
nuclide. The summations for the individual faces have
already been given in the earlier tables of this section. It
does not appear expedient to account here for the
individual samples (which would increase the volume of
data manyfold), since most of the information is
summarized in typical profiles shown below for samples

ORNL—DWG 68-14519

Om
B
i

20 30 40
DISTANCGE FROM SURFACE (mils)

10 50

Fig. 9.8. Concentration profile for 137¢s in impregnated
CGB graphite, sample P-92.

75

removed after run 14, We note that improved hot-cell
milling techniques permitted recovery of 95 to 99% of
the removed material for these samples, with little cross
contamination indicated.

The results of these procedures on two samples of
MSRE (CGB) graphite are shown in Figs. 9.8—9.14. The
sample labeled PS8 was a CGB graphite exposed slightly
above the core midplane and was inserted after run 11.
Sample X-13 was exposed slightly below midplane and
was inserted after run 7, being withdrawn and returned
after run 11.

Data for the various nuclides are presented as semi-
logarithmic plots of activity per unit mass of graphite vs
cut depth. Although such a plot permits easy display of
the data, it does tend to obscure the general fact that

ORNL-DWG 68-1456
1012 — i R | S —]
89,
. pu—— Sr
0" \
X
N X-13, WIDE FACE
o —
i h—-l-——_.__—__'__ 140g, |
,0%
o
E
(5]
~
£
o A
44 \ ; 89
0 S P e — B g
o~
< i
\\
\\
T
.
10'° \\
~
1905y
10°
0 10 20 30 40 50

DISTANCE FROM SURFACE (mils)

Fig. 9.9. Concentration profiles for ®°Sr and '*°Ba in two
samples of CGB, X-13 wide face, and P-58.
76

3 ORNL-DWG 68-14520

10
Tl T vt D ORNL-DWG 68-14515
S . . ;
; [oLusl I _—
)
S i e e ‘
2 ) i :
i i '
- ! O SRR P
1010 |
- ] 89g, —
T T |
5 B | D S -
|| ParALLEL TO PLANES — ‘:\, j 107 e =
| :
i 140g, }
2 | ! b ,
N | | |
o? 0 WIDE FACE '*%ga | NARROW FAGE '
® NARROW FACE '*%Bg =1 —— 1 95 Nb
& wiDE Face %% DO
5 o NaRrrow Fack %% 10'2
i -
2 ’7
;
8
o , [ n
. = :
- ! i ! 10”
5 —qt f i ]
} o I — ———
£ ‘ ! i i ~ [
c \ . i i E .
o ! a i
o 2 I : o i
- |
107 ——
=l
1010 |
; ——PERPENDICULAR TO PLANES
|
— —
RS - | . ;
2 —.&&‘k - : — i
\% 895, ( i
|06 \\ \ ' '4 :
[ S e
\ ~ 9
F %—— 10
5 & 1408_4_0 ; l"‘fi - ; -
1 -1
i :
1 o i ..—._-—._ -—
2 ’ '
e o]
10° | —— !
8 i —
I 10 - : A — O
; o : : } . :
5 ' i ; |
i e ‘ |
! i T B
) E— . . | SN S A
i ]
10* L 107 |
0 10 20 30 a0 50 o} 10 20 30 40 50
DISTANGE FROM SURFAGE (mils! DISTANCE FROM SURFACE (mils)
Fi . . 89 140, . : , . 95 95 .
ig. 9.10. Concentration profiles for °’Sr and Ba in Fig. 9.11. Concentration profiles for *°Nb and *°Zr in CGB

pyrolytic graphite. graphite, X-13, double exposure.
10'3 ORNL-OWG 68-14514
- 3
A - —————— o
® WIDE FACE
A a NARROW FACE
1012 |l
-
T
1o'! \.
1
{
1
|
‘\
)
{ \
g 10" !
a 1|
© i
'
|
‘\
A
10° !
= o

\/NARROW FACE

/<

WIDE FACE

F 4

——————

—

N
[ S —

107

o 10 20

30 40

DISTANCE FROM SURFACE (mils)

Fig. 9.12. Concentration profiles for 103Ru on two faces of

the X-13 graphite specimen, CGB, double exposure.

50

77

ORNL-DWG 68 - 14518
10
WIDE FACE
. 'Y 406Ru
a 194ce
o Ylce
10"
¢
a9
o'
o
Nt
£
(=8
o
<r
107 |{
i
\ .
A / N
\ / AW
“%u—/' \ch
-)a\ N
0o — %= 106R,
N7 < \
| i ~ ———
- S AY
1 N N
] [ \\ . N
1 — _\ \\ \_
L - N
R e\
1%\ J
I\
o e, c—
._;_\\
107
o} 10 20

30 40 50
DISTANCE FROM SURFACE {mils)

Fig. 9.13. Concentration profiles for '°®Ru, '*'Ce, and
144Ce in CGB graphite, sample Y-9.
ORN_L-DWG 68-14517

o\ |
\ ‘
1 :
- '"_! - ]
i\
\\
N\
L\ |
10‘0 = \\ \ \
| \— ~
1 N — 1
s [ 1\ \ S~ L
'\ N rot _
| \ \___.l Ce, X-13
o \ \ : ‘ \
9 \\ i \
e = S |
— S '
5 | 1 i ™~ '
. V\ ) N L
\5 \\ | o~ e, P-58
® 2 \\\ \\\
10° \\\ \\
\‘(_- B &___— ‘4408,)(—13
5 ™\ e
AN
N\
2 \\
10’ \\ 149¢e, P-58
|
5
2
10°
0 10 20 30 40 50

DISTANCE FROM SURFACE (mils)

Fig. 9.14. Concentration profiles for 141ce and '*%Ce in
two samples of CGB graphite, X-13, wide face, and P-38.

by far the greatest amount of any nuclide was to be
found within a very few mils of the surface. Conse-
quently what the fission product profiles tend to
display is the behavior of the small fraction of the
deposited nuclide which penetrates beyond the first few
mils.

Additional data on samples from this set of specimens
were obtained by Cuneo and co-workers, using a
technique developed by Evans. The technique offered

78

less possibility of cross contamination of samples.
According to this technique the specimen was generally
cut longitudinally and at the midplane, and a core was
drilled to the outer surface. The core was then glued to
a cold graphite coupon, which in turn was glued to a
machined steel piston. This piston, the position of
which could be measured accurately using a dial
micrometer, fitted into a holder which was moved on a
piece of emery paper. The resulting powder was Scotch
taped in place, and the total activity of various nuclides
was determined using a gamma-ray spectrometer.

9.3.2 Results. Results using this technique are shown
in Figs. 9.15-9.17.

The material found on or in the graphite doubtless
emerged from the adjacent salt. Transport from salt can
occur by fission product atom recoil from adjacent fuel
salt, by the deposition of elemental fission products
diffusing out of salt as atoms or borne by salt as
colloids, by chemical reaction of salt-soluble species
with graphite, by diffusion of gases from salt and
deposition onto graphite, and by physical transport of
salt into graphite, either by pressure permeation of
cracks, by wetting the graphite, or by sputtering
processes due to fission spikes in salt close to the
graphite surfaces.

Of these, there is no indication of reaction of
fluorides with graphite (with niobium a.possible excep-
tion), and volatile substances are not thought to be a
factor, the noble-gas nuclides excepted. Furthermore,
the graphite did not appear to be wet by salt.
Occasionally there was an indication that salt entered

.cracks in the graphite, and this could be a factor for a

few samples.

The nuclides most certain to be found at greatest
depths should be the daughters of the noble gases.
Profiles for 8°Sr and !#°Ba are shown in Fig. 9.9.

9.3.3 Diffusion mechanism relationships. Among the
ways in which fission products might enter into and
diffuse in graphite are (1) recoil of fission fragments
from adjacent salt, (2) diffusion into graphite of noble
or other short-lived gases originating in salt which on
decay will deposit a nonmobile daughter, (3) formation
of fission products from uranium that was found at that
depth in the graphite, (4) migration of fission products
by a surface diffusion mechanism.

In our consideration of any mechanisms for the
migration of fission products, we will seek (1) estimates
of the amount of activity on unit area, summed over all
depths (expressed as disintegrations per minute per
square centimeter), (2) the surface concentration
(expressed as disintegrations per minute per gram of
graphite), and (3) the variation of concentration with
79

ORNL-DWG 6B-10755R

, 1
1018 _ - : - —rbem— ! M
o : —a— M‘CE I
‘ . [ . —a— 103y,
. ! ; w— 95
FREE-FLOWING SALT | ' Nb
SURFACE B : —e— 7y :
- . : ! _‘;__ I R - e Mdpg i
. el R 4 |
R = ——Tani—2 N L L T -
e U S il i o gy :
1
- - — [ R A ; — - i SERNUU IS S A
!
y H A ———— —
3 N
o' -
- il
S )%/
¢ - Cs 137¢s
© rfl—o—
£ »
E o
f 1015 ;flfi\ 14080 ,
Q 1
v = _ //
§ N R = s
~ | '\'. \\ —td 14080 ‘41ce _._;T'I'A' ]
® VY ~ v
T;; '.\ \__\ * ¥ A""/ Il
E g — —-._
" L[] 7
© 14 \.\J — '7 J l‘
w 10 ) TH— 7—7 Y-l -
€ —A) hY y 4 y .1
S [—¥y ¥ LY ri y b4
< = N FA=s A o 1]
e LN FAVESWANID
AN / 4 17'
mCe Y — —raZ .
S \ X, 7 / "/
-\ +- v ~— 144(:8
o \'Ory st — SN v -
13 Ce v —A—/
10 N X ~ 7 —]
—— - 7 ya /955,
X, e AN _/ f TV ——
AN N 103 L
N\ A — 7 Ru
—a— s A
w-— S i
.
\'<52f \\ &
~ ~ . .
f ™~ . :
10'2 | —a—t
5 -~ -~ STAGNANT SALT
- SURFACE ———— =
! —— |_..._. —
Eomtl et I .
o"

Q 10 20 30 40 50 60 7O B8O 90 100 10 120 {30 440 150 460 {70 180 {90 200 210
DEPTH IN GRAPHITE {mils}

Fig. 9.15. Fission product distribution in unimpregnated CGB (P-55) graphite specimen irradiated in MSRE!cycle ending March
25, 1968.
atoms of nuclide /cm? of grophite

80

0'® . ORNL-DWG 68-10754R
| —
— ]
95Nb
—— 1“mce _..__103Ru
—— e :
X103 R, —a— 103, i
07 37 . -\_
—— 95Nb ] I (0
_._ 95 Zr - : 1
—o— 140g,
1 ]
-
16 I
(o] -
95 -
o5 Nb k:l
J‘ ZrP'I
X 137 B
e
{ b —o——T—0—_ o
ios | I ) Y - :k
I = A
v 140, 190g o
HAL AN i
1y \ = , — .
. /
RS- L '
14 L,:'\\ 7 N _.Z s
10 Yt 144 - = 7——Jm1ai~ ]
A {‘,‘/ Ce v . S oA F A’ 4 Ce
N - 7 1%
ITRIANY ,. yd AN ) -
L R Y — Vi .
. \ \ == _ - . pd 1_7-:
\ oaP 855 A/ il
N —‘\K_ X * —— i 3 N
o, = Y (AN FUS PO
N - Py & ¥
& e Wey -
N > i - : T y
144, 5 . ]
€ —/—} CRACK NOTED -
4 IN 463~-464
M /i SAMPLE WHILE <
/ GRINDING _
* P -
— 4
102

0 10 20 30 40 50 60 70 30 320 330 340 350 360 370 380 390 400 440 420 430 440 450 460
OEPTH IN GRAPHITE (mils)

Fig. 9.16. Fission product distribution in impregnated CGB (V-28) graphite specimen irradiated in MSRE cycle ending March 25,
1968.

atoms of nuclide/cm3 of graphite

i7

S,
@

81

ORNL —DWG 68 -10753R

FREE FLOWING SALT SURFACE

STAGNANT SALT SURFACE

|
1
!
|
i
|

o] 10

20 30 40

50

60 7O 80 90 00

DEPTH IN GRAPHITE

110 120 130

(mils)

140 150 {60 170 {180 190 200 210

Fig. 9.17. Distribution in pyrolytic graphite specimen irradiated in MSRE for cycle ending March 25, 1968.
depth — usually expressed as the depth required to
halve (or otherwise reduce) the concentration. We
should also try to relate the calculated activity to the
calculated inventory activity of fuel salt (expressed as
disintegrations per minute per gram of salt) developed
for the exposure period.

Since 25% of the fissions occurring in salt within one
range unit will leave the salt and doubtless enter the
adjacent graphite, and if we use as applicable to salt the
range of light and heavy fragments in zirconium,
determined as 7.5 and 5.9 mg/cm? respectively, we can
calculate the accumulated recoil activity:

0.0053 for heavy
0.0073 for light °’

_ inv. dpm
g salt

recoil dpm/cm?® X IX {
where 7 is the ratio of local to core average neutron
flux. (I ~ 2 to 4 for core center specimens, depending
on axial position.)

Only in the case of salt-seeking nuclides is recoil a
dominant factor.

The range of fission products entering a graphite of
density 1.86 was determined'® as 3.07 mg/cm? for
95Y and 2.51 mg/cm? for '*%Xe; this corresponds to
16.5 and 13.5 u, so that the penetration should be
limited to a nominal 0.6 mil.

The transport of fission gases in graphite has been
reported! !!? for representative CGB graphites.

The diffusion of noble gases in graphite also involves
diffusion through boundary layers of the adjacent salt.
We will express the behavior in graphite as a function of
the entering flux, and then use this as a boundary
condition for diffusion from salt.

At steady state the diffusion of a short-lived gaseous
nuclide into a plane-surfaced semi-infinite porous solid
has been shown by Evans'? to be characterized by the
following:

Jg = Cg(eDeN)'? = CefDg |
B=(eNDg)''?,
N(y) =Cg exp(—By) ,

where
Cg = atom concentration in surface gas phase,
Jg = atom flux entering suriace,
€ = total porosity of graphite, '
D¢ = Knodsen diffusion coefficient in graphite,

N(y) = concentration of atoms in gas phase in pores at
depth y.

82

The surface concentration of a nonmobile daughter in
graphite resulting from the decay of a diffusing short-
lived precursor is obtained from the accumulation
expression

dC,

dt /o
where C, is the daughter concentration per gram of
graphite. Integrating this across the power history of

the run, we obtain for the activity (a,) of the daughter
in disintegrations per minute per gram of graphite;

- ?\lclge
PG

'“'?\'ZCQ >

_JGofi (inv.)run .
(a2)o = PG [FOY/massJ ’ (1)
further,
@, (¥) = A(a;), exp(—By) ; (2)

where

J° = gas nuclide flux into graphite at unit power,

(inv.);yp is the activity inventory accumulated by salt
during run,

F® = fission rate at unit power in given mass,
Y = fission yield.

The total accumulation for unit surface integrated
over all depths follows as

(inv.)ryn

2 0
dpm/em” =Hg" 1 Y/mass -

It remains to determine the flux into the graphite, J°
at unit power, by considering conditions within the salt.

The short-lived noble-gas nuclide is generated volu-
metrically in the salt; the characteristic distance
(Dg/N)'/? is about 0.06 cm for a 200-sec nuclide and
about 0.013 cm for a 10-sec nuclide. The salt in a
boundary layer at 0.013 ¢cm from the wall of a fuel
channel with a width of 1 ¢m will flow more slowly
than the bulk salt and will require perhaps 100 sec to
traverse the core. For nuclides of 10 sec or less half-life
and as an approximation for longer-lived nuclides,
Kedl'® showed that such flow terms could be neg-
lected, so that at steady state

7C; MG O
ar¥ " D, D,

3

where Qg is the volumetric generation rate in core salt,
D; is diffusion coefficient in salt, C; is the local con-
centration in salt, and r is the distance from the slab
channel midplane.

Boundary conditions are:
1. at midplane:

dC,q

r=O,dT

2. at either surface:

r=ry, Ji=Jg,
where

Js = flux in salt at surface;

whereby
(dCs) I CeleNDg)'?
“N\dr/ry Ds T D
and
Cs =CeK,
where

K. = Ostwald solubility coefficient .

Integration and satisfaction of the midplane boundary
condition yields:

C, = C cosh (r/MDg) + Q/\ .

The second boundary condition evaluates C:

_ o |
€= K (DsleD)! /2 sinh (ro~/N/Dy) + cosh (ron/NDy)

We may now obtain

1
K, +(eDg/Ds)'/? coth (rov/NDy)

Because coth (rov/N/Ds) ~ 1 and K < (eDg/Ds)'/? ,
Z=DsN'"?
as

0
_ [F Y X salt vol.
mass core vol.

Psait -

QO

83

We now are able to write the desired expression for
the activity of nonmobile daughters of short-lived noble
gases which diffuse into graphite.

1. Surface activity, disintegrations per minute per gram
of graphite:

= (i Psalt
a;), = (inv. XIX——=22
(@), = (v )eun X X 52000
salt vol. eDs\'/?2
x R
core vol. Dg
2. Activity per gram of graphite at any depth, disinte-

grations per minute:

2 (y) =(a2), exp —Py ,
B=(eN/Dg)'
Y12 =(n2)/8.

. Total accumulation for unit surface, integrated over
all depths:

dpm _

_ salt vol. Dy
2

core vol. X Psalt X

(inv.)run X I X
A

cm

Table 9.7 applies these results to the specimens
removed at the end of run 14. A comparison with
observations given earlier in Table 9.3 is commented on
later.

A third possibility of developing activity within the

_graphite is from the traces of uranium found in the

graphite.
Here the relationship is:

As2(U)) = (inv.)yup X I X salt vol.
52 MV-Jrun core vol.

2351 conc. in graphite at given depth

X 235U conc. in 1 g of salt -7

under the assumption that the uranium was present in
the graphite at the given location for essentially all of
the run. The 235U concentration in fuel salt was about
15,500 ppm;

salt vol. 71 ft3 N
corevol. 23 ft3

3.

The concentration at various depths of 233U in a
specimen of CGB graphite is shown in Fig. 9.18. The
surface concentration of about 70 ppm soon falls to
84

102 ORNL-DWG 68-14526 near 1 ppm. Similar data were obtained for all
specimens.
Table 9.8 shows that for '#*Ce, °%Zr, '°3Ru, and
© PY WIDE (1) 1317 at depths greater than about 7 mils the activity
o PY NARROW {11) ) _
®Y-9 could be accounted for as having been produced by the |
. ‘ trace of 235U which was present at that depth. |
. The total quantity of 35U associated with graphite
surfaces was quite small. Total deposition ranged
. between 0.15 and 2.3 ug of 233U per square centi-
O meter, with a median value of about 0.8 ug of 235U per
N\ square centimeter. This is equivalent to less than 2 g of
E \\L}R_ 235U on all the system graphite surface (about 1 on
g \ e flow-channel surfaces) out of an inventory of 75,000 g
2 = \ of 235U in the system.
g .Q\ % 9.3.4 Conclusions from profile data. As an overview
\ of the profile data the following observations appear
Y-9 valid.
‘00 O < \___ * - i
A —— | Let us roughly separate the depths into surface (less
AN —~— Py 1l than 1 mil), subsurface (about to 7 mils), moderately
\\ — deep (about 7 to 20 mils), and deep (over 20 mils).
=t S~ PYL For salt-secking nuclides, !°*Ru and '°Ru and ''1
—o— . —_— the moderately deep and deep regions are a result of
235U penetration and fission. We have noted in an
earlier section that for salt-seeking nuclides the total
» activity for unit surface was in fair agreement with
e o 10 20 20 a0 o recoil effects. Profiles indicate that almost all of the
DISTANCE FROM SURFACE (mils) activity for these nuclides was very near the surface,
Fig. 9.18. Uranium-235 concentration profiles in CGB and consistent with this. For “these the only region for
pyrolytic graphite. which evidence is not clear is the subsurface (1 to 7
Table 9.7. Calculated specimen activity parameters after run 14 based on
diffusion calculations and salt inventory
Chain 89 91 137 140 141
Gas half-life, sec 191 9.8 234 16 1.7
D¢, cm? [sec 1.SE-§ 1.5E -5 1.2E -5 1.2E -5 12E -5
Dg, cm?/sec 14E -5 14E - 5 13E-5 13E -5 1.3E -5
€ 0.1 0.1 0.1 0.1 0.1
Halving depth, mils 56 13 55 14 4.7
Daughter 89gr Wy 137¢5 140p, 141 e
Inventory,? dis/min per gram of salt 1.1E11 1.3E11 4 ,2E94 1.7E11 1.5E11
Surface concentration,? dis/min per gram of graphite 1.2E11 14E11 5.0E9 2.0E11 1.8E11
Total activity,? dis min™ cm™2 4.6E10 1.2E10 1.9E9 1.9E10 5.6E9

%Inventory here is total at end of run 14. Carryover from prior runs is insignificant by the end of run 14 except in the case of
137Cs, where operation prior to end of run 11 contributes about half the inventory at the end of run 14, 319 days later.
b Assumes a local relative flux of 1.
mils), where the values, though declining rapidly, may
be higher than explicable by these mechanisms.

The nuclides having noble-gas precursors (ie., ®°Sr,
140B, 141Ce 91Y) do clearly exhibit the results of
noble-gas diffusion. The slopes and surface concentra-
tions are roughly as estimated. The total disintegrations
per minute per square centimeter is in-accord.

In the case of '37Cs, the levels are considerably too
low in the moderately deep and deep regions, indicating
that cesium was probably somewhat mobile in- the
graphite. Additional evidence on this point is presented
later.

This leaves 5Nb, **Mo, and possibly '32Te and
129MTe, These nuclides appear to have migrated in
graphite, and in particular there is about 10 times as
much niobium as was explicable in terms of the parent
93 Zr present or the 23° U at that depth. Such migration
might occur because the nuclides were volatile fluorides
or because they form stable carbides at this temperature
and some surface diffusion due to metal-carbon chemi-
sorption occurred. The latter possibility, which seems
most likely, seems also to explain the traces of 235U
found having migrated into the graphite.

9.4 Other Findings on Surveillance Specimens

In visual examination of surveillance specimens, slight
amounts of flush salt and, on occasion, dark green fuel
salt were found as small droplets and plates on the
surface of specimens, particularly on faces that had

85

been in contact with other specimens. A brown
dusty-looking film was discerned on about half of the
fuel-exposed surfaces, using a 30-power microscope.
Examination by electron microscopy of a surface film
removed by pressing acetone-dampened cellulose ace-
tate tape against a graphite surface revealed only
graphite diffraction patterns. Later, spectrographic
analysis showed that appreciable quantities of stable
molybdenum isotopes were present on many surfaces.
Presumably these were not crystalline enough to show
electron diffraction patterns.

Thin transverse slices of five specimens were ex-
amined by x radiography. Many of the salt-exposed
surfaces and some other surfaces appeared to have a
thin film of heavy material less than 10 mils thick.

Results of spectrographic analyses of samples from
graphite surface cuts are shown in Table 9.9, expressed
as micrograms of element per square centimeter, Data
for zirconium, lithium, and iron are not included here
since they showed too much scatter to be usefully
interpreted.

About 15 ug of fuel salt would contain 1 ug of
beryllium. Similarly, about 13 ug of Hastelloy N might
contain 1 ug of chromium and also about 9 ug of nickel
and perhaps 2 ug of molybdenum. Thus the spectro-
graphic analysis for beryllium corresponds to about
660 ug of fuel salt per square centimeter. The nickel
analyses correspond to 7-9 ug of Hastelloy N per
square centimeter, and the chromium and molybdenum
(except for a high value) are in at least rough
agreement.

Table 9.8. List of milled cuts from graphite for which the fission product content could be
approximately accounted for by the uranium present

Milled Cut Numbers?

Graphite
5 -

Sample 99Mo 132 129, 103p, 106, 95Ny, 957, 89g; l40g, 141c, 1440, 131
P-77 9,10 6—~10 6,10 4-10 10 3-10
X-13 wide 10 710 7—10 7—10 7--10 7—=10
X-13 narrow 10 5—-10 5-10 5—-10 5—-10 3-10
Y-9 10 10 10
P-58 10 10 10 7—10 7—10 7—10
P-92 9,10 5—10 5—-10 5--10 8—-10 7—-10
K-1 wide 5-7 5=7 5=10 1 1,5-10
K-1 narrow 5-9 5-9 5-9 5-9
PG 1 2 3-10 3—-i10 3-6,8 3-6 2-10 2-10 2-=10 2—-10 2--10 3-10
PG I 10 7,10 2—-10 10 2,8—-10 5-10

Nominally, in mils, cut No. 1 was l/2; 2, I/2; 3,1,4, 2; 5, 3;6,5;7,8; 8 10; 9, 10; 10, 10.

5The samples are listed in order of distance from the bottom of the reactor.
The quantities of molybdenum are too high to
correspond to Hastelloy N composition and strongly
suggest that they are appreciably made up of fission
product molybdenum.

Between the end of run 11 and run ]4 about 4400
effective full-power hours were developed, and MSRE
would contain about 140 g of stable molybdenum
isotopes formed by fission, or about 46 ug per square
centimeter of MSRE surface. The observed median
value of 9 would correspond to a relative deposit
intensity of 0.2, a magnitude quite comparable with the
0.1 median value reported for ?? Mo deposit intensities
on graphite for this set of stringers.

Aliquots of two of these samples were examined mass
spectrographically for molybdenum isotopes. Table

Table 9.9. Spectrographic analyses of graphite
specimens after 32,000 MWhr

Micrograms per Square Centimeter

86

9.10 gives the isotopic composition (stable) for natural
molybdenum, fission product molybdenum (for suit-
able irradiation and cooling periods), and the samples.

This table suggests that the deposits contained com-
parable parts of natural and fission product molyb-
denum. Some preferential deposition of the 95 and 97
chains seems indicated, possibly by stronger deposmon
of niobium precursors.

Determinations of lithium and fluorine penetration
into MSRE graphite reported by Macklin, Gibbons, and
co-workers' 1% were made by inserting samples across
a collimated beam of 2.06-MeV protons, measuring the
'F(p,ay)* 0 intense gamma ray and neutrons from

e "Li(p,n)"Be reaction. Graphite was appropriately
abraded to permit determination of these salt constit-
uents at various depths, up to about 200 mils.

Data for the two specimens — Y-7, removed after run
11, and X-13, removed after run 14 — are shown in
Figs. 9.19-9.25.

These data for each specimen show, plotted logarith-
mically, a decline in concentration of lithium, and

Graphite of Specimen fluorine with depth. Possibly the simplest summary is
Sample Be Mo Cr Ni that, for specimen X-13 removed after run 14, the
lithium, fluorine, and, in. a similar specimen, ?3°U
NR-5W 41 declined in the same pattern. The lithium-to-fluorine
NR-5N 4.6 ratios scattered around that for fuel salt (not that for
PossW 0.457 LiF). The 23%U content, though appearing slightly high
P-77W 1.61 17.6 (it was based on a separate sample), followed the same
P-77N 1.0 34. pattern, indicating no remarkable special concentrating
X-13W 1.1 4.5 effect for this element. The data for sample X-13 might
X-13N 0.7 7.4 0.15 be reproduced if by some mechanism a slight amount of
Y-ow 0.4 9.3 1.03 4.7 fuel salt had migrated into the graphite.
P-S8W 0.4 8.7 0.49 5.5 The pattern for the earlier sample Y-7, removed after
P-goW 2.00 11.3 run 11, is similar with respect to lithium. But the
P-9ON 2.67 fluorine values continue to decrease with depth, so that
Poco-W 0.60 9.0 0.55 6.6 below about 20 mils there is an apparent deficiency of
Poco-N 0.62 10.5 fluorine. Any mechanism supported by this observation
Pyr 1 would appear to require independent migration of
Pyr Il fluorine and lithium.
MR-10W 1.3 It thus appears possible that slight amounts of fuel
MR-10N 41 salt may have migrated into the graphite, and the
presence of uranium (and resultant fission products)
Table 9.10. Percentage isotopic composition of molybdenum on surveillance specimens
92 94 95 96 97 98 100
Natural 15.86 9.12 15.70 16.50 9.45 23.75 9.62
Fission 0 0 2225 0 25-24 25-24 27-26
Samples 4.0;5.0 2.9;3.8 28.6;27.3 4.8;4.9 30.2;29.6 15.1;15.2 14.1;14.4
Redetermination 34;3.3 2.1;1.8 28.9; 30.3 3.9;3.6 32.7,33.5 15.9;14.9 13.1;12.6

87

ORNL-DWG 6B-14 531

oo 2 s g 1000
o FIRST SURFACE TO CENTER
\\ 4 SECOND SURFACE TO CENTER
S0 N e 235y IN SIMILAR SPECIMEN 500
— —
wfi$ - Fln\b
3 o
20 n P W AR 200
Ha-h L{’ J g o,
‘Q\ & hYs f’
s o
‘g * N Li E
a 10 100 &
o [~ %
— u—— —
<
al 235U|>
5 - 50
A
H-E—.
‘\AL—‘\
& 4 r‘ \‘
\
e i Ayl %
7|
1 10
{ 0 100 1000

DISTANCE FROM

SURFACE ({mils)

Fig. 9.19. Lithium concentration as a function of distance from the sutfacg, specimen Y-7.

ORNL-DWG 68-2545R

500
‘ o MEASURING INWARD FROM FIRST SURFACE
Ao MEASURING OUTWARD TOWARD SECOND
SURFACE
200
N
A \
100 \
LY
T 7
o 50 e
o {___Ij
= - A-NYA] CENTER OF
) ~alg SAMPLE
2 20 ! l
€
e
— 10
FT' —-0 %
5
2
i
10 2 5 10 2 5 100 2 5 1000

DEPTH (mils)

!
0.005 004 002 005 04 02 05 10

DEPTH {cm}

Fig. 9.20. Lithium concentration as a function of distance
from the surface, specimen X-13.

ORNL-OWG 68-2544R

1000 r T
Nt o MEASURING INWARD FROM FIRST FHH
500 i . SURFACE T
51T a MEASURING OUTWARD TowarD [
\\; SECOND SURFACE L
200 1
o o \o\‘
& 100 - ¥
S
o]
5 P
—_ oM
E 20 R,
(=% oL 0
— \\
c'r‘»" 10 "
- HHICENTER OF
5 SAMPLE H
T N l
2 AN
1 %\\

[oh] 2 5 10 2 S5 10 2 S5 100 2 5 1000
DEPTH {mils)
1 | 1 ' ! ] 1 1 1 1 1
w32 5 w022 5 w0'2 5 1
DEPTH (cm)

Fig. 9.21. Fluorine concentrations in graphite sample Y-7,
exposed to molten fuel salt in the MSRE for nine months.
Measurements were made as the sample was ground away in
layers progressing from the first surface to the center (open
circles) and then from the interior toward the second surface
(closed triangles). The distances shown are as measured from
the nearest surface exposed to molten salts.
88

ORNL-DWG 68-14530
1000 I I N I T
I O 1
O FIRST SURFACE TO CENTER
A SECOND SURFACE TO CENTER
500 ® 235N SIMILAR SPECIMEN
ffl’zt‘-_"-‘r‘"—"
o \\ \
o S Em
200 . ‘T \;\
A
. y
\, 2 / o
|-c>.-+\ Np\ \\ A .\
o / W "
100 ‘ P 19¢
- W
e %]
< ‘\t; ‘nu Uk"
H o) 0G 19
50 T X fis F
\\ } b ¥
'y \\
T~
\w\ 235,
20 P,
\
® \
10 . -
™~
o
5
q 10 100

DISTANCE FROM SURFACE (mils)

Fig. 9.22. Fluorine concentration as a function of distance from the surface, specimen X-13.

500 ORNL-DOWG 68-14534
| ||
l [
-=0-- SAMPLE Y-7
—e— SAMPLE X-13
200 :
- P
(o] N N
A
100 .
A\
g e
N\
a 50 I N <1
St O .‘...
— M \\
* 3. - 1N|fli'—u\"\“
® ‘f ¢ ot \
\ [ 1)
\
0 \\
~ N pa—————
n) -.’
T
5 |
L 2 5 10 20 50 100 200 500

DISTANCE FROM SURFACE , mils

Fig. 9.23. Comparison of lithium concentrations in samples Y-7 and X-13. -
ORNL-DWG 68-44532

50

R

BEIE—— -

o FIRST SURFACE TO CENTER

e SECOND SURFACE TO CENTER

20

o
®|

RATIO FOR LiF

1 2 5 10 20

50 100 200 500 1000

DISTANCE FROM SURFACE { mils)

Fig. 9.24. Mass concentration ratio, F/Li, vs depth, specimen X-13.

ORNL-DWG 68— 14533

. \ N SAMPLE X—13

N SAMPLE Y-7

(4]
i
W,

AN

N

{ 2 5 10 20 50 100 200 500
DISTANCE FROM SURFACE (mits)

Fig. 9.25. Comparison of fluorine concentrations in samples
Y-7 and X-13, a smooth line having been drawn through the
data points.

within the graphite may largely be explained by this.
Microcracks, where present, would provide a likely
path, as would special clusters of graphite porosity.

References

1. W. H. Cook, “MSRE Material Surveillance Tests,”
MSR Program Semiannu. Progr. Rep. Feb. 28, 1965,
ORNL-3812, pp. 83—86.

2. W. H. Cook, “MSRE Materials Surveillance Test-
ing,” MSR Program Semiannu. Progr. Rep. Aug. 31,
1965, ORNL-3872, p. 87.

3. C. H. Gabbard, Design and Construction of Core
Irradiation-Specimen Array for MSRE Runs 19 and 20,
ORNL-TM-2743 (Dec. 22, 1969).

4. S. S. Kirslis, F. F. Blankenship, and L. L.
Fairchild, “Fission Product Deposition on the Fifth Set
of Graphite and Hastelloy-N Samples from the MSRE
Core,” MSR Program Semiannu. Progr. Rep. Aug. 31,
1970, ORNL4622, pp. 68-70.

5. R. J. Kedl, Fluid Dynamic Studies of the Molten-
Salt Reactor Experiment (MSRE) Core, ORNL-
TM-3229 (Nov. 19, 1970).

6. J. R. Engel and P. N. Haubenreich, Temperatures
in the MSRE Core during Steady-State Power Opera-
tion, ORNL-TM-378 (Nov. 5, 1962).

7. R.H. Perry, C. H. Chilton, and S. D. Kirkpatrick,
eds., Chemical Engineers’ Handbook, 4th ed., McGraw-
Hill Book Co., Inc., New York, 1963, p. 5-21.

8. W. C. Yee, A Study of the Effects of Fission
Fragment Recoils on the Oxidation of Zirconium,
ORNL-2742, Appendix C (April 1960).

9. E. L. Compere and S. S. Kirslis, “Cesium Isotope
Migration in MSRE Graphite,” MSR Program Semi-
annu. Progr. Rep. Aug. 31, 1971, ORNL4728, pp.
51-54.

10. R. B. Evans III, J. L. Rutherford, and R. B.
Perez, “Recoil of Fission Products, II. In Heterogeneous
Carbon Structures,” J. Appl Phys. 39, 3253-67
(1968).

11. A. P. Malinauskas, J. L. Rutherford, and R. B.
Evans I, Gas Transport in MSRE Moderator Graphite,
I. Review of Theory and Counter Diffusion Experi-
ments, ORNL-4148 (September 1967).

12. R. B. ‘Evans III, J. L. Rutherford, and A. P.
Malinauskas, Gas Transport in MSRE Moderator
Graphite, II. Effects of Impregnation, IllI. Variation of
Flow Properties, ORNL-4389 (May 1969).

13. R. J. Kedl, 4 Model for Computing the Migration
of Very Short-Lived Noble Gases into MSRE Graphite,
ORNL-TM-1810 (July 1967).

14. R. L. Macklin, J. H. Gibbons, E. R1cc1 T. H.
Handley, and D. R. Cuneo, “Proton Reaction Analysis
for Lithium and Fluorine in MSR Graphite,” MSR

90

Program Semiannu. Progr. Rep. Feb. 29, 1968, ORNL-
4254, pp. 119-27.

15. R. L. Macklin, J. H. Gibbons, and T. H. Handley,
Proton Reaction Analysis for Lithium and Fluorine in
Graphite, Using a Slit-Scanning Technique, ORNL-
TM-2238 (July 1968).

16. R. L. Macklin, J. H. Gibbons, F. F. Blankenship,
E. Ricci, T. H. Handley, and D. R. Cuneo, ‘“Analysis of
MSRE Graphite Sample X-13 for Fluorine and
Lithium,” MSR Program Semiannu. Progr. Rep. Aug.
31, 1968, ORNL-4344, pp. 146--50.
91

10. EXAMINATION OF OFF-GAS SYSTEM COMPONENTS OR SPECIMENS
REMOVED PRIOR TO FINAL SHUTDOWN

The off-gas from the pump bowl carried some salt
mist, gaseous fission products, and oil vapors. On a few
occasions, restrictions to flow developed, and in replac-
ing the component or removing the plugging substance,
samples could be obtained which provide some insight
into the nature of the fission product burden of the
off-gas. In addition, a special set of specimens was
installed in the “jumper line” flange near the pump
bowl after run 14 and was examined after run 18.

10.1 Examination of Particle Trap
Removed after Run 7

The Mark I particle trap,' which replaced the filter in
off-gas line 522 just upstream of the reactor pressure
control valve, was installed in April 1966, following
plugging difficulties experienced in February and March
1966. It was replaced by one of similar design in
September 1966, permitting its examination. The ori-
ginal plugging problems were attributed to polymeriza-
tion of oil vapors originating in the entry into the pump
bowl of a few grams of lubricating oil per day.

The particle trap accepted off-gas about 1 hr flow
downstream from the pump bowl. Figure 10.1 shows
the arrangement of materials in the trap. The incoming
stream impinged on stainless steel mesh and then passed
through coarse (1.4 1) and fine (0.1 u) felt metal filters.
The stream then passed through a bed of Fiberfrax and
finally out into a separate charcoal bed before continu-
ing down the off-gas lines to the main charcoal beds.

FINE METALLIC FILTER

Black deposits were found on the Yorkmesh at the
end of the entry pipe, as seen in Fig. 10.2. The mesh
metal was heavily carburized, indicating operating
temperatures of at least 1200°F (the gas stream at this
point was much cooler). The radiation level in some
parts of this region was about 10,000 R/hr for a probe
in the inlet tube.

In addition to an undetermined amount of metal
mesh wire, a sample of the matted deposit showed 35%
weight loss on heating to 600°C (organic vapor), with a
carbon content of 9%.

Mass spectrographic analysis showed 20 wt % Ba, 15
wt % Sr, 0.2 wt % Y, and only 0.01 wt % Be and 0.05
wt % Zr, indicating that much of the deposit was
daughters of noble-gas fission products and relatively
little was entrained salt.

Gamma-ray spectrometry indicated the presence of
137CS, 89Sr’ 1('.)3}{u or 106_RU, llOmAg’ 95Nb’ and
14014, Lack of quantitative data precludes a detailed
consideration of mechanisms. However, much of the
deposit appears to be the polymerization products of
oil. Salt mist was in this case largely absent, and the
fission products listed above are daughters of noble
gases or are noble metals such as were found deposited
on specimens inserted in the pump bowl. One consist-
ent model might be the collection of the noble-metal
nuclides on carbonaceous material (soot?) entrained in
the pump bowl in the fuel salt and discharged from
there into the purge gas; such a soot could also adsorb
the daughters of the noble gases as it existed as an

ORNL-DWG 66-11444R

FIBERFRAX
COARSE METALLIC FILTER (LONG FIBER) - THERMOCOUPLE (TYP)
LOWER WELD~ SAMPLE 9] "N — g
AR N ke o
7 _— \ —"i T =
3 : N i
=) | ! WA 2
i - | _ ———
N | g ) NI G b s T N\
'S = e s
e L
— SAMPLES A= LoampLe 3A -8B
58 AND &
STAINLESS STEEL MESH NICKEL BAFFLE (8)

Fig. 10.1. MSRE off-gas particle trap.
§ -

92

. fl;ff

Fig. 10.2. Deposits in particle trap Yorkmesh.

aerosol in the off-gas. Particles of appropriate size,
density, and charge could remain gas-borne but be
removed by impingement on an oily metal sponge.
Although the filtering efficiency was progressively
better as the gas proceeded through the trap, by far the
greatest activity was indicated to be in the impingement
deposit, indicating that most of the nongas activity
reaching this point was accumulated there (equivalent
to about 1000 full-power hours of reactor operation)
and that the impinging aerosol had good collecting
power for the daughters of the noble gases. The aerosol
would have to be fairly stable to reach this point
without depositing on walls, which implies certain
limits as to size and charge. Evidently the amounts of
noble metals carried must be much less than the
amounts of daughters of noble gases (barium, stron-
tium) formed after leaving the pump bowl. Barium-138

(about 6% chain yield: 17-min Xe - 32-min Cs ~ stable
Ba) comprises most of the long-lived stable barium.
Thus it appears that if the amounts of noble metals
detected by mass spectrometry were small enough,
relative to barium, to be unreported, the proportions of
noble metals borne by off-gas must be small, although
real. No difficulties were experienced with the particle
trap inserted in September 1966, and it has not been
removed from the system.

10.2 Examination of Off-Gas Jumper
Line Removed after Run 14

After the shutdown of MSRE run 14, a section of
off-gas line, the jumper section of line 522, was
removed for examination.? This line, a 3-ft section of
Y,-in.-ID open convolution flexible hose fabricated of

type 304 stainless steel with O-ring flanges on each end,
was located about 2 ft downstream from the pump
bowl. The upstream flange was a side-entering flange
which attaches to the vertical line leaving the pump
bowl, while the downstream flange was a top-entering
flange which attaches to the widened holdup line. The
jumper discussed here, the third used in the MSRE, was
installed prior to run 10, which began in December
1966.

After the shutdown, the jumper section was trans-
ferred to the High Radiation Level Examination Labo-
ratory for cutup and examination. Two flexible-shaft
tools used to probe the line leading to the pump bow!l
were also obtained for examination. On opening the
container an appreciable rise in hot-cell off-gas activity
was noted, most of it passing the cell filters and being
retained by the building charcoal trap. As the jumper
section was removed from the container and placed on
fresh blotter paper, some dust fell from the upstream
flange. This dust was recovered, and a possibly larger
amount was obtained from the flange face using a
camel’s-hair brush. The powder locked like soot; it fell
but drifted somewhat in the moving air of the cell, as if
it were a heavy dust. A sample weighing approximately
8 mg read about 80 R/hr at “contact.” The downstream
flange was tapped and brushed over a sheet of paper,
and similar quantities of black powder were obtained
from it. A sample weighing approximately 15 mg read
about 200 R/hr at contact. Chemical and radiochemical
analyses of these dust samples are given later.

The flanges each appeared to have a smooth, dull-
black film remaining on them but no other deposits of
significance (Fig. 10.3). Some unidentified bright flecks
were seen in or on the surface of the upstream flange.
Where the black film was gently scratched, bright metal
showed through.

Short sections of the jumper-line hose (Fig. 10.4)
were taken from each end, examined microscopically,
and submitted for chemical and radiochemical analysis.
Except for rather thin, dull-black films, which smoothly
covered all surfaces including the convolutions, no
deposits, attack, or other effects were seen.

Each of the flexible probe tools was observed to be
covered with blackish, pasty, granular material (Fig.
10.5). This material was identified by x ray as fuel-salt
particles. Chemical and radiochemical analyses of the
tools will be presented later.

Electron microscope photographs (Fig. 10.6) taken of
the upstream dust showed relatively solid particles of
the order of a micron or more in dimension, surrounded
by a material of lighter and different structure which
appeared to be amorphous carbon; electron diffraction
lines for graphite were not evident.

93

An upstream 1-in. section of the jumper line near the
flange read 150 R/hr at contact; a similar section near
the downstream flange read 350 R/hr.

10.2.1 Chemical analysis. Portions of the upstream
and downstream powders were analyzed chemically for
carbon and spectrographically for lithium, berylium,
zirconium, and other cations. In addition, 233U was
determined by neutron activation analysis; this could be
converted to total uranium by using the enrichment of
the uranium in the MSRE fue] salt, which was about
33%.

Results of these analyses are shown in Table 10.1.

Analyses of the dust samples show 12 to 16% carbon,
28 to 54% fuel salt, and 4% structural metals. Based on
activity data, fission products could have amounted to
2 to 3% of the sample weight. Thereby 53 to 22% of
the sample weight was not accounted for in these
categories or spectrographically as other metals. The
discrepancies may have resulted from the small amounts
of sample available. The sample did not lose weight
under a heat lamp and thus did not contain readily
volatile substances.

10.2.2 Radiochemical analysis. Radiochemical analy-
ses were obtained for the noble-metal isotopes '!! Ag,
106 Ry, 23Ry, ?Mo, and °5Nb; for ®%Zr; for the rare
earths '*7Nd, '%?Ce, and !%'Ce; for the daughters of
the fission gases krypton and xenon: °'Y, 8°8r, ®°8i,
199Ba, and '37Cs; and for the tellurium isotopes
132Te, '2°Te, and *2'I (tellurium daughter). These
analyses were obtained on samples of dust from
upstream and downstream flanges, on the approxi-
mately 1-in. sections of flexible hose cut near the
flanges, and on the first flexible-shaft tool used to
probe the pump off-gas exit line.

Data obtained in the examination are shown in Table
10.2, along with ratios to inventory.

It appeared reasonable to compare the dust recovered
from the upstream or downstream regions with the
deposited material on the hose in that region; this was
done for each substance by dividing the amount
deposited by the amount found in 1 g of the associated
dust.

Values for the inlet region were reasonably consistent
for all classes of nuclides, indicating that the deposits
could be regarded as deposited dust. The median value
of about 0.004 g/cm indicates that the deposits in the
inlet region corresponded to this amount of dust. A
similar argument may be made with respect to the
downstream hose and outlet dust, which appeared to be
of about the same material, with the median indicating
about 0.016 g/cm. Ratios of outlet and inlet dust values
had a median of 1.5, indicating no great difference
between the two dust samples.
94

PHOTO 1856 - 74

Fig. 10.3. Deposits on jumper line flanges after run 14.

95

R-42973

Fig. 10.4. Sections of off-gas jumper line flexible tubing and outlet tube after run 14.
Table 10.1. Analysis of dust from MSRE off-gas jumper line

Inlet Flange (wt %) Outlet Flange (wt %)

As Determined Constituent As Determined Constituent

L1 3.4 7.3

LiF 12.6 27.1
Be 1.7 4.2

BeF2 8.9 21.9
Zr 2.74 1.4

Zer 5.0 2.6
235y 0.358 0.596

UF[_ {total) 1.4 2.4
Carbon 1014 1517

12 16
Fe 3 2
Cr ~C0.01
Ni ~] 1 (4.5)
Mo ~0.4 0.4
Al ™1 1
Cu 0.1 G.1
FP’s (max)® (™~ (™3
Total found 47 78

Assumes chain deposition rate constant throughout power history (includes only chains detemined).
Table 10.2. Relative quantities of elements and isotopes found in off-gas jumper line®

b

Sample Inlet Dust Outlet Dust Upstream Hose Downstream Hose Flexible Tool MSRE Inventory
Element or . Yield® (per g) (per g) (per ft) (per ft) (total) Decay Rate
Isotope 1/2 (%) (1016 dis/min)
Element
Li 0.066 0.14 0.015 0.027 0.066
Be 0.057 0.14 0.006 0.010 0.031
Zr 0.054 0.027 0.004 0.004 0.00003
235y 0.050 0.083 0.009 0.023 0.0014
e 23 32
Isotope
1llag 7.6 d 0.0181 ~31 47 ~1 ~28 0.41 0.235
106py 365 d 0.392 10 33 21 125 6.9 2.53
103y 39.7 d 2,98 5.7 11 3.9 40 1.3 32.4
?9Mo 67 hr  6.07 2.8 88 2.9 48 1.6 88.5
95Nb 35 6,26 0.54 1.2 0.047 0.065 61.4
Szr 65 6.26 ~0.013 0.025 ~0,0003 <0.009 <0.0002 65.3
147Ng 11.1 d 2.37 <0.06 <0.02 <0.003 <0.02 <0.0009 20.3
144ce 285 d 5.58 ~0.027 0.039 ~0.001 ~0.01 ~0.0007 40.8
14lce 33 d 6.44 0.0004 0.008 ~(.0003 ~0,001 <0.0001 71.1
ly 58 5.83 2.5 (90) 7.0 (250) 0.32 (12) 1.4 (50) 0.06 (2.0) 61.5 (1.69)
140p, 12.8 d 6.39 3.6 (140) 1.8 (66) 0.75 (28) 3.5 (130) 0.13 (5.0) 77.6 (2.05)
89¢r 50.5 yr  4.72 120 (260) 150 (320) 13 (29) 71 (150) 0.15 (0.33) 50.4 (23.2)
137¢s 20.2 yr  6.03 150  (300) 110 (210) 13 (26) 62 (120) 1.5 (3.0) 2.17 (1.12)
90g, 28 yr  5.72 3.6 (30) 28 (230) 3.0 (25) 13 (110) 2.14 (0.256)
1321e 78  hr 421 8.7 14 1.1 9.0 0.29 60.6
129mTe 37 d 0.159 28 61 4.6 30 1.7 1.74
131 8.05 d 2,93 6.4 2.5 0.9 3.2 0.27 37.7

®Ratio of amount found in sample to 10~% x MSRE inventory., The fission product inventory was computed from the power history since startup, as-

suming full power equals 8 Mw,

bValues in parentheses are corrected for fraction of rare-gas precursor entering pump bowl, assuming 100% stripping and negligible return in stripped

salt,

“Taken from Nuclear Data Library for the Fission Product Program by M. R. Trammell and W. A. Hennenger (Westinghouse Astro-Nuclear Laboratory,
Pittsburgh, Pa.), WANL-TME-574 (rev. 1), Nov, 17, 1966. Independent yields of chain members are given, and all yields normalized to 200%; vield for

129mpe differs from other published values,

L6
Fig. 10.5. Deposit on flexible probe.

98

The electron microscope photographs of dust showed
a number of fragments 1 to 4 u in width with sharp
edges and many small pieces 0.1 to 0.3 u in width
(1000 A to 3000 A).

The inventory values used in these calculations
represent accumulations over the entire power history
(2 — to); the more appropriate value would of course
be for the operating period (¢, — ¢, ) only:

fry—to=le,—t, T~y eXP [-A(2 - 41)] -

The second term, representing the effect of prior

accumulation, is important only when values of A(#, —

t;) are suitably low (less than 1). So, except for

366-day '°6Ru, 2-year '25Sb, 30-year ?°Sr, and

30-year ' 27 Cs, correction is not particularly significant.
We shall frequently use the approach

obs __ obs inv. (total)
inv. (present period) inv. (total) ~ inv. (present period)’

where

inv. (total) _ It 1
inv. (present period) [y, ¢o — 1, ¢, exp [-M(t; - 1,)]

_ 1
V= [t —to/lty—14]) exp [-N(t2 — £1)]

If the prior inventory were relatively small,

][1 - fo/ft2 — l'0<1 >

or the present period relatively long with respect to
half-life, A(z, — #;) > 1, then the ratio

1total/1nv'(present period)

approaches 1. It can never exceed the ratio

El:PHtotal/EFPI_I(total after prior period) -

(EFPH is accumulated effective full-power hours.)
Examination of the data in terms of mechanisms will
be done later in the section.

10.3 Examination of Material Recovered
from Off-Gas Line after Run 16

At the end of run 16 a restriction existed in the
off-gas line (line 522) near the pump, which had
developed since the line was reamed after run 14.2°2 To
100

clear the line and recover some of the material for
examination, a reaming tool with a hollow core was
attached to flexible metal tubing. This was attached to
a “May pack” case and thence to a vacuum pump
vented into the off-gas system. The May pack case held
several screens of varied aperture and a filter paper. The
specialized absorbers normally a part of the May pack
assembly were not used.

The tool satisfactorily opened the off-gas line. A small
amount of blackish dust was recovered on the filter
paper and from the flexible tubing.

Analysis of the residue on the filter paper is shown in
Table 10.3. The total amount of each element or
isotope was determined and compared with the amount
of “inventory” fuel salt that should contain or had
produced such a value.

The constituent elements of the fuel salt appear to be
present in quantities indicating 4 to 7 mg of fuel salt on
the filter paper, as do the isotopes 1*°Ba, '%*Ce, and
®5Zr, which usually remain with the salt. It is note-
worthy that *33U is in this group, indicating that it was
transported only as a salt constituent and that the salt
was largely from runs 15 and 16.

Table 10.3. Material recovered from MSRE off-gas
line after run 16

Corrected to shutdown December 16, 1968

Inventory Found
(per milligram : -
of salt) Total Ratio to inventory
Elements
In milligrams
Li 0.116 0.80 7
Be 0.067 0.35 5
Zr 0.116 0.47 4
U-233 0.0067 0.0396 6
Fission products
In disintegrations per minute
Sr-89 2.9E6 3.13E8 110
Cs-137 4.1E6 4.31E8 110
Ba-140 4.1E6 1.77E6 8
Y-91 5.2E6 1.2E8 23
Ce-144 4.1E7 1.48E8 4
Zr-95 7.4E6 3.08E7 4
Nb-95 9.4E6” 1.40E9 150“
Mo-99 3.1E4 2.76E7 900
Ru-106 2.8E6 (9.8E2% 3.51E6 1400¢
Te-129m 2.3E4 9.2E7 4000
I-131 9.8E4 1.01E6 10

@Inventory set to zero for fuel returned from reprocessing
September 1968.

The isotopes #%Sr and *37Cs, which have noble-gas
precursors with half-lives of 3 to 4 min, are present in
significantly greater proportions, consistent with a
mode of transport other than by salt particles.

The “noble-metal” isotopes * SNb, **Mo, 1°¢Ru, and
129MTe were present in even greater proportions,
indicating that they were transported more vigorously
than fuel salt. Comparison with inventory is straight-
forward in the case of 2.79-day °°Mo and 34-day
129mTe  since much of the inventory was formed in
runs 15 and 16. In the case of 367-day ! °% Ru, although
a major part of the run 14 material remains undecayed,
salt samples during runs 15 and 16 show little to be
present in the salt; if only the '°®Ru produced by
2330 fission is taken into account, the relative sample
value is high.

The fuel processing, completed September 7, 1968,
appeared also to have removed substantially all ®5Nb
from the salt. Inventory is consequently taken as that
produced by decay of ®5Zr from run 14 after this time
and that produced in runs 15 and 16. Thus the
“noble-metal” elements appear to be present in the
material removed from the off-gas line in considerably
greater proportion than other materials. It would
appear that they had a mode of transport different
from the first two groups above, though they may not
have been transported all in the same way.

There remains 8.05-day *3*1. The examination after
run 14 of the jumper section of the off-gas line found
appreciable [, which may have been transferred as
30-hr '3 ™Te. In the present case, essentially all the
1311 inventory came from a short period of high power
near the end of run 15. Near-inventory values were
found in salt sample FP 16-4, taken just prior to the
end of run 16. Thus it appears that the value found here
indicates little * 3?1 transferred except as salt.

10.4 Off-Gas Line Examinations after Run 18

After run 18 the specimen holder installed after run
14 in the jumper line outlet flange was removed, and
samples were obtained.

The off-gas specimens were exposed during 5818 hr
to about 4748 hr with fuel circulation, during runs 15,
16, 17, and 18. During this period, 2542 effective
full-power hours were developed.

Near the end of run 18, a plugging of the off-gas line
(at the pump bowl) developed. Restriction of this flow
caused diversion of off-gas through the overflow tank,
thence via line 523 to a pressure control valve assembly,
and then into the 4-in. piping of line 522. These valves
can be closed when it is desired to blow the accumu-
lated overflow salt back into the pump bowl, but are
normally open. A flow restriction also developed in line
523 near the end of run 18, and the valve assembly was
removed for examination.

Data obtained from both sets of examinations will be
described below.

The off-gas line specimen assembly was placed after
run 14 in the flange connecting the jumper line exit to
the entry pipe leading to the 4-in. pipe section of line
522. The specimen holder was made of 27 in. of
Y,-in.-OD, 0.035-in.-wall stainless steel tubing, with a
flange insert disk on the upper end. Four slotted
sections and one unslotted section occupied the bottom
17 in. of the tube; about 8 of the upper 10 in. were
contained within the % -in. entry pipe; all the rest of
the specimen holder tube projected downward into the
4-in. pipe section. The specimen arrays included a
holder for electron microscope screens, a pair of
closed-end diffusion tubes, and a graphite specimen.

A hot-cell photograph of the partially segmented tube
after exposure is shown in Fig. 10.7. The electron
microscope screens were not recovered. Data from the
diffusion tubes and graphite specimen will be presented
below. In addition, two sections were cut from the
10-in. unstotted section, of particular interest because
normally all the off-gas flow passed through this tubing.
These segments of tubing were then plugged, and the

101

exterior was carefully cleaned and leached repeatedly
until the leach activity was quite low; the sections were
then dissolved and the activity determined. Data from
these sections are presented in Table 10.4.

From the exterior of the tube, slightly above the
upper slots, a thin black flake of deposited material was
recovered weighing about 20 mg. The tube, after
removal of the flake, is shown in Fig. 10.8. The
underlying metal was bright and did not appear to have
interacted with the flake substance. Analysis of the
flake is shown in Table 10.4.

The quantity per centimeter was divided by that for 1
g of flake substance for each nuclide; the general
agreement of values indicated that they were doubtless
from the same source and that the deposit intensities on
the two sections were about 1 and 6 mg/cm respectively
(based on median ratio values). These values will be
referred to later. Observed values are also shown for
deposits on the upstream handling “bail.”

Data were also obtained for fission product deposi-
tion on the graphite specimen (narrow and wide faces)
and on consecutive dissolved 1-in. scrubbed segments of
Ys-in.- and Y;-in.-ID diffusion tubes closed on the upper
ends. The upper parts of these tubes contained packed
sections of the granular absorbents Al, O3 and NaF.

Data on these specimens are shown in Table 10.5. As
useful models for examination of the data have not

Fig. 10.7. Off-gas line specimen holder as segmented after removal, following run 18.
Table 10.4. Data on samples or segments from off-gas line specimen holder removed following run 18

Inventory? Flake Tube section 1 Tube section 2
Total upstream
per gram Amount  Ratio to inventory  Amount per  Ratio to MSRE Ratio to Amount per  Ratio to MSRE Ratio to “bail” deposit
of salt per gram for 1 g of salt centimeter total inventory 1 g of flake centimeter total inventory 1 g of flake
Elements
In milligrams
Li 113 2.89 0.026 <0.016 3.2E-11 <0.0054 0.019 3.7E-11 0.00642 0.20
Be 67.6 7.11 0.105 ~(.008 2.7E—11 ~0.0011 ~0.012 4.1E-11 ~0.0011 0.92
Zr ¢
U-233 6.7 0.517 0.078 0.14
Fission products
In disintegrations per minute
Sr-89 1.332E11 1.8E12 13 1.0E12 1.7E-6 0.011 1.3E12 2.2E-6 0.014 4.0E12
Y91 1.206E11 8.8E9 0.07 1.5E10 2.8E-8 0.033 1.9E10 3.5E--8 0.041 3.0E9
Ba-140  1.534El1 6.4E9 0.04 1.5E10 2.2E-8 0.046 2.4E10 3.5E-8 0.073
Cs-137  5.450E9 2.8E10 5 8.3E9 3.5E-7 0.0058 1.0E10 4.1E-1 0.0069 3.6E10
Ce-141  2E11 8.0E6 0.00004 1.9E7
Ce-144  6.109E10 S4E7 0.0009 3.3E8
Nd-147 ~5E11 1.6E9 0.03
Zr-95 1.254E11 1.6E8 0.0012 1.0E7 1.9E-11 0.0013% 2.0E7 3.7E-11 0.0025
Nb-95 8.920E10 1.4E11 1.6 4.7E9 1.2E-8 0.0007 1.0E10 2.6E-8 0.0014 2.6E10
Ru-103  4.495E10 1.2E11 2.7 4.3E9 2.2E-8 0.0007
Ru-106 3.464E10 2.5E10 7 7.9E8 5.2E-9 0.0006
Te-129  1.872E10 2.3E10 1.2 1.2E9 1.4E-8 0.0010 1.9E9 2.3E-8 0.0016
1-131 8.0E10 1.5E12 4.3E-6 7.9E9 2.3E-8

@This inventory covers the entire MSRE operating history ; as usual, however, 235 Nb is taken as zero at start of run 15.
bMedian.

A1) |
Table 10.5. Specimens exposed in MSRE off-gas line, runs 15—-18

Inventory for g]r;jfi?t?;g;gn?:n Radioactivity on Y-in.-ID diffusion tube (dis/min) Radioactivity on Y-in.-ID diffusion tube (dis/min)

Nuclide 1 g of salt (dis min~! ¢cm~?) Bottom Second Third Fourth  Al,04 NaF Second Third Fourth  Al;0; NaF

(dis min~! g71) . inch inch inch inch granules  granules inch inch inch granules  granules

Wide face  Narrow face

Sr-89 1.3E11 3.6E11 4 4E10
Y-91 1.2E11 1.1E9 4 4E7
Ba-140 1.5E11 3.5E10 7.9E6 1.1E7 2.5E7 1.5E8 <3E8 1.5E8 7.0E8 3.5E8 1.6E8 2.8E10 2.6E9
Cs-137 5.4E9 4.4E9 6.2E8 3.3E7 2.4E7 14E8 1.3E9 6.5E8 4.6E8 8.0E8 2.7E8 5.8E9 7.0E9 1.1E10
Ce-144 6.1E10 1.2E6 5.2ES 2.7ES 1.9E5 3.4E6 29E4 <1Eé6 <1E6 <4E4 <SES <2E6 <2.4E6 1.5E7
Zr-95 1.3E11 5.3E6 6.7ES 1.1E5 <SE4 <lES5 <1ES 2.1E5 <4Es 1.3E6 <6ES  <3E6 <1E6 <3E6
Nb-95 8.9E10 2.8E7 a 7.9E5 4.6E6 1.0E6 1.9E9 1.9E9 4 4E7 i4E8 1.8E7 3.0E7 <8Eé6
Ru-103 4.5E10 1.5E7 1.1E6 1.9E6 1.5E6 7.9ES5 S5.7ES 1.2E6 1.9E6 3.5E8 6.6E7 <3E6 <1E7 <6E6
Ru-106 3.5E9 3.5E7 1.3E7 2.7E6 1.2E6 1.1E6 1.1E6 5.2E6 4.7E6 26E7 64E6 S5.7E6 <8E6 <4E6
Sb-125 2.9E8 8.5E4 3.7ES
Te-129 1.9E10 3.3E7 3.0E8

Negligible.

€01
104

R-48664

Fig. 10.8.

been found, these data will not be considered further
here.

10.5 Examination of Valve Assembly
from Line 523 after Run 18

The overflow line from the pump bowl opens at
about the spray baffle, descends vertically, spirals
around the suction line below the pump bowl, then
passes through the top of the toroidal overflow tank,
and terminates near the bottom of the tank. Gas may
pass through this line if no salt covers the exit into the
overflow tank or if the pressure is sufficient (due to
clogging of regular off-gas lines, etc.) to cause bubbling
through any salt that is present.

In addition, helium (about 0.7 std liter/min) flows in
through two bubblers to measure the liquid level. Gas
from these bubblers, along with any off-gas flow, will
pass through line 523 to enter the 4-in.-diam part of the
main off-gas line 522.

From time to time the salt accumulated in the
overflow tank was returned to the pump bowl. This was
accomplished by closing a control valve in line 523
between the overflow tank and the entry point into line
522. The entry of the bubbler gas pressurized the
overflow tank until return of the accumulated sait to
the pump bowl permitted the gas to pass into the pump
bowl; the control valve was then reopened.

Section of off-gas line specimen holder showing flaked deposit, removed after run 18.

Near the end of run 18, clogging of line 523, which
had been carrying most of the off-gas flow for several
weeks, was experienced. This was attributed to plugging
in the valve assembly, and after shutdown the assembly
was replaced. The removed equipment was transported
to the High Radiation Level Examination Laboratory,
where it was segmented for examination.

The vertical %-in. entry and exit lines to the valve
assembly were spaced about 39 in. apart, terminating in
O-ring flanges. Following the entry flange the line
continued upward, across, and downward to the entry
port of the automatic control valve HCV-523. Gas flow,
proceeded past the plunger into the cylinder containing
the bellows plunger seal and upward into L-shaped exit
holes in the flange, leaving the valve flange horizontally.
Curved piping then led to a manual valve (V-523),
normally open, which was entered upward. The hori-
zontal exit pipe from this valve curved upward and then
downward to the exit flange.

Examination of HCV-523 did not reveal an obstruc-
tion to flow. All surfaces were covered with sootlike
deposits over bright metal; samples of this were
recovered. Some samples of black dust were obtained
from sections of the line between the two valves.

In the normally open V-523 the exit flow was
downward. A blob of rounded black substance (about
the size of a small pea) was found covering the exit
105

Table 10.6. Analysis of deposits from line 523 (undetermined quantity)

Expressed as equivalent grams of inventory salt

Ratio of observed value to that for 1 g of inventory salt

Inventory ;
for1g HCV-523 Line
of fuel salt Leach Dust Dust Dust Dust Dust V-523
Leach
Elements
In milligrams
Li 116 0.080 0.025 0.006 <0.002 <0.0005 0.009
Be 67 1.5 14 0.036 0.003 0.033
U 6.7 0.54 0.12 0.23 0.12 0.12 0.013 0.095
(% U-235) (€10)] 38 (24) (44) 29
(% U-238) 62) 59) (61) - (54) 61)
Fission products
In disintegrations per minute

Sr-89 1.3E11 12 0.019 0.065 0.004 0.28 0.35 5.0
Sr-90 5.4E9 230 160 17
Y-91 1.2E11 <(.006 <0.005
Ba-140 1.5E11 0.039 0.21 0.043 0.038 0.002 0.011 0.022
Cs-137 5.5E9 21 i3 41 1.2 0.31 0.33 94
Ce-141 1.9E11 0.001 0.002
Ce-144 6.1E10 0.009 0.013
Nd-147 5.6E10 <0.05 0.44 0.036 0.009 0.010 0.002 <0.04
Zr-95 1.3E11 0.004 0.40 0.017 0.006 0.0003 0.0015 0.006
Nb-95 8.9E10 7.5 430 27 5.2 6.4 9.2
Ru-103 4.5E10 14 0.57 0.18 0.13 4.6 0.036 13
Ru-106 3.5E9 18 .56 0.19 0.13 4.3 0.041 20
Ag-111 6.7E8 <30 300 24 160 0.78 19 <20
Sb-125 2.9E8 56 82
Te-129 8.3E9 17 700 717 28 4.6 6.3 15
[-131 8.1E10 1.3 38 15 4.9 0.036 14 - 1.5

aperture, in addition to a hard black deposit on the
hexagonal side of the entry port.

The black material which covered the V-523 port was
glossy, quite hard and frangible, and filled with bubbles
1 mm in diameter. Examination by W. W. Parkinson?* is
summarized.

The plug material analyzed 44% carbon, 0.3% beryl-
lium, and 0.1% lithium.

Leaching by CCl; at room temperature yielded a
solution indicated by its infrared spectrum to contain a
saturated hydrocarbon having considerable branching.
The leach residue, comprising over two-thirds of the
original sample, in all probability was a saturated
cross-linked and insoluble hydrocarbon. The source of
this material was believed to be a hydrocarbon (lubri-
cant or solvent) which evaporated or decomposed at
elevated temperature with vapors being carried to the
valve where they condensed. Condensation was thought
to be followed by cross-linking, probably radiation-
induced, to render the material insoluble.

The off-gas service of line 523 and its valve assembly
was sporadic over the full prior history of the reactor.
A good account of the amount and duration of the flow
is not available, and so a detailed examination of data in
terms of mechanisms related to such history will not be
attempted.

The total quantities of the various elements and
nuclides in the samples from each valve and the
intervening line are shown in Table 10.6, where they are
expressed as fractions of the inventory for 1 g of salt.

In spite of unknown mass and uncertain deposition
schedule, several conclusions are evident. The uranium
was largely 235U and 232U, doubtless deposited during
that phase of operations. This is substantiated by high
values of ?°Sr and '37Cs (both long-lived) compared
with 50.4-day 3°Sr, indicating that the average accumu-
lation rate was higher than the recent.

The noble metals generally were relatively high; *SNb
far exceeded °5Zr, and the two ruthenium isotopes
(*°3Ru and '°®Ru) were consistent with each other
and appreciably higher than the salt-seeking substances
(Li, Be, °5Zr, 1%1Ce, '%%Ce, 147Nd, '*°Ba, etc.).
Silver-111 was quite high! Antimony-125 and '2°™Te
were quite high. Todine-131 was lower than '2°™Te in
each sample. On balance the pattern is not much
different than that seen for other deposits from the gas
phase, wherever obtained. The absolute quantities were
not really very large, and no analysis would have been
called for had not the plugging that developed during
the use of this line for off-gas flow required removal.

10.6 The Estimation of Flowing Aerosol
Concentrations from Deposits on Conduit Walls

We can observe the amounts of activity or mass
deposited on off-gas line surfaces. To find out what this
can tell us about MSRE off-gas behavior, we must
consider the relevant mechanics of aerosol deposition.
We will obtain a relation between the amount enter-
ing a conduit and the amount depositing on a given
surface segment by cither diffusional or thermophoretic
mechanisms. The accumulation of such deposits over
extended  operating periods will then be related to
observable mass or activity in terms of inventory values,
fraction to off-gas, etc.

Diffusion coefficients of aerosol particles are cal-
culated using the Einstein-Stokes-Cunningham equa-
tion,’

A=1.254+044e—1.097/1

106

At a pressure of 5 psig and assuming a helium
collision diameter (¢) of 2.2 A, the mean free path is
calculated from the usual formula:

1

m(2)1/2 pg? ’

)

where n is the number of atoms per cubic centimeter.

Values calculated for the diffusion coefficient of
particles of various diameters in S psig of helium are
shown in Table 10.7.

10.6.1 Deposition by diffusion. The deposition of
aerosols on conduit walls from an isothermal gas
stream in laminar flow is the result of particle
diffusion and follows the Townsend equation,”*® which
is of the form

Nn=Hng X a; eXp [_th/(Q/D)] )

where n is the concentration of gas-borne particles at a
distance X from the entrance and Q is the volumetric
flow rate. The coefficients a; and b; are numerical
calculated constants.

The derivative gives the amount deposited on unit
length (n,) in terms of the amount entering the conduit,
Hy:

D
0

Values of the coefficients are:

ng=—dnjdz = ny — X a;b; exp [-b,X/(Q/D)] .

i a; b;
where [/ is the mean free path, r the particle radius, and
. . 1 0.819 11.488
n the viscosity of the gas. 5 0.097 70.070
The viscosity of helium® is n = 4.23 X 107¢71-5/ 3 0.032 178.91
(T0-826 — 0.409). 4 0.0157 338.0
Table 10.7. Diffusion coefficient of particles in 5 psig of helium
Diameter of Diffusion coefficient (cm?/sec) at temperature of —
particle (&) 923°K 873°K 533°K 473°K 433°K
3 5.250 4.880 2.530 2.160 1.920
10 4.73E-1 4.39E-1 2.28E-1 1.95E-1 1.73E-1
30 5.26E-2 4.88E-2 2.54E-2 2.17E-2 1.93E-2
100 4.74E-3 4.40E-3 2.29E-3 1.96E-3 1.74E-3
300 5.31E-4 4.93E—4 2.58E-4 2.21E—-4 1.97E-4
1,000 4.90E-5 4.56E-5 243E-5 2.09E-5 1.87E-5
1,700 1.74E-5 1.62E—5 8.83E-6 7.65E-6 6.89E—6
3,000 5.89E-6 5.51E-6 3.11E-6 2.73E-6 2.48E-6
10,000 7.05E-7 6.70E-7 4.43E-17 4.06E-17 3.81E-7
30,000 1.46E—7 1.41E-7 1.08E-7 1.02E-7 9.75E-8

For suitably short distance [i.e., X < 0.01 (Q/Db;)]
the exponential factors approach unity, and we may
write

ne =~

D
L Q X 27.
This is valid in our case for particles above 1000 A at
distances below about 2 m.
As a point for later reference, in the case of particles
of 1700 A in 523°K gas flowing at 3300 std cm® of
helium per minute and 5 psig,

27X 8.8X 1076
80

f:s'_: :3)(10—6_

g

10.6.2 Deposition by thermophoresis. Because the
gas leaving the pump bowl (at about 650°C) was
cooled considerably (temperature below S500°F in
the jumper region), the heat loss through the con-
duit walls implies an appreciable radial temperature
gradient. A temperature gradient in an aerosol results in
a thermally driven Brownian motion toward the cooler
region, called thermophoresis. This effect, for example,
causes deposition of soot on lamp chimney walls. The
velocity of particles smaller than the mean free path is
independent of size or composition. At diameters
somewhat greater than the mean free path, particle
velocities are again independent of diameter but are

107

diminished if the thermal conductivity of the particle is

much greater than the gas. (The effect of thermal
conductivity is not appreciable in our case.)
For particles in helium the radial velocity is given as

dT
V= 0.0024(T/300) —-.
r

The radial heat transfer conditions determine both
the internal radial gradient and also the axial tempera-
ture decrease for given flow conditions. Simple stepwise
models can be set up and deposition characteristics
calculated. One such model assumed Y% -in.-ID conduit
(1-in.-OD), slow slug flow at 3300 std cm®/min and 5
psig, with about 50 W of beta heat per liter in the gas,
an arbitrary wall conductance, and horizontal tube
convective loss to a 60°C ambient. With a 650°C inlet
temperature we found the values given in Table 10.8.
We probably do not know the conditions affecting heat
loss too much better than this, which is believed to be
reasonably representative of the MSRE conditions.

It is evident that for particles of sizes above about
100 A the thermophoretic effect was dominant in

Table 10.8. Thermophoretic deposition parameters
estimated for off-gas line

Distance

from inlet Temperature n/ng nJ/ng
cO
(cm)
10 546 0.86 12% 1072
50 306 0.58 3.9x 1073
100 191 0.46 1.6 %1073
150 147 0.40 9.1x10™
200 132 0.36 6.8x 107

producmg the -observed depositions (ng/ny was about 6
to 16 X 10™* for thermophoresis and about 3 X 107¢
for diffusion of 1700-A particles).

The insensitivity of thermophoresis to particle size
indicates that the observed mixture of sizes could be
approximately representative of that emergmg in the
pump bowl off-gas stream.

Thus the ratio Z = n./ny of deposit per unit length,
ng, to that entering the conduit, ny, can be estimated
on a thermophoretic basis and on a diffusional basis.
The thermophoretic effect is appreciably greater in the
regions where the gas is cooled appreciably if particle
sizes are abovz about 100 A. We will assume below that
values of Z are available.

10.6.3 Relationship between observed deposition
and reactor loss fractions. Four patterns of deposi-
tion will be described, and the laws relating the ob-
served deposition to the reactor situation will be
indicated.

1. Stable species in constant proportion to salt
constituents follow the simple relation ny = ny/Z. Thus,
given a value of Z, the total amount entering the off-gas
system can be obtained from an observation of the
deposit on unit length. For jumper-line situations Z is
about 0.001 (within a factor of 2). Steady deposition
can be assumed. '

2. A second situation relates to the transport of
nuclides which are not retained in the salt and part of
which may be transported promptly into the off-gas
system, for example, the noble metals. In developing a
relationship for this mechanism, we define /.. to as the
total MSRE inventory of the nuclide at time ¢,
produced after time ¢,. If the fraction of production
which goes to off-gas is fr(0G), it may be shown that n,
= Z X fi(OG) X I;_¢,- Inasmuch as we have inventory
values at various times,

If——f'):It"—Ito €Xp [—?\(t’—t())] .
In many cases, I1/l;—¢, = 1.
For many of the noble metals, ng/l; ~ 1077 to 107%;
thereby

_1
0.001

An extension of this case will involve holdup of the
nuclide in a “pool” for some average period. If the
holdup time is significantly lower than the half-life, the
effect is not great and should involve an additional
factor et AT,

For appreciable holdup periods and complex power
histories, integral equations are not presented. Dif-
ferential equations and a several-compartment model,
numerically integrated through the power history,
could be used to produce a value of the total atoms to
off-gas (ng).

3. A third mechanism of transport is applicable to
fission products which remained dissolved in salt, with
some salt transported into the off-gas, either as discrete
mist particles or conceivably as small amounts accumu-
lated on some ultimately transported particle, possibly
as the result of momentary vaporization of salt along a
fission spike.

If we assume a continuous transport during any
interval of reactor power,

fi(0G) < X 1077 <0.001 .

dC
—=PFy —AC
a7

and
dng
—=WZC — \n, ,
dt §

where C is the number of atoms per gram of salt, P is
reactor power, F is the number of fissions per gram at
unit power in unit time, y is fission yield, W is the rate
of salt transport to off-gas in grams per unit time, Z =
ng/ng, and ng is the number of atoms in unit length of
deposit.

From these for interval { we find:

PF
Ci=CiyeMit =2 (1 - ¢ M)

and
. Wi v
Msi =Mi—1)e” % {Cf—fl\tfe M

+P—];:y [1-(1 +?\t,-)e‘“f]} .

The equation for C; is evidently simply a version of the
inventory relationship. It appears possible to carry the
second equation through the power history. If we take
into account the lack of transport when the reactor is
drained and assume that the rate of salt transported to
off-gas is constant while the salt is circulating, we may
write W; = Wf;, where f; = 0 when drained and 1 when
circulating. Then the WZ/A term can be factored out,
and the term

IG(t - t0)=  f; {Gro hage ™

+PL§Z [1—-(1+ ?\z‘,-)e"“i]} e MTs=T0)

can be obtained by computation through the power
history in conjunction with the inventory equation: .

WZ
ns = TIG(IS — to) .

The evaluation of IG(+ — ), the gas deposit
inventory, has not yet been done.

4. The final case considers the daughters of noble
gases, insofar as they are produced by decomposition in
the off-gas stream. Here the mechanism changes: The

~daughter atoms diffuse relatively rapidly to the walls, so

that to a good approximation the rate of deposition of

daughter atoms on the walls is the rate of decay of

xenon Krypton in the adjacent gas. Thus we must, for

short-lived parent atoms, define the time of flow (1)

from the conduit entrance to the point (x) in question.
Let

@)= [ 4w pe) ax.

where p is the gas density at the point x (at 7, P), Wis
the mass inlet rate of the gas, and. A is the cross-
sectional area of the conduit.
We may now show that the accumulated activity of
the daughter in disintegrations per minute per centi-
meter at a particular point is

A, A(x
IQ(—x())e_MT(x) fr(0G) Iz (¢~ 1o) »
where fr(OG) is the fraction of the parent production
which enters the conduit and I,(;—¢,) is the MSRE
inventory activity of the daughter for the period (z, ¢;).
Note in particular that because deposition of atoms
(rather than particulates) is rapid and the daughter
atoms are formed in flow, no Z factor appears here.
Actually the daughter atoms of noble gases are
ubiquitous in that they might deposit from the gas onto
particulate surfaces and be transported in that fashion.
A major part is contained in the salt and would of
course move with that. However, either of these
alternative mechanisms is indicated to result in much
less deposition than that which occurs by deposition
from decay of a short-lived parent in the adjacent gas.
We now proceed to examine some of the data
presented above in the light of these relationships.

10.7 Discussion of Off-Gas Line Transport

10.7.1 Salt constituents and salt-seeking nuclides.
The deposit data for salt constituent elements and
salt-seeking nuclides can be interpreted as total amounts
of salt entering the off-gas system over the period of
exposure, using the equation presented earlier, ny =
ng/Z.

We have available to us the deposition per unit length
on segments from the jumper-line corrugated tubing
after run 14 and the specimen holder tube after run 18.
The deposits presumably occurred by simple diffusion
of particulates to the walls, or by thermophoresis. The
thermophoresis mechanism is over 100-fold more rapid
here. For the regions under consideration, calculations
indicate that within a factor of about 2, the ther-
mophoretic deposition per centimeter would be about
0.001 times that entering the tube, relatively independ-
ently of particle size. A similar rate could occur by
diffusion alone for particles of 100 A ‘but electron
microscope photographs showed particles 1000 to 3000
A, as well as larger.

However, the only way for the off-gas to have cooled
to the measured levels at the jumper line requires radial
temperature gradients to lose the heat, and the ther-
mophoretic effect of such gradients is well established.

Consequently, we conclude the thermophoretic effect
must have controlled deposition in the off-gas line near

109

the pump bowl. Thus we use a value ngfng = 0.001.
Amounts of gas-borne salt estimated to have entered
the off-gas system are shown in Table 10.9.

The deposit data for the samples within a given period
are passably consistent. Using the thermophoretic depo-
sition factor, the data indicate that the amounts of salt
entering the off-gas system during the given periods
were of the order of only a few grams.

The calculation for the radioactive nuclides was
crude; only the accumulation across runs 13—14 and
17—18 was considered, and this was treated as a single
interval at average power, prior inventory being ne-
glected. These assumptions should not introduce major
error, however. We conclude this indicates that major
quantities of particulate material did not pass beyond
the jumper line into the off-gas system.

About 60% of the particulate material leaving the
pump bowl should be transported to the walls by
thermophoresis in the first 2 m or so from the pump
bowl.

Diffusion alone would result in rates several hundred-
fold slower (for 1700-A particles; this varies approxi-
mately inversely with the square of particle diameter).
Consequently, if this mechanism controlled the ob-
served deposits, it would imply that considerably more
particulate material left the pump bowl and passed
through the jumper line.

10.7.2 Daughters of noble gases. The deposit activity
observed for the daughters of noble gases can be used to
calculate the fraction of production of the parent noble
gas which enters the off-gas system. Diffusion of
daughter atoms is more rapid than thermophoresis, and
deposition is assumed to occur at the same tube
positions that the parent noble gas undergoes decay in
the adjacent gas.

Table 10.9. Gra:hs of salt estimated to enter off-gas system

Runs 15-18,4748 hr

Runs 10—-14, 9112 hr

Basis O_f Upstream  Downstream Siz(ig:fn Si:i:zfn
C u

slevlation  hose hose tube 1 tube 2
Li 2.3 4.2 0.15 0.17
Be 0.9 1.5 0.12 0.12
Zr 0.6 0.6

U-235 1.4 4

Zr-95 0.15 4 0.2 04
Ce-141 0.2 0.8

Ce-144 0.6 6

Nd-147 5 43

110

Table 10.10. Estimated percentage of noble-gas nuclides entering the off-gas based on deposited
daughter activity and ratio to theoretical value for full stripping

Runs 10-14 Runs 15-19
G Daugh Upstream Downstream Specimen holder Specimen holder
as aughter hose hose tube 1 tube 2

Percent Ratio Percent Ratio Percent Ratio Percent Ratio
191-sec Kr-89 - Sr-89 0.8 0.06 4.0 0.28 4.0 0.28 _ 5.2 0.36
33-sec Kr-90 Sr-90 : 0.04 0.04 0.17 0.19
9.8-sec Kr-91 Y-91 0.07 1.00 0.003 0.04 0.004 0.06 0.004 0.06
16-sec Xe-140 Ba-140 0.004 0.03 0.019 0.12 0.005 0.03 0.008 0.05
234-sec Xe-137 Cs-137 1.3 0.07 6.0 0.32 4.6 0.25 5.4 0.29

The fraction of noble gas (1) entering the off-gas
system is calculated from daughter (2) activity:

obs dis min ™! ¢cm ™! inventory total
fr(1) = _ X1 s ,
inventory total 2 inventory period/

v flow rate X i o—TiAx

area A

The values shown below were calculated assuming a

flow rate of about 80 cm?>/sec and a delay (7, ) of about
2 sec between the pump bowl and the deposition point.
All daughter nuclides which result from the decay of a
noble-gas isotope are assumed to remain where depos-
ited. _
. The indicated percentage of production entering the
off-gas was compared with the amount calculated to
enter the off-gas if full stripping of all of the noble-gas
burden of the salt entering the pump bowl occurred,
with no entrainment in the retu:n flow, no holdup in
graphite or elsewhere, etc. The results are indicated in
Table 10.10. The magnitudes appear plausible.

Most of the values for the longerlived gases (3°Kr
and '27Xe) are 25 to 32% of the theoretical maximum,

observed activity

fr(off-gas) = MSRE inventory

MSRE inventory 1
MSRE period inventory Z’

where Z is again the ratio of the amount deposited per
centimeter to the amount entering the off-gas system
(here, of the nuclide in question). This factor as before
is approximately 0.001 if the thermophoresis mecha-
nism is dominant.

The period of operation was long (runs 10 to 18
extended over 380 days, runs 15 to 18 over 242 days)
with respect to the halflife of most noble-metal
nuclides, so that the ratio of the MSRE inventory to the
MSRE period inventory is not much above unity (in the
greatest case, 367-day ' ®®Ru, it is less than 1.1 for runs
10 to 14, and for runs 15 to 18 the ratio is below 2.7,
in spite of the shorter period, the longer prior period,
and changes in fission yield).

Table 10.11, Estimated fraction of noble-metal
production entering off-gas system

indicating that the net stripping was only partially Runs 1014 Runs 15-18
complete, possibly attributable to bubble return, in- Nuclid Specimen  Specimen
. uclide  Upstream  Downstream

complete mass transfer, and graphite holdup. hose hose holder holder

The ratio values for °Kr, ®'Kr, and !%°Xe mostly tube 1 tube 2
are 0.06 to 0.04; the net strippi stob

¢ met stripping appears 10 be oo 000002 0.00001  0.00003

somewhat less for these shorter-lived nuclides. Slow Mo-99 0.00010 0.00160 _
mass transfer from salt to gas phases could well account Ru-103  0.00012 0.00130 0.00002
for both groups. Ru-106  0.00074 0.00450 0.000005

10.7.3 Noble metals. The activity of deposits of  Agl1l11  0.00009 0.00260
noble-metal nuclides can be used to estimate the %zgg 8'888(1)2 g‘ggégg 0.00001 0.00002
fraction of production that entered the off-gas system. 131 0.00003 0.00010 0.00430 0.00002

The relationship employed is

Thus, to a useful approximation,

obs activity per cm

X
MSRE inventory

1
fr(off-gas) = 7
We tabulate in Table 10.11 this fraction, using for Z the
value of 0.001 as before.

These values, of course, indicate that only negligible
amounts of noble metals entered the off-gas system,
tenths to hundredths of one percent of production. We
believe that the assumptions involved in the above
estimate are acceptable and consequently that the
estimates do indicate the true magnitude of noble-metal
transport into off-gas.

Obviously, the estimated values depend directly on
the inverse of the deposition factor, Z. If only the
diffusion mechanism were active (which we doubt), the
estimated amounts transported into the off-gas would
be increased several hundredfold (300 X 7). Even in this
situation the estimated fractions of noble metals trans-
ported into the off-gas would mostly be of the order of
a few percent or less.

We consequently believe that the observed activities
of noble metals in off-gas line deposits indicate that
only negligible, or at most minor, quantities of these
substances were transported into the off-gas system.

111

References

1. A. N. Smith, “Off-Gas Filter Assembly,” MSR
Program Semiannu. Progr. Rep. Aug. 31, 1966, ORNL-
4037, pp. 714-77. ‘

2. E. L. Compere, “Examination of MSRE Off-Gas
Jumper Line,” MSR Program Semiannu. Progr. Rep.
Aug. 31, 1968, ORNL-4344, pp. 206—10.

3. E. L. Compere, “Examination of Material Re-
covered from MSRE Off-Gas Line,” MSR Program
Semiannu. Progr. Rep. Feb. 28, 1969, ORNL-4396, pp.
144—45, '

4. W. W. Parkinson (ORNL Health Physics Division),
“Analysis of Black Plug from Valve in MSRE,” letter to
E. L. Compere (ORNL Chemical Technology Division),
Jan. 21, 1970.

5. Cited by J. W. Thomas, The Diffusion Battery
Method for Aerosol Particle Size Determination,
ORNL-1649 (1953), p. 48.

6. C. F. Bonilla, Nuclear Engineering Handbook, ed.
H. Etherington, McGraw-Hill, New York, 1959, p.
9-25. ‘

7. C. N. Davies, ed., Aerosol Science, Academic, New
York, 1966.

8. N. A. Fuchs, The Mechanics of Aerosols, Macmil-
lan, New York, 1964.
112

11. POST-OPERATION EXAMINATION OF MSRE COMPONENTS

Operation of the Molten Salt Reactor Experiment was
terminated on December 12, 1969, the salt drained, and
the system placed in standby condition. In January
1971 a number of segments were removed from
selected components in the reactor system for exami-
nation. These included a graphite bar and control rod
‘thimble from the center of the core, tubing and a
segment of the shell from the heat exchanger, and the
sampler-enricher mist shield and cage from the pump
bowl. The examination of these items is discussed
below. It was expecient to extract parts of the original
reports in preparing this summary.

11.1 Examination of Deposits from the Mist Shield
in the MSRE Fuel Pump Bowl

In January 1971 the sampler cage and mist shield
were excised from the MSRE fuel pump bowl by using
a rotated cutting wheel to trepan the pump bowl top.
The sample transfer tube was cut off just above the
latch stop plug penetrating the pump bowl top, the
adjacent approximately 3-ft segment of tube was
inadvertently dropped to the bottom of the reactor cell
and could not be recovered. The final ligament attach-
ing the mist shield spiral to the pump bowl top was
severed with a chisel. The assembly was transported to
the High Radiation Level Examination Laboratory for
cutup and examination.

Removal of the assembly disclosed the copper bodies
of two sample capsules that had been dropped in 1967
and 1968 lying on the bottom of the pump bowl. Also
on the bottom of the bowl, in and around the sampler
area, was a considerable amount of fairly coarse
granular, porous black particles (largely black flakes
about 2 to 5 mm wide and up to 1 mm thick). Contact
of the heated quartz light source in the pump bowl with
this material resulted in smoke evolution and appar-
ently some softening and smoothing of the surface of
the accumulation.

A few grams of the loose particles were recovered and
transferred in a jar to the hot cells; a week later the jar
was darkened enough to prevent seeing the particles
through the glass. An additional quantity of this
material was placed loosely in the assembly shield
carrier can. Samples were submitted for analysis for
carbon and for spectrographic and radiochemical analy-
ses. The results are discussed below.

The sampler assembly as removed from the carrier can
is shown in Fig. 11.1. All external surfaces were covered
with a dark-gray to black film, apparently 0.1 mm or

more in thickness. Where the metal of the mist shield
spiral at the top had been distorted by the chisel action,
black eggshell-like film had scaled off, and the bright
metal below it appeared unattacked. Where the metal
had not been deformed, the film did not flake off.
Scraping indicated a dense, fairly hard adherent black-
ish deposit.

On the cage ring a soft deposit was noted, and some
was scraped off; the underlying metal appeared unat-
tacked. The heat of sun lamps used for in-cell photogra-
phy caused a smoke to evolve from deposits on bottom
surfaces of the ring and shield. This could have been
material, picked up during handling, similar to that seen
on the bottom of the pump bowl.

At this time, samples were scraped from the top,
middle, and bottom regions of the exterior of the mist
shield, from inside the bottom, and from the ring. The
mist shield spiral was then cut loose from the pump
bowl segment, and cuts were made to lay it open using
a cutoff wheel. A view of the two parts is shown in Fig.
11.2.

In contrast to the outside, where the changes between
gas (upper half) and liquid (lower) regions, though
evident, were not pronounced, on the inside the lower
and upper regions differed markedly in the appearance
of the deposits.

In the upper region the deposits were rather similar to
those outside, though perhaps more irregular. The
region of overlap appeared to have the heaviest deposit
in the gas region, a dark film up to 1 mm thick, thickest
at the top. The tendency of aerosols to deposit on
cooler surfaces (thermophoresis) is called to mind. In
the liquid region the deposits were considerably thicker
and more irregular than elsewhere, as if formed from
larger agglomerates.

In the area of overlap in the liquid region, this kind of
deposit was not observed, the deposit resembling that
on the outside. If we recall that flow into the mist
shield was nominally upward and then outward through
the spiral, the surfaces within the mist shield are
evidently subject to smaller liquid shear forces than
those outside or in the overlap, and the liquid was
surely more quiescent there than elsewhere. The con-
ditions permit the accumulation and deposition of
agglomerates.

The sample cage deposits also were more even in the
upper part, becoming thickest at and on the latch stop.
The deposit on the latch stop was black and hard,
between 1 and 2 mm thick. Deposits on the cage rods

113

R-5363!

Fig. 11.1. Mist shield containing sampler cage from MSRE pump bowl.

below the surface (see Figs. 11.3 and 11.4) were quite
irregular and lumpy and in general had a brown-tan
(copper or rust) color over darker material; some
whitish material was also seen. Four of the rods were
scraped to recover samples of the deposited material.
After a gamma-radiation survey of the cage at this time,
the unscraped cage rod was cut out for metallographic
examination; another rod was also cut out for more
thorough scraping, segmenting, and possible leaching of
the surfaces.

The gamma radiation survey was conducted by
lowering the cage in %, -in. or smaller steps past a 0.020-
by 1.0-in. horizontal collimating slit in 4 in. of lead.
Both total radiation and gamma spectra were obtained
using a sodium jodide scintillation crystal. The radiation
levels were greatest in the latch stop region at the top of

the cage and next in magnitude at the bottom ring.
Levels along the rods were irregular but were higher in
the liquid region than in the gas area even though
considerable material had been scraped from four of the
five rods in that region. In all regions the spectrum was
predominantly that of 367-day '°®Ru and 2.7-year
1258, and no striking differences in the spectral shapes
were noted.

Analyses of samples recovered from various regions
inside and outside the mist shield and sampler cage are
shown in Table 11.1. The samples generally weighed
between 0.1 and 0.4 g. The radiation level of the
samples was measured using an in-cell G-M probe at
about 1-in. distance and at the same distance with the
sample surrounded by a Y;-in. copper shield (to absorb
the 3.5-MeV beta of the 30-sec '°®™Rh daughter of

114

R-54220

6

Fig. 11.2. Interior of mist shield. Right part of right segment overlapped left part of segment on left.
Fig. 11.4. Deposits on sampler cage. Ring already scraped.

116

198 Ru). Activities measured in this way ranged from 4
R/hr (2 R/hr shielded) to 180 R/hr (80 R/hr shielded),
the latter being on a 0.4-g sample of the deposit on the
latch stop at the top of the sample cage.

Spectrographic and chemical analyses are available on
three samples: (1) the black lumpy material picked up
from the pump bowl bottom, (2) the deposit on the
latch stop at the top of the sample cage, and (3)
material scraped from the inside of the mist shield in
the liquid region. The material recovered from the
pump bowl bottom contained 7% carbon, 31%
Hastelloy N metals, 3.4% Be (18% BeF,), and 6% Li
(22% LiF). Quite possibly this included some cutting
debris. The carbon doubtless was a tar or soot resulting
from thermal and radiolytic decomposition of lubri-
cating oil leaking into the pump bowl. It is believed that
this material was jarred loose from upper parts of the
pump bowl or the sample transfer tube during the chisel
work to detach the mist shield.

The hard deposit on the latch stop contained 28%
carbon, 2.0% Be (11% BeF,), 2.8% Li (10% LiF), and
12% metals in approximate Hastelloy N proportions,
again possibly to some extent cutting debris.

The sample taken from the inner liquid region of the
mist shield contained 2.5% Be (13% BeF,), 3.0% Li
(11% LiF), and 18% metals (with somewhat more
chromium and iron than Hastelloy N); a carbon analysis
was not obtained.

In each case, about 0.5 to 1% Zr (about 1 to 2%
ZrF,) was found, a level lower than fuel salt in
proportion to the lithium and beryllium. Uranium
analyses were not obtainable; so we cannot clearly say
whether the salt is fuel salt or flush salt. Since fission
product data suggest that the deposits built up over
appreciable periods, we presume that it is fuel salt.

In all cases the dominant Hastelloy N constituent,
nickel, was the major metallic ingredient of the deposit.
Only in the deposit from the mist shield inside the
liquid region did the proportions of Ni, Mo, Cr, and Fe
depart appreciably from the metal proper. In this
deposit a relative excess of chromium and iron was
found, which would not be attributable to incidental
metal debris from cutting operations. It is also possible
that the various Hastelloy N elements were all subject
to mass transport by salt during operation, and that
little of that found resulted from cutup operation.

We now come to consideration of fission product
isotope data. These data are shown in Table 11.2 for
deposits scraped from a number of regions. The activity
per gram of sample is shown as a fraction of MSRE
inventory activity per gram of MSRE fuel salt to
eliminate the effects of yield and power history;
Table 11.1. Chemical and spectrographic analysis of deposits from mist shield in the MSRE pump bowl

Radiation level

. 3
Samp?le (R/hr @ 1 in)) Weight  Percent . ,T_l _;. Percent Li Percent Be Percent Zr* Percent Ni¥ Percent Mo® Percent Cr Percent Fe® * Percent Mn®
Location . (mg) C (dismin~'g™") » A
‘ (shielded)
Pump bowl 10(5) 402 3.1 El0 6.00 3.42 0.5-1.0 20--30 2-4 1-2 0.5-1.0 0.5-1.0
- bottom 251 108 71
Latch stop 130(60) 291 1.85 El1 2.75 2.01 ©0.5-1.0 5-10 24 0.5-1.0 0.5-1.0 <0.5
180(80 365 28
Inside, 40(17) 179 4.7 E10 3.03 2.52 0.5-1.0 5-10 2-4 3-5 2-4 <0.5
liquid :
region
MSRE fuel 11.1 6.7 11.1
(nominal)
Hastello_y N 69 16 7 5 ~1
(nominal)

9Semiquantitative spectrographic determination.

LTT
Table 11.2. Gamma spectrographic (Ge-diode) analysis of deposits from mist shield in the MSRE pump bowl

99 95Nb 103, 106R, 125gp 127mT, 1370y 957,49 1444
Half-life 2.1 X 105 years 35 days 39.6 days 367 days 2.7 years 105 days 30 years 65 days 284 days
(after 25 Z1)
Inventory, dis min~! g~} 24 uglg 8.3E10 3.3 ElO 3.3E9 3.7 E8 2.0E9 6.2 E9 9.9 E10 59EI10
Sample activity per gram expressed
as fraction of MSRE inventory
activity/grams fuel salt® »
Pump bowl bottom, 0.23 36 52 20 17 2.1 0.13:0.03 0.10
loose particles
Latch stop 265 364 1000 5.7 69 3.1 0 0
Top
Outside 122 167 328 104 98 9.5 0 0
Inside : 42+ 14 237 + 163 273 1000 69 43 0 0
Middle outside 2711 531 646 563 54 8.0
Below liquid surface
Inside No. 1 354 164 224 271 143 0.6 0 0.13 + 0.02
Inside No. 2 463 148 292 310 72 98 1.1 0 0
Cage rod 305 221 692 198 187 0.2 0 0
Bottom .
Outside 0, <60) (0, <60) 1210 167 232 3.2 0 0
Inside 16+ 9 15+ 9 189 51 27 0:32 0 0.11

9Background values (limit of detection) were as follows: ° >Zr, 2—9 E10; '*%Ce, 1-2 E10; ' 3*Cs, 2-9 E8; !'Ag, 1-3 E9; ! 5Eu, 1 E8-2 E9.

bUncertainty stated (as + value) only when an appreciable fraction (>10%) of observed.

811
materials concentrated in the same proportion should
have similar values.

We first note that the major part of these deposits
does not appear to be fuel salt, as evidenced by low
values of ?5Zr and 1#*Ce. The values 0.13 and 0.11 for
144 Ce average 12%, and this is to be compared with the
combined 24% for LiF plus BeF, determined spec-
trographically, as noted above. These would agree well
if fuel salt had been occluded steadily as 24% of a
growing deposit throughout the operating history.

For '37Cs we note that samples below liquid level
inside generally are below salt inventory and could be
occluded fuel salt, as considered above. For samples
above the liquid level inside, or any external sample,
values are two to nine times inventory for fuel salt.
Enrichment from the gas phase is indicated. Houtzeel?
has noted that off-gas appears to be returned to the
main loop during draining, as gas from the drain tanks is
displaced into a downstream region of the off-gas
system. However, our deposit must have originated
from something more than the gas residual in the pump
bowl or off-gas lines at shutdown. An estimate substan-
tiating this is as follows.

With full stripping, 3.3 X 10' 7 atoms of the mass 137
chain per minute enter the pump bowl gas. About half
actually go to off-gas, and most of the rest are
reabsorbed into salt. If we, however, assume a fraction f
is deposited evenly on the boundaries (gas boundary
area about 16,000 cm?), the deposition rate would be
about 2 X 10'3f atoms of the 137 chain per square
centimeter per minute. Now if in our samples the
activity is 7 relative to inventory salt (1.4 X 10! 7 atoms
of 137Cs per gram) and the density is about 2, then the
time ¢ in minutes required to deposit a thickness of X
centimeters would be

t_1.4><1017XI><2><X

=14X lO“—I—X min.
2X 101 %f f

In obtaining our samples, we generally scraped at least
0.1 g from perhaps 5 cm?, indicating a thickness of at
least about 0.01 ¢m, and 7 values were about 4, whence
t is about 600/f.

Thus, even if all (f ~ 1) the 137 chain entering the
pump bowl entered our deposits, 600 min of flow
would be required to develop their !37Cs content —
too much for the 7-min holdup of the pump bowl or
even the rest of the off-gas system, excluding the
charcoal beds.

It appears more likely that ! 37Cs atoms, from '3 7Xe
atoms decaying in the pump bowl, were steadily
incorporated to a slight extent in a slowly growing
deposit.

119

The noble-metal fission products, 35-day °5Nb,
39.6-day '°3Ru, 367-day !°¢Ru, 2.7-year !258b, and
105-day 27" Te, were strongly present in essentially
all samples. In all cases, 35-day ®5Nb was present in
quantities appreciably more than could have resulted
from decay of °3Zr in the sample.

Antimony-125 appears to be strongly deposited in all
regions, possibly more strongly in the upper (gas) region
deposits. Clearly '25Sb must be considered a noble-
metal fission product. Tellurium-127m was also found,
in strong concentration, frequently in similar propor-
tion to the '2%Sb of the sample. The precursor of

- 127mTe s 3.9-day !'27Sb. It may be that earlier

observations about fission product tellurium are in fact
observations of precursor antimony isotope behavior,
with tellurium remaining relatively fixed.

The ruthenium isotopes were present in quantities
comparable with those of ?5Nb, 1258b, and '>7™Te.
If the two ruthenium isotopes had been incorporated in
the deposit soon after formation in the salt, then they
should be found in the same proportion to inventory.
But if a delay or holdup occurred, then the shorter-lived
103 Ru would be relatively richer in the holdup phase as
discussed later, the activity ratio ! ®*Ru/!°®Ru would
exceed the inventory ratio, and material deposited after
an appreciable holdup would have an activity ratio
103Ru/! ®¢Ru which would be less than the inventory
value. Examination of Table 11.2 shows that in all
samples, relatively less '°?Ru was present, which
indicates that the deposits were accumulated after a

~ holdup period. This appears to be equally true for

regions above and below the liquid surface. Thus we
conclude that the deposits do not anywhere represent
residues of the material held up at the time of
shutdown but rather were deposited over an extended
period on the various surfaces from a common holdup
source. Specifically this appears true for the lumpy
deposits on the mist shield interior and cage rods below
the liquid surface.

Data for 2.1 X 10°-year °°Tc are available for one
sample taken from the inner mist shield surface below
the liquid level. The value, 1.11 X 10* ug/g, vs
inventory 24 ugf/g, shows an enhanced concentration
ratio similar to our other noble-metal isotopes and
clearly substantiates the view that this element is to be
regarded as a noble-metal fission product. The con-
sistency of the ratios to inventory suggests that the
noble metals represent about 5% of the deposits.

The quantity of noble-metal fission products held up
in this pump bowl film may not be negligible. If we
take a median value of about 300 times inventory per
gram for the deposited material, take pump bowl area
in the gas region as 10,000 cm? (minimum), and assume
deposits 0.1 mm thick (about 0.02 g/cm?; higher values
were noted), the deposit thus would have the equivalent
of the content of more than 60 kg of inventory salt.
There was about 4300 kg of fuel salt; so on this basis,
deposits containing about 1.4% or more of the noble
metals were in the gas space. At least a similar amount
is estimated to be on walls, etc., below liquid level; and
no account was taken for internal structure surfaces
(shed roof, deflector plates, etc., or overflow pipe and
tank).

Since pump bowl surfaces appear to have more (about
10 times) noble-metal fission products deposited on

them per unit area than the surfaces of the heat -

exchanger, graphite, piping, surveillance specimens, etc.,
we believe that some peculiarities of the pump bowl
environment must have led to the enhanced deposition
there.

We first note that the pump bowl was the site of
leakage and cracking of a few grams of lubricating oil
each day. Purge gas flow also entered here, and
hydrodynamic conditions were different from the main
loop.

The pump bowl had a relatively high gas-liquid
surface with higher agitation relative to such surface
than was the case for gas retained as bubbles in the
main loop. The liquid shear against walls was rather less,
and deposition appeared thickest where the system was
quietest (cage rods). The same material appears to have
_ deposited in both gas and liquid regions, suggesting a
common source. Such a source would appear to be the
gas-liquid interfaces: bubbles in the liquid phase and
droplets in the gas phase. It is known that surface-
seeking species tend to be concentrated on droplet
surfaces.

The fact that gas and liquid samples obtained in
capsules during operation had '°3Ru/!'°®Ru activity
ratios higher than inventory and deposits discussed here
had '°3Ru/!°SRu activity ratios below inventory
suggests that the activity in the capsule samples was
from a held-up phase that in time was deposited on the
surfaces which we examined here.

The tendency to agglomerate and deposit in the less
agitated regions suggests that the overflow tank may
have been a site of heavier deposition. The pump bowl
liquid which entered the overflow pipe doubtless was
associated with a high proportion of surface, due to
rising bubbles; this would serve to enhance transport to
the overflow tank.

The binder material for the deposits has not been
established. Possibilities include tar material and per-
haps structural- or noble-metal colloids. Unlikely,
though not entirely excludable, contributors are oxides

120

formed by moisture or oxygen introduced with purge
gases or in maintenance operations. The fact that the
mist shield and cage were wetted by salt suggests such a
possibility.

11.2 Examination of Moderator Graphite
from MSRE

A complete stringer of graphite (located in an axial
position between the surveillance specimen assembly
and the control rod thimble) was removed intact from
the MSRE. This 64.5-in.-long specimen was transferred
to the hot cells for examination, segmenting, and
sampling.

11.2.1 Results of visual examination. Careful exami-
nation of all surfaces of the stringer with a Kollmorgen
periscope showed the graphite to be generally in very
good condition, as were the many surveillance speci-
mens previously examined. The corners were clean and
sharp, the faint circles left upon milling the fuel channel
surfaces were visible and appeared unchanged, and the
surfaces, with minor exceptions described below, were
clean. The stringer bottom, with identifying letters and
numbers scratched on it, appeared identical to the
photograph taken before its installation in MSRE.

Examination revealed a few solidified droplets of
flush salt adhering to the graphite, and one small.
(0.5-cm?) area where a grayish-white material appeared
to have wetted the smooth graphite surface. One small
crack was observed near the middle of a fuel channel.
At the top surface of the stringer a heavy dark deposit
was visible. This deposit seemed to consist in part of
salt and in part of a carbonaceous material; it was
sampled for both chemical and radiochemical analysis.
Near the metal knob at the top of the stringer a crack in
the graphite had permitted a chip (about 1 mm thick
and 2 cm? in area) to be cleaved from the flat top
surface. This crack may have resulted from mechanical
stresses due to threading the metal knob into the
stringer (or from thermal stresses in this metal-graphite
system during operation) rather than from radiation or
chemical effects.

11.2.2 Segmenting of graphite stringer. Upon com-
pletion of the visual observation and photography and
after removal of small samples from several locations on
the surface, the stringer was sectioned with a thin
Carborundum cutting wheel to provide five sectiofis of
4-in. length, three thin (10- to 20-mil) sections for x
radiography, and three thicker (30- to 60-mil) sections
for pinhole scanning with the gamma spectrometer.
Each set of samples contained specimens from near the
top, middle, and bottom of the stringer. The large
specimens, from which surface samples were subse-

121

quently milled, included (in addition) two samples from

intermediate positions.

11.2.3 Examination of surface samples by x-ray
diffraction. Previous attempts to determine the chemi-
cal form of fission products deposited on graphite
surveillance specimens by x-ray reflection from flat
surfaces failed to detect any element except graphitic
carbon. A sampling method which concentrated surface
impurities was tried at the suggestion of Harris Dunn of
the Analytical Chemistry Division. This method in-
volved lightly brushing the surface of the graphite
stringer with a fine Swiss pattern file which had a
curved surface. The grooves in the file picked up a small
amount of surface material, which was transferred into
a glass bottle by tapping the file on the lip of the bottle.
In this way, samples were taken at the top, middle, and
bottom of the graphite stringer from the fuel-channel
surface, from the surface in contact with graphite, and
from the curved surface adjacent to the control rod
thimble.

Three capillaries were packed and mounted in holders
which fitted into the x-ray camera.

Samples from the fuel-channel surfaces yielded very
dark films, which were difficult to read. Many weak
lines were observed in the x-ray patterns. Since other
analyses had shown Mo, Te, Ru, Tc, Ni, Fe, and Cr to
be present in significant concentrations on the graphite
surface, these elements and their carbides and tellurides
were searched for by careful comparison with the
observed patterns.

In all three of the graphite surface samples analyzed,
most of the lines for Mo, C and ruthenium metal were
certainly present. For one sample, most of the lines for
Cr,C; were seen. (The expected chromium carbide in
equilibrium with excess graphite is Cr;C,, but nearly
half the diffraction lines for this compound were
missing, including the two strongest lines.) Five of the
six- strongest lines for NiTe, were observed. Molybde-
num metal, tellurium metal, technetium metal, chro-
mium metal, CrTe, and MoTe, were excluded by
comparison of their known pattern with the observed
spectrum.’ These observations (except for that of
Cr,C3) are in accord with expected chemical behavior
and are significant in that they represent the first
experimental identification of the chemical form of
fission products known to be deposited on the graphite
surface. )

11.2.4 Milling of surface graphite samples. Surface
samples for chemical and radiochemical analyses were
milled from the five 4-in.-long segments from the top,
middle, bottom, and two intermediate locations on the

graphite stringer using a rotating milling cutter 0.619 in.
in diameter. The specimen was clamped flat on the bed
of the machine, and cuts were made from the flat
fuel-channel surface and from one of the narrower flat
edge surfaces on both sides of the fuel channel. The
latter surfaces contacted the similar surfaces of an
adjacent stringer in the MSRE core. After the sample
was clamped in position the milling machine was
operated remotely to take successive cuts 1, 2, 3, 10,
20, 30, 245, 245, and 245 mils deep to the center of
the graphite stringer. The powdered graphite was
collected in a tared filter bottle attached to a vacuum
cleaner hose during and after each cut. The filter bottle
was a plastic cylindrical bottle with a circular filter

" paper covering a number of drilled holes in the bottom.

This technique collected most of the thinner samples
but only about half of the larger 245-mil samples. After
each sampling, the uncollected powder was carefully
removed with the empty vacuum cleaner hose.

Before samples were cut from the narrow flats the
corners of the stringer bars were milled off to a width
and depth of 66 mils to avoid contamination from the
adjacent stringer surfaces. Then successive cuts 1, 2, 3,
10, 20, and 30 mils deep were taken. Finally, a single
cut 62 mils deep was taken on the opposite flat
fuel-channel surface. Only the latter cut was taken on
the two stringer samples from positions halfway be-
tween the bottom and middle and halfway between the
middle and top of the stringer.

11.2.5 Radiochemical and chemical analyses of
MSRE graphite. The milled graphite samples were
dissolved in a boiling mixture of concentrated H, SO,
and HNO; with provision for trapping any volatilized
activities. The following analyses were run on the
dissolved samples:

1. Gamma spectrometer scans for '°SRu, !2°Spb,

134CS,137CS, 1 IOAg, 144Ce,952r,95Nb,and°°C0.

. Separate radiochemical separations and analyses for
898r, ®%8r, 127Te, and ®H. A few samples were
analyzed for ®° Tc.

. Uranium analyses by both the fluorometric and the
delayed neutron counting methods.

. Spectrographic analyses for Li, Be, Zr, Fe, Ni, Mo,
and Cr. (High-yield fission products were also looked
for but not found.)

Uranium and . spectrographic analyses. Table 11.3
gives the uranium analyses (calculated as 233 U) by both
the fluorometric method and -the delayed neutron
counting method. The type of surfaces sampled, the
number of the cut, and the depth of the cut for each
122

Table 11.3. Chemical analyses” of milled samples

C 233Us ppm, [
ut and Depth, Total U, ppm, . Metal,
Sample b - . delayed Li, ppm Be; ppm Zr, ppm
type mils fluorometric ppm
neutron
1 1 Blank 0-6 <10 <2 - -
2 2 Blank 6-30 1 0.1 +0.05 <2 <1 - -
3 1FC 0-2 28 35515 360 320 1600 -
4 2 FC 1-3 21 269+ 1.2 340 170 - -
) 3FC 3-6 8 11.5+0.8 - — — -
7 S5FC 16-36 2 26+0.2 - - — -
11 9 FC 556--801 3 2603 310 200 - 820 Fe
12 1E 0-2 <1 (M 22.7+3.5 250 180 — High Fe
13 2E 2-3 <30 8.6=x213 110 50 — -
14 3E 3-6 9 5.5£0.7
17 6 E 36—66 3 50+£03
18 1 Deep 0-62 2 24 0.1 40 20 - -
19 1 FC 0-5 21 21.5+06 220 150 970 Mo,1100 Ni
20 2FC 1-17 9 10.1 £ 0.8 190 100 - —
21 3FC 3-10 4 7.1+04
23 S FC 16—-40 <1 1.0+ 0.7
26 8 FC 311-556 <2 0.8 0.6 10 610
217 1E 0-3 3 3.8+0.2 150 90 1400 High Fe
28 2E 1-5 <8 59+0.1 230 110
29 3E 3-8 3 57+04
31 1 Deep 0-62 4 33:20.1 80 50 70 150 Ni
32 1 Deep 0-62 14 13.0+0.2 120 90 80 180 Ni
33 1 FC 0-0.2 12 18.1+1.3 1400 290 High High Fe + Mo
34 2FC 0-3 36 29.2=+1.3 410 240 500 2900 Ni
35 3FC 3-6 18 20.0: 09
36 5E 16-36 1 2.5+0.1
39 6 FC 36-66 3 3501
43 1E 0-1 51 118+ 6 1000 400 8000 High Fe
44 2E 1-3 46 36.6 2 1.6 40 270 550 220 Ni
45 3E 3-6 8 7.8+0.7
48 6 E 36-66 2 3.6+0.2
49 1 Blank 0-2 <10 <2 - -
50 2 Blank 2-30 <1 0.1 + 0.04 <7 <0.3 <40 <70 Ni

?Dashes in the body of the table represent analyses showing none present. Blanks indicate the analyses were not done.
The number is the number of cut toward the interior starting at the graphite surface. *FC” stands for a fuel channel surface,
“E” for a narrow edge surface, and “Deep” for a first cut about 62 mils deep from a fuel channel surface.
€“High” indicates an unbelievably high concentration (several percent).

sample are also shown in the table. Samples between 19
and 29 were inadvertently tapered from one end of the
specimen block to the other so that larger-than-planned
ranges of cut depth were obtained. Samples from 3 to
18 were taken from the topmost graphite stringer
specimen, those from 19 to 31 and sample 36 were
from the middle specimen, and those from 32 to 48
were taken from the bottom specimen.
~ In view of the fact that the uranium concentrations
were at the extreme low end of the applicable range for
the fluorometric method, the agreement with the
delayed neutron counting method was quite satis-
factory. The data suggest that the sizable variations

between different surfaces (e.g., the three deep-cut
samples 18, 31, and 32) were real. Sizable variations
also exist in uranium concentrations in the deep interior
of different regions of the stringer. These values range
from 2.6 ppm at the top to 0.8 ppm at the middle to 3
ppm at the bottom. :

The concentration profiles indicated by the data in
Table 11.3 were generally similar to those previously
observed both on the surface and in the interior.
Concentrations dropped a factor of 10 in the first 16
mils. A tough calculation of the total 233U in the
MSRE core graphite indicates about 2 g on the surface
and about 9 g in the interior of the graphite. These low
values indicate that uranium penetration into modera-
tor graphite should not be a serious problem in
large-scale molten-salt reactors.

The fact that fluorometric values for total uranium
and the delayed neutron counting values for 233U
agreed (with the 233U value usually larger than the
total uranium value) indicates that little uranium
remained in the graphite from the operation of the
MSRE with 235U fuel. Apparently, the 238U and 235U
previously in the graphite underwent rather complete
isotopic exchange with 233U after the fuel was
changed. The finding of 150 to 2900 ppm nickel in a
few of the surface samples is probably real. The main
conclusions from the spectrographic analyses are that
adherent or permeated fuel salt accounted for the
uranium, lithium, and beryllium in half the samples and
that a thin layer of nickel was probably deposited on
some of the graphite surface.

Radiochemical analyses. Since the graphite stringer
samples were taken more than a year after reactor
shutdown, it was possible to analyze only for the
relatively longlived fission products. However, the
absence of interfering short-lived activities made the
analyses for long-lived nuclides more sensitive and
precise. The radiochemical analyses are given in Table
11.4, together with the type and location of surface
sampled, the number of the milling cut, and the depth
of the cut for each sample.

The species '25Sb, '°¢Ru, 1% Ag, ®*5Nb, and 127 Te
showed concentration profiles similar to those observed
for noble-metal fission products (*®Mo, *2° Te, 122 Te,
103Ry, 196Ru, and ?5Nb) in previous graphite surveil-
lance specimens, that is, high surface concentrations
falling rapidly several orders of magnitude to low
interior concentrations. The profiles for °°Zr were of
similar shape, but the interior concentrations were two
or three orders of magnitude smaller than for its
daughter ®*Nb. It is thought that ®* Zr is either injected
into the graphite by a fission recoil mechanism or is
carried with fuel salt into cracks in the graphite; the
much larger concentrations of **Nb (and the other
noble metals) are thought to result from the deposition
or plating of solid metallic or carbide particles on the
graphite surface. The °°Sr (33-sec °°Kr precursor)
profiles were much steeper than those previously
observed in surveillance specimens for 3°Sr (3.2-min
89Kr precursor), as expected. An attempt to analyze
the stringer samples for ®®Sr also was unsuccessful. It is
difficult to analyze for one of these pure beta emitters
in the presence of large activities of the other.

Surprisingly high concentrations of tritium were
found in the moderator graphite samples (Table 11.4).

123

* The tritium concentration decreased rapidly from about

10'! dis min™' g7! at the surface to about 10° dis
min™' g7! at a depth of ¥4 in. and then decreased
slowly to about half this value at the center of the
stringer. If all the graphite in the MSRE contained this
much tritium, then about 15% of the tritium produced
during the entire power operation had been trapped in
the graphite. About half the total trapped tritium was
in the outer Y, 4-in. layer.

Similarly high concentrations of tritium were found
in specimens of Poco graphite (a graphite characterized
by large uniform pores) exposed to fissioning salt in the
core during the final 1786 hr of operation. Surface
concentrations as high as 4.5 X 10'° dis min™* g™!
were found, but interior concentrations were below 108
dis min™' g7, much lower than for the moderator
graphite. This suggests that the graphite surface is
saturated relatively quickly but that diffusion to the
interior is slow.

If it is assumed that the surface area of the graphite
(about 0.5 m?/g) is not changed by irradiation (there
was no dependence of tritium sorption on flux), there
was 1 tritium per 100 surface carbon atoms. Since the
MSRE cover gas probably contained about 100 times as
much hydrogen (from pump oil decomposition) as
tritium, a remarkably complete coverage by chemi-
sorbed hydrogen is indicated.

An overall assessment of tritium behavior in the
MSRE? and proposed MSBR’s* is presented elsewhere.

The data on fission product deposition on and in the
graphite, based on Table 11.4, have been calculated as
(observed activity per square centimeter) divided by
(inventory activity/total area), as was done for surveil-
lance specimens earlier. The resulting relative deposit
intensities, shown in Table 11.5, can, of course, then be
compared with values reported for other nuclides, other
specimens, or other times. If all graphite surfaces were
evenly covered at the indicated intensity, the fraction
of total inventory in such deposits would be 74% of the
relative deposit intensity shown. For top, middle, and
bottom regions and for ‘channel and edge (graphite-to-
graphite) surfaces, values are shown for surface and
overall deep cuts.

As in the case of surveillance specimens, intensities
for salt-seeking nuclides are at levels appropriate for
fission recoil (about 0.001), the noble-gas daughters
(*7Cs and °°Sr) are 10 to 20 times as high, and the
noble metals notably higher, though, as we shall see,
not as high on balance as on metals. Where comparisons
can be made, most of the deposit was indicated to be at
the surface. Values for *Nb are far above °5Zr,
indicating that ®*Nb was indeed deposited; the graphite
Table 11.4. Radiochemical analyses of graphite stringer samples

Sample - Cut and Depth, : Disintegrations per minute per gram of graphite on 12-12-69 GO
No. type? mils 1256y, 106R, 1277, 9SNb  110Ag 99T 134¢s 137¢g  90g; 144 g 957, - '52Eu 154p, 60Co 3H
1 Thin blank 0-6 . <8E5 <4E6 - <TE9 <9ES <3ES <BE8 5.79E5 6.22E6 <2E8 2..'93E7 3.12E7 1.06E7 9.15E5
2 Thick blank 6-30 <B8E9 8.53E5 <7E8 <9E4 <3E4 <T7E4 1.56E6 <2E7 1.85E6 1.83E6 6.81E5 2.38E6
3 1FCT 0-2 3.1E10 7.31E10 8.95E10 <4E12 <1E9 ~330 <3E8 1.0E9 8.7E9 8.49E9 <1.8E10 <4.8E8 <5E8 ’ 3.63E8 5.92E10
4 2FCT 1-4 1.14E9 3.74E8 342E9 <2.7E11 2.84E8 ~13 1.06E7 2.39E8 1.1E10 4.26E8 <I1.1E9 1.76E7 2.37E7 44E7 1.21E10
5 3FC,T 3-7 5.52E8 1.08E9 1.81E9 <IEll 2.75E7 <7E6 2.39E8 3.88E9 1.72E8 <2E9  1.50E8 1.51E8 3.77E7 6.31E9
6 4 FC,T 6-17 1.76E8 3.72E8 5.23E8 <3E10 <6E7 4.39E6 349E8 3.38E9 5.27E7 <3ES8 <_1E7 <1.3E7 1.51E7 1.75E9
7 S FC,T 16-37 1.07E8 2.54E8 3.50E8 <1E10 <3.6E6 8.81E6 3.96E8 1.24E9 3.82E7 <2E8 <7E6 <5E6 1.14E6 1.85ES
8 6 FC,T 36-67 S5.91E7 2.02E8 1.96E8 <9E9 <3.3E6 2.03E6 2.69E7 2.78E8 3.79E7 <2E8 <9E6 <4.8E6 1.70E6 1.03E9
9 7 FC,T - 66-312 <TE6 290E7 1.39E7 2.73E10 <9ES5 3.32E7 3.00E8 7.23E7 1.69E8 2.95E8 <3E6 <3E6 <1.7E6 1.36E9
10 8 FC,T 311-557 <BE6 <2E7 489E6 <54E9 <I1E6 247E8 1.46E8 1.12E7. <4.1E7 <I1.2E8 <3E6 <2E6 <7E5 6.32E8
11 9 FC,T 556—-802 <1E7 <3.3E7 2.00E7 <T7E9 <3E6 3.04E8 1.64E8 2.12E6 <5E7 <2.0E8 <4E6 <16E6 <2E6 8.16E8
12 1E, T 0-2 3.02E9 1.82E10 1.07E10 1.59E12 3.62ES8 140E8 7.46E8 5.86E9 4.84E9 1.09E10 <8E7 <4.2E6 2.25E8 1.78E10
18 1 Deep,T 0-62 145E9 6.08E9 6.52E9 1.23E11 <3E7 <1E7 2.14E8 1.13E9 <2E8 <6ES8 <1E7 <7.9E7 1.05E7 3.48E9
19 1 FCM 0-5 249E9 9.20E9 1.21E10 1.57E12 5.22ES8 395E8 1.10E9 1.13E10 6.68E9 1.25E10 <6E7 <4E7 2.73E8 5.72E9
20 2FCM 1-7 1.50E9 ‘ 1.20E10 1.03E10
22 4 FCM 6-20 8.33E9
23 5 FCM 16-40 1.28E6 3.19E9
24 6 FC.M 36-70 7.95E8
26 8 FCM 311-556 <8E6 <39E9 S5.79E6 <3EI0 <3E6 2.85E8 <2E8 1.22E7 5.58E7 <2ES§ <3E6 <7.8ES5 6.11E8
27 1EM 0-3 9.10E9 7.13E9 291E9 -1.22E12 2.54E8 2.86E8 B8.50E9 8.67E9 4.52E9 8.50E9 <4E7 <3E7 1.77E8 1.56E10
28 2EM 1-5 493E7 1.83E8 2.39E8. 2.24E11 2.37E7 2.83E7 5.28E8 6.94E9 4.06E8 <I1E9 7.07E6 7.34E6 >2.30E9
36 ° SEM 16—-36 1.07E7 17.34E7 4.79E7 9.85E10 <4E6 3J40E7 5.14E8 2.18E8 <6ES§ 3.71E6 3.53E6 2.99E9
31 1 Deep,M 0-62 391E9 1.86E10 9.08E9 4.90E11 <9E7 9.72E7 9.13E8 2.26E9 1.10E§8 <2E9 <1E8 <6E7 1.24E8 5.50E9
32 1 Deep,B 0—62 145E9 4.62E9 5.84E9 2.16E11 <6.3E7 1.96E8 1.13E9 243E9 6.14E8 1.13E9 <3E7 <I1.6E7 4.30E7 8.45E9
33 1 FC,B 0--0.2 296E11 4.19E11 8.22E11 2.13E13 <2E9 1.37E9 4.69E9 1.07E10 5.09E10 <1E11 <3E9 <3E9 6.82E9 1.59El11
34  2FCB 0-3 5.76E9 1.15E9 1.75E10 3.04El12 1.57E9 1.76E8 6.43E8 8.13E9 4.67E9 1.14E10 <2.7E7 1.63E8. 4.31E10
35 3FC,B 3-6 1.27E9 3.29E9 3.22E9 3.78El1l1 2.76E8 1.17E7 2.66ES§ 451E8 <2E9 <1.8E7 1.94E7 1.26E10
38 5 FC,B 16—-36 3.27E8 8.38E8 7.27E8 <B8.3EI0 <5.7E6 2.20E7 1.17E9 <1E8 <6E8 6.26E6 196E7 3.46E9
40 7 FC,B 66-311 3.13E7 1.04E8 S5.20E7 <1.0E10 <2E6 1.03E8 1.33ES8 <3E7 <2E8 <1.6E6 3.23E6 5.69E8
42 9 FC,B 556-801 <I1E7 <39E7 6.19E6 <4EI0 <2Eé6 3.13E8 2.93E8 1.69E7 6.91E7 <3.2E8 <lE6 4.21E6 7.23E8
43 1EB 0-1 3.05E10 6.94E10 8.06E10 8.49E12 <7.8E8 8.57E8 2.83E9 1.11E10 2.42E10 4.58E10 <3.5E8 <2E8 1.27E9 9.18E10
44 2EB 1-3 3.45E8 8.30E8 1.06E9 1.50E12 S5.07ES8 3.07E7 2.71ES8 9.76E8 <4E9 <1.6E7 1.34E8 1.46E10
49 Thin blank 0-2 1.33E7 S5.09E7 14E11  <9ES <1.0E6 1.54E6 1.77E7 3.25E7 2.10E9 1.94E7 23E7 8.56E6 4.65E6
50 Thick blank 2-30 2.39E6 8.61E6 7.02E6 <3.1E9 <2ES <1.3E5 2.90E5 1.80E6 <I1ES8 4.69E6 4.96E6 1.07E6 1.79E6
51 Top knob 2.24E10 1.94E10 <1E12 <2E8 <8E7 1.0E8 <T7E8 <1E10 <3E7 247E9 "
KD Top chips 0-30 2.83E10 2.19E11 8.8EI0 5.82E12 <5.7E8 3200 <24E8 1.0E9 1.17E10 <3E9 <3.2E10 1.6E8 1.06E8 1.27E10

3.7E8 3.3E9 2.0E9 8.3E10 9ES 1.3E8. 6.2E9 6.1E9 $.9E10 9.9E10 9.6E9

9The number is the number of cut toward the interior starting at the graphite surface. “FC” stands for a fuel channel surface, “E” for a narrow edge surface, and “Deep” for a
first cut about 62 mils deep from a fuel channel surface. T, M, and B represent samples from the top, middle, and bottom specimens of the graphite stringer, respectively.

174!
125

Table 11.5. Fission products in MSRE graphite core bar after removal in cumulative values of ratio to inventory

Location Type ]();flts]; 137¢s 90g; 144 ce 957 9Nb  106Ry 1277 125g,
Top Channet 0-2 0.0008 0.007 0.0007 0.24°  0.11 0.024 0.41
0—-800 0.087 0.072 0.0032 0.0024 0.69 0.14 0.090 0.50
Edge 0-2 0.0008 0.006 0.0005 0.0007 0.12  0.037 0.036 0.054
Edge 0-62 0.0007 0.038 : 0.30 0.38 0.68 0.79
Middle Channel 0-3 0.0029 0.030 0.002 0.31 0.05 0.10 0.11
0-550 0.071  0.003 0.12
Edge 0-3 0.013 0.014 0.0008 0.0008 0.15 0.021 0.015 0.24
0-36 0.020 0.029 0.0011 0.0009 0.26 0.024 0.018 0.25
Edge 0-62 0.030 0.075  0.0004 1.00 1.16 0.94 2.14
Edge 0-62 0.037 0.081 0.0021 0.0023 0.53 0.29 0.60 0.79
Bottom Channel 0-3 0.0015 0.014 0.0012 0.0011 0.49 0.09 0.36 0.68
0-800 0.069 0.017 0.0024 0.0011 0.58 0.14 0.43 0.84
Edge 0-3 0.0017: 0.006 0.0015 0.0015 045 0.07 0.14 0.27

Inventory (dis/min per gram of salt)

6.2E9 6.1E9 S9E10 9.9E10 8.3E10 3.3E9 2.0E9  3.7EB8

“Includes significant < values.

values appear to be higher than those reported in the
next section for metal surfaces, but the passage of a
dozen half-lives between shutdown and determination
may have reduced the precision of the data.

11.3 Examination of Heat Exchangers and Control
Rod Thimble Surfaces

Among the specimens of component surfaces excised
from the MSRE were two segments of control rod
thimble and one each of heat exchanger shell and
tubing. In addition to the fission product data we
report here, these specimens were of major interest
because of grain-boundary cracking of surfaces con-
tacting molten salt fuel, as discussed elsewhere.’

Successive layers were removed electrochemically

from the samples using methanol—10% HCl as electro-
lyte. These solutions were examined both spectro-
graphically and radiochemically. The sample depth was
calculated from the amount of nickel and the specimen
geometry, etc. Concentration profiles (relative to
nickel) for the fuel side of the specimen of heat
exchanger tube are shown in Fig. 11.5 for constituent
elements, fission product tellurium, and various fission
product nuclides. It is of interest to note that concen-
trations 3 mils below the surface were 1 to 10% of
those observed near the surface. Values this high could

be the result of a grain-boundary transport of the
nuclides. :

For the various samples, deposit concentrations were
obtained by summing the amounts determined in
successive layers. These are shown, expressed as ratios
to inventory concentration divided by total MSRE area
(3.0 X 10% ¢m?), in Table 11.6.

As discussed for surveillance specimen and other
deposit data, this ratio, the relative deposit intensity,
permits direct comparison with deposits on other
surfaces, for example, graphite. To calculate the frac-

tion of inventory that deposits on a particular kind of

surface, the relative deposit intensity should be multi-
plied by the fraction of MSRE surface represented by
the deposit. The metal surface was 26% of the total
MSRE surface.

The most strongly deposited nuclides were ' 2° Sb and
'27Te, with most values above 1, indicative of a

selective strong deposition. The variation in values from

surface to surface suggests that generalizing any value to
represent all metal surfaces will probably not be very
accurate. Also, no values are available for certain large

.areas — in particular, the reactor vessel surfaces. In spite
.of all these caveats, the data here for tellurium and

antimony are consistent with similar data from the
various surveillance specimens in indicating that these
elements are very strongly deposited on metal surfaces,
as well as somewhat less strongly on graphite.

It is sufficient here to note that less strong but
appreciable deposition of ruthenium, technetium, and
niobium was found.

ORNL-DWG 7{- 7106
HEAT EXCHANGER TUBE

{ 2 3

00 Lme——e—ere—e—Ni—
? o/‘""—_.—_. *- Mo E
10! - ,.——4-—9———‘--: o— CFre_E
102 —
B~ . o . n 3
2 -3 I \.- 7]
0 I
B \ © ° T :
SRy o ¢
- S~ T
f’ E \-%Tc E
& 1075 | kN | -
Z =\, N\ | 3
= N {ppm _
= 6 I\ -\- -A'tr [ 7
« 10 N #1370, 3
= = e o 125 -

< 2 \ Sb
T 2T a\‘ e | -
NG R
-8 | > P%r_]|
10~ = 27, 3

-\, ] \»
0 N
- =
-10 ¢ ol e
0" <, 9b =
0 { 2 3 4
DEPTH (mils)

Fig. 11.5. Concentration profiles from the fuel side of an
MSRE heat exchanger tube measured about 1.5 years after
reactor shutdown. (Arrows indicate level was less than or equal
to that given.)

126

11.4 Metal Transfer in MSRE Salt Circuits

Cobalt-60, is formed in Hastelloy N by neutron
activation of the minor amount of %° Co (0.09%) put in
the alloy with nickel; the detection of ®°Co activity in
bulk metal serves as a measure of its irradiation history,
and the detection of °°Co activity on surfaces should
serve as a measure of metal transport from irradiated
regions. Cobalt-60 deposits were found on segments of
coolant system radiator tube, on heat exchanger tubing,
and on core graphite removed from .the MSRE in

. January 1971.

The activity found on the radiator tubing (which
received a completely negligible neutron dosage) was
about 160 dis min™' cm 2. This must have been

-transported by coolant salt flowing through heat

exchanger tubing activated by delayed neutrons in the
fuel salt. The heat exchanger tubing exhibited sub-
surface activity of about 3.7 X 10® dis/min per cubic
centimeter of metal, corresponding to a delayed neu-
tron flux in the heat exchanger of about 1 X 10'°. If
metal were evenly removed from the heat exchanger
and evenly deposited on the radiator tubing throughout
the history of the MSRE, a metal transfer rate at full
power of about 0.0005 mil/year is indicated.

Cobalt-60 activity in excess of that induced in the
heat exchanger tubing was found on the fuel side of the
tubing (3.1 X 10° dis min ™" ¢m™2) and on the samples
of core graphite taken from a fuel channel surface (5 X
10% to 3.5 X 107 dis min~! ¢m™2). The higher values
on the core graphite and their consistency with fluence
imply that additional activity was induced by core
neutrons acting on >?Co after deposition on the
graphite.

The reactor vessel (and annulus) walls are the major

~ metal regions subject to substantial neutron flux. If

these served as the major source of transported metal
and if this metal deposited evenly on all surfaces, a
metal loss rate at full power of about 0.3 mil/year is
indicated. Because deposition occurred on both the
hotter graphite and cooler heat exchanger surfaces,
simple thermal transport is not indicated. Thermo-
dynamic arguments preclude oxidation by fuel.

One mechanism for the indicated metal transport
might have 10% of the 1.5 W/cm?® fission fragment
energy in the annular fuel within a 30-u range deposited
in the metal and a small fraction of the metal sputtered
into the fuel. About 0.4% of the fission fragment
energy entering the metal resulting in such transfer
would correspond to the indicated reactor vessel loss
rate of 0.3 mil/year. If this is the correct mechanism,
reactors operating with higher fuel power densities

adjacent to metal should exhibit proportionately higher
loss rates.

11.5 Cesium Isotope Migration in MSRE
Graphite

Fission product concentration profiles were obtained
on the graphite bar from the center of the MSRE core
which was removed early in 1971. The bar had been in
the core since the beginning of operation; it thus was
possible to obtain profiles for 2.1-year '3?Cs (a
neutron capture product of the stable '?>Cs daughter
of 5.27-day '33Xe) as well as the 30-year !37Cs
daughter of 3.9-min '37Xe. These profiles, extending
to the center of the bar, are shown in Fig. 11.6.

Graphite surveillance specimens exposed for shorter
periods, and of thinner dimensions, have revealed
similar profiles for !27Cs.® Some of these, along with
profiles of other rare-gas daughters, were used in an
analysis by Kedl” of the behavior of short-lived noble
gases in graphite. Xenon diffusion and the possible
formation of cesium carbide in molten-salt reactors
have been considered by Baes and Evans.®

An appreciable literature on the behavior of fission
product cesium in nuclear graphite has been developed
in studies for gas-cooled reactors by British investi-
gators, the Dragon Project, Gulf General Atomic
workers, and workers at ORNL.

The profiles shown in Fig. 11.6 indicate significant
diffusion of cesium atoms in the graphite after their
formation. The 5.27-day half-life of '®3Xe must have
resulted in a fairly even concentration of this isotope
throughout the graphite and must have produced a flat
deposition profile for '33Cs. This isotope and its
neutron product ! 3*Cs could diffuse to the bar surface
and could be taken up by the salt. The !3*Cs profile

127

shows that this occurred. The ! ®*Cs concentration that
would accumulate in graphite if no diffusion occurred
has been estimated from the power history of the
MSRE to be about 2 X 10'*® atoms per gram of
graphite (higher if pump bowl xenon stripping is
inefficient), assuming the local neutron flux was a
minimum of four times the average core flux. The
observed ' 2*Cs concentration in the bar center, where
diffusion effects would be least, was about 5 X 10'*
atoms of '3*Cs per gram of graphite. The agreement is
not unreasonable.

Data for 30-year '37Cs are shown in Fig. 11.6. For
comparison, the accumulated decay profile of the
parent 3.9-min !37Xe is shown as a dotted line in the

ORNL-DWG 71-13199

1018 _
= ) E
i LACCUMULATED '*"xe =
'\’/ DISINTEGRATION (est) —
o 10'7 |= =
= E\ =
5 =\ =
o \ —
< 106 \ 137 _
: Al _ =
S 15 _: I\\ |
QL_) 10 = 4 \ -
a = LJ =
o o ] =
g [ \ _
5 10" = 13404 =
D [— j—
g — b
5 BAR CENTER LINE —
@ 43

10 =
12 7

0 100 200 300 400 500 600 700 800

DEPTH (mils)

Fig. 11.6. Concentration of cesium isotopes in MSRE core
graphite at given distances from fuel channel surface.

Table 11.6. Fission products on surfaces of Hastelloy N after termination
of operation expressed-as (observed dis min 1 em 2)/(MSRE inventory/total
MSRE surface area)

Surface inventory *SNb e '®Re %Ry ?TTe gy
Control rod thimble, bottom 0.14 1.2 1.5 0.50 1.65 33
Control rod thimble, middle 1.0 0.73 0.58 042 0.51 1.4
Mist shield outside, liquid 0.26 0.73 0.27 0.38 0.89 2.8
Heat exchanger, shell 0.33 1.0 0.10 0.19 1.4 2.6
Heat exchanger, tube 0.27 1.2 0.11 0.54 2.6 4.3

MSRE inventory divided by total MSRE surface area (dis min 1 em _2)
8.3E10 9E5 3.3E10 3.3E9 2.0E9

3.7E8

upper left of the figure. This was estimated assuming
that perfect stripping occurred in the pump bowl with a
mass transfer coefficient from salt to central core
graphite of 0.3 ft/hr® and a diffusion coefficient of
xenon in graphite (10% porosity) of 1 X 107°
cm?fsec.'® :

Near the surface the observed '27Cs profile is lower
than the estimated deposition profile; toward the center
the observed profile tapers downward but is about the
es imated deposition profile. This pattern should de-
velop if diffusion of cesium occurred. The central
concentration is about one-third of that near the
surface. Steady diffusion into a cylinder'! from a
constant surface source to yield a similar ratio requires
that Dt/r? be about 0.14. For a cylinder of 2 ¢cm radius
and a salt circulation time of 21,788 hr, a cesium
diffusion coefficient of about 7 X 107° cm?/sec is
indicated.

Data developed for cesium-in-graphite relationships in
gas-cooled reactor systems'? at temperatures of 800 to
1100°C may be extrapolated to 650°C for comparison.
The diffusion coefficient thereby obtained is slightly
below 1071 °; the diffusion coefficient for a gas (xenon)
is about 1 X 1075 c¢m?/sec. Some form of surface
diffusion of cesium seems indicated. This is further
substantiated by the sorption behavior reported by
Milstead! ® for the cesium—nuclear-graphite system.

In particular, Milstead has shown that at temperatures
of 800 to 1100°C and concentrations of 0.04 to 1.6 mg
of cesium per gram of graphite, cesium sorption on
graphite follows a Freundlich isotherm (1.6 mg of
cesium on 1 m? of graphite surface corresponds to the
saturation surface compound CsCg). Below this, a
Langmuir isotherm is indicated. In the MSRE graphite
under consideration the **7?Cs content was about 10" ®
atoms per gram of graphite, and the 133 and 135
chains would provide similar amounts, equivalent to a
total content of 0.007 mg of cesium per gram of
graphite. At 650°C, in the absence of interference from
other adsorbed species, Langmuir adsorption to this
concentration should occur at a cesium partial pressure
of about 2 X 107'® atm. At this pressure, cesium
transport via the gas phase should be negligible, and
surface phenomena should control.

To some extent, rubidium, strontium, and barium
atoms also are indicated to be similarly adsorbed and
likely to diffuse in graphite.

It thus appears that for time periods of the order of a
year or more the alkali and alkaline earth daughters of
noble gases which get into the graphite can be expected
to exhibit appreciable migration in the moderator
graphite of molten-salt reactors.

128

11.6 Noble-Metal Fission Transport Model

It was noted elsewhere'? that noble metals in MSRE
salt samples acted as if they were particulate con-
stituents of a mobile “pool” of such substances held up
in the system for a substantial period and that evidence
regarding this might be obtained from the activity ratio
of pairs of isotopes.

Pairs of the same element, thereby having the same
chemical behavior {e.g., ! °*Ru and ! ®4Ru), should be
particularly effective. As produced, the activity ratio of
such a pair is proportional to ratios of fission yields and
decay constants. Accumulation over the operating
history yields the inventory ratio, ultimately propor-
tional (at constant fission rate) only to the ratio of
fission yields. If, however, there is an intermediate
holdup and release before final deposition, the activity
ratio of the retained material will depend on holdup
time and will fall between production and inventory
values. Furthe_rmore,' the material deposited after such a
holdup will, as a result, have ratio values lower than
inventory. (Values for the isotope of shorter half-life,
here '°3Ru, will be used in the numerator of the ratio
throughout our discussion.) Consequently, comparison
of an observed ratio of activities (in the same sample)
with associated production and inventory ratios should
provide an indication of the “accumulation history™ of
the region represented by the sample. Since both
determinations are for isotopes of the same element in
the same sample (consequently subjected to identical
treatment), many sampling and handling errors cancel
and do not affect the ratio. Ratio values are thereby
subject to less variation. ’ '

We have used the ' °2Ru/' °®Ru activity ratio, among
others, to examine samples of various kinds taken at
various times in the MSRE operation. These include salt
and gas samples from the pump bowl and other
materials briefly exposed there at various times.

Data are also avaijlable from the sets of surveillance
specimens removed from runs 11, 14, 18, and 20.
Materials removed from the off-gas line after runs 14
and 18 offer useful data. Some information is avail-

able! 3 from the on-site gamma spectrometer surveys of

the MSRE following runs 18 and 19, particularly with
regard to the heat exchanger and off-gas line.

11.6.1 Inventory and model. The data will be dis-
cussed in terms of a “compartment” model, which will
assign first-order transfer rates common for both
isotopes between given regions and will assume that this
behavior was consistent throughout MSRE history.
Because the half-lives of '®*Ru and '°¢Ru are quite
different, 39.6 and 367 days, respectively, appreciably
different isotope activity ratios are indicated for differ-
ent compartments and times as simulated operation
proceeds. A sketch of a useful scheme of compartments
is shown in Fig. 11.7.

We assume direct production of '®*Ru and !°®Ru in
the fuel salt in proportion to fission rate and fuel
composition as determined by MSRE history. The
material is fairly rapidly lost from salt either to
“surfaces’ or to a mobile “particulate pool” of agglom-
erated material. The pool loses material to one or more
final repositories, nominally ‘‘off-gas,” and also may
deposit material on the “surfaces.” Rates are such as to
result in an appreciable holdup period of the order of
50 to 100 days in the “particulate pool:” Decay, of
course, occurs in all compartments.

Material is also transferred to the “drain tank™ as
required by the history, and transport between com-
partments ceases in the interval.

From the atoms of each type at a given time in a
given compartment, the activity ratio can be calculated,
as well as an overall inventory ratio.

We shall identify samples taken from different regions
of the MSRE with the various compartments and thus
obtain insight into the transport paths and lags leading
to the sampled region. It should be noted that a
compartment can involve more than one region or kind
of sample. The additional information required to
establish the amounts of material to be assigned to a
given region, and thereby to produce a material balance,
is not available.

129

In comparison with the overall inventory value of
103Ru/ ®®Ru, we should expect “surface” values to
equal it if the deposited material comes rapidly and
only from “salt” and to be somewhat below it if, in
addition, “particulate” is deposited. If there is no direct
deposition from “salt” to “surface,” but only “particu-
late,” then deposited material should approach “off-
gas” compartment ratio.

The “off-gas” compartment ratio should be below
inventory, since it is assumed to be steadily deposited
from the ‘“particulate pool,” which is richer in the
long-lived '°SRu component than production, and
inventory is the accumulation of production minus
decay.

The particulate pool will be above inventory if
material is transferred to it rapidly and lost from it at a
significant rate. Slow loss rates correspond to long
holdup periods, and ratio values tend toward inventory.

Differential equations involving proposed transport,
accumulation, and decay of !°3Ru and !°¢Ru atoms
with respect to these compartments were incorporated
into a fourth-order Runge-Kutta numerical integration
scheme which was operated over the full MSRE power
history. ' :

The rates used in one calculation referred to in the

“discussion below show rapid loss (less than one day)

from salt to particulates and surface, with about 4%
going directly to surface. Holdup in the particulate pool
results in a daily transport of about 2% per day of the
pool to “off-gas,” for an effective average holdup

ORNL-DWG 70-¢3503

SURFACES

FISSION SALT

DECAY

PARTICULATE
POOL

OFF - GAS
(AND OTHER SINKS)
FOR PARTICLES

(t1)

DRAIN TANKS
(AND HEEL)

{ALL TRANSPORT CEASES DURING
PERIOD THAT SALT IS DRAINED
FROM SYSTEM)

I

Fig. 11.7. Compartment model for noble-metal fission transport in MSRE.
period of about 45 days. All transport processes are
assumed irreversible in this scheme. '

11.6.2 Off-gas line deposits. Data were reported in
Sect. 10 on the examination after run 14 (March 1968)
of the jumper line installed after run 9 (December
1966), on the examination after run 18 (June 1969) of
parts of a specimen holder assembly from the main
off-gas line installed after run 14, and on the exami-
nation of parts of line 523, the fuel pump overflow
tank purge gas outlet to the main off-gas line, which
was installed during original fabrication of MSRE.
These data are shown in Table 11.7.

For the jumper line removed after run 14, observed
ratios range from 2.4 to 7.3. By comparison the
inventory ratio for the net exposure interval was 12.1.
If a holdup period of about 45 days prior to deposition
in the off-gas line is assumed, we calculate a lower ratio
for the compartment of 7.0. It seems indicated that a
holdup of ruthenium of 45 days or more is required.

Ratio values from the specimen holder removed from
the 522 line after run 18 ranged between 9.7 and 5.0.
Net inventory ratio for the period was 19.7, and for
material deposited after a 45-day holdup, we estimated
a ratio of 12.3. A longer holdup would reduce this
estimate. However, we recall that gas flow through this
line was appreciably diminished during the final month
of run 18. This would cause the observed ratio values to
be lower by an appreciable factor than would ensue
from steady gas flow all the time. A holdup period of
something over 45 days still appears indicated.

Flow of off-gas through line 523 was less well known.
In addition to bubbler gas to measure salt depth in the
overflow tank, part of the main off-gas flow from the
pump bowl went through the overflow tank when flow
through line 522 was hindered by deposits. The
observed ratio after run 18 was 8 to 13.7, inventory was
9.6, and for material deposited after 45 days holdup the
ratio is calculated to be 5.8. However, the unusually
great flow during the final month of run 18 (until
blockage of line 523 on May 25) would increase the
observed ratio considerably. This response is consistent
with the low value for material from line 522 cited
above. So we believe the assumption of an appreciable
holdup period prior to deposition in off-gas regions
remains valid. _

11.6.3 Surveillance specimens. Surveillance speci-
mens of graphite and also selected segments of metal
were removed from the core surveillance assembly after
exposure throughout several runs. Table 11.8 shows
values of the activity ratios for *®3Ru/*®®Ru for a
number of graphite and metal speciméns removed on
different occasions in 1967, 1968, and 1969. Insofar as

130

deposition of these isotopes occurred irreversibly and
with reasonable directness soon after fission, the ratio
values should agree with the net inventory for the
period of exposure, and the samples that had been
exposed longest at a given removal time should have
appropriately lower values for the ! °3Ru/! °%Ru activ-
ity ratio.

Examination of Table 11.8 shows that this latter view
is confirmed — the older samples do have values that are
lower, to about the right extent. However, we also note
that most observed ratio values fall somewhat below the
net inventory values. This- could come about if, in
addition to direct deposition from salt onto surfaces,
deposition also occurred from the holdup “‘pool,”
presumed colloidal or particulate, which was mentioned
in the discussion of the off-gas deposits. Few of the
observed values fall below the parenthesized off-gas
values. This value was calculated to result if all the
deposited material had come from the holdup pool.

Table 11.7. Ruthenium isotope activity ratios
of off-gas line deposits
)d

dis/min '23Ruy
Calculated

Observed vs calculated (
) dis/min 19%Ry

Sample Observed

I. Jumper Section of Line 522, Exposed December 1966
to March 25, 1968 .

Flextoo! 2.4 Production? = 58
Upstream hose 24 Net inventory 12.1
Downstream hose 4.1

Inlet dust 7.3 Net deposit if:
Outlet dust 44 45-day holdup 7.0

90-day holdup 5.5

I1. Specimen Holder, Line 522, Exposed August 1968
to June 1969

Bait end 9.7 Production® 43
Flake 5.0 Net inventory 19.7
- Tube sections 54,59
Recount 7/70, corr 6,2 Net deposit if:

45-day holdup 12.3
90-day holdup 9.3

1. Overflow Tank Off-Gas Line (523), Exposed 1965
to June 1969

Valve V523 8.0 Inventory 9.6
Line 523 11.3,13.7, 13.3
"Valve HCV 523 13.3, 12.5, 10.0p{Net deposit if:

45-day holdup 5.8
90-day holdup 4.3

Corrected to time of shutdown.

b2351.238 fyel, with inbred 23%Pu.

€233y fuel, including 2.1% of fissions from contained 235U
and 4.3% from 239Pu.
131

Table 11.8. Ruthenium isotope activity ratios of surveillance specimens

Observed vs calculated(

dis/min '03Ru>

dis_/min 106py

Observed Ratios,

Calculated Values of Ratio

Exposure, Material
Runs Median Undeslined Plus Deposition
Netlnventory ¢ o Holdup (45 dayy O Gas
8-11 Graphite 20,922,234 25,227, 52 25.5 19.2 (15.9)
8-14 Graphite 9,512 11.4 8.4 (6.8)
12-14 Graphite 11,12,9(>12), 13,2 (>13), 14, (>14), 16, 17 16.7 11.6 (9.3)
Metal 10,211,125 15
8—18 Graphite 601778 9.8 7.2 (5.7
Metal basket  (3.5)? '
15-18 Graphite 2,10,11,2 12,213, 14, 15,2 16, 21, 27 19.7 14.9 (12.3)
Metal 6,7,28,9,10 '
19-20  _ Graphite 10,211, 12,2 13,% 14, 3200 21.7 13.4 (9.6)
Metal 6,7,8,10911,212,2 13,215

4s. S. Kirslis and F. F. Blankenship, MSR Program Semiann. Progr. Repr. Aug. 31, 1968, ORNL-4344 pp. 115-41.
bs. S. Kirslis and F. F. Blankenship, MSR Program Semiann. Progr. Repr. Aug. 31, 1967, ORNL4191, pp. 121-28.

€F. F. Blankenship, personal communication.

dg. 1. Compere, MSR Program Semiann. Progr. Rept. Aug. 31, 1968, ORNL-4344, pp. 206—-10.

°F. F. Blankenship, S. S. Kirslis, and E. L. Compese, MSR Program Semiann. Progr. Rept. Aug. 31, 1969, ORNL-4449, pPp.

104—7; Feb. 28, 1970, ORNL-4548, pp. 10410,

Therefore we conclude that the surface deposits did not
occur only by deposition of material from the “particu-
late pool”; calculated values are shown which assume
rates which would have about two-thirds coming from a
particulate pool of about 45 days holdup. The agree-
ment is not uncomfortable. .

In general, the metal segments showed lower values
than graphite specimens similarly exposed, with some
observed values below any corresponding calculated
values. This implies that in some way in the later part of
its exposure the. tendency of the metal surface to
receive and retain ruthenium isotope deposits became
diminished, particularly in comparison with the graph-
ite specimens. Also, the metal may have retained more
particulate and less directly deposited material than the
graphite, but on balance the deposits on both types of
specimen appear to have occurred by a combination of
the two modes.

11.6.4 Pump bowl samples. Ratio data are available
on salt samples and later gas samples removed from the
pump bowl spray shield beginning with run 7 in 1966.
Similar data are also available for other materials
exposed from time to time to the: gas or liquid regions
within the spray shield. Data from outer sheaths of
double-walled capsules are included. Data for the
activity ratios (dis/min '°3Ru)/(dis/min '®®Ru) are
shown for most of these in Fig. 11.8. In this figure the

activity ratios are plotted in sample sequence. Also
shown on the plot are values of the overall inventory
ratio, which was calculated from power history, and the
production ratio, which was calculated from yields
based on fuel composition. This changed appreciably
during runs 4 to 14, where the plutonium content
increased because of the relatively high 23®U content
of the fuel. The !°¢Ru yield from 23°Pu is more than
tenfold greater than its yield from 233U or 233 U. The
plutonium content of the fuel did not vary nearly as
much during the 233U operation and was taken as
constant. . A

Also shown are lines which have been computed
assuming a particulate pool with average retention
periods of 45 and 100 days. The point has been made
previously! 3 that noble-metal activity associated with
any materials exposed in or sampled from the pump
bowl is principally from this mobile pool rather than
being dissolved in salt or occurring as gaseous sub-
stances. Consequently, similar ratios should be en-
countered for salt and gas samples and surfaces of
various materials exposed in the pump bowl.

Examination of Fig. 11.8 indicates that the prepon-
derance of points fall between the inventory line and a
line for 45 days average retention, agreeing reasonably
well with an average holdup of between 45 and 100
days — but with release to off-gas, surfaces, and other

s
132

ORNL-DWG 70~-13504

72
Jrrrarerrerrrbrgrrgrrrrnrrer e ettt e Ty T T T T T T T i T T T T T I T I T T T T T e T T
68 - \ ® SALT SAMPLES INSIDE
© SALT SAMPLES OUTSIDE
A GAS SAMPLES INSIDE
& GAS SAMPLES OUTSIDE
64 N ? UNCERTAIN ANALYTICAL DATA
\ PRODUCTION
—-—— INVENTORY
--------- PARTICLES, 45 DAY AVG. HOLDUP
6o \ —--—— PARTICLES, 100 DAY AVG HOLDUP
AY O STAINLESS STEEL
¥ HASTELLOY N
© LIQUID
A X ¥ INTERFACE
56 o ® Gas
® OTHER ADDITIONS, EXPOSURES
OR LADLE SAMPLES
52 -
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48
o
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540 ) N T s -
= : S eI '
a [] ’ '
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[ v} v > o e [ B Ni
’ LY
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28 —~ — ' ‘\ ; B T
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A * * W : e, e g :
) L o | T et 7
/ \ o [ ! ! Sa
24 o— v ® = a 1 & Fefa i 2
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- A ] T e b E—— K *’
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20 “of — L RN SV 4 Sy Py = 25 a4 = o—— e — Y
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\ i La A / ° 0/ A
16 \ :' " A 2 % 4“_' R ) / a A ’/'
3\ : l ¢ ; “fa fl" a oA
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81e Pew! ( 7 o? g - i a — 7
H o . T Remmemas . "
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4 ;’I : — < &0 ._;- 4
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] ’/ o? ¢ . 4 A
. i, f
Uit T e la VAl T LR
2 11 20 8 22 46 51 53 58 26 20 63 67 3R S T 2 7 {7 25 2 6 14 19 29 46 6 13 45 19 23 28 36 38 42 346- 54 56 58 62 €5 73IC 77 79 9 (g 32
5 {2 22 12 45 SO S2 54 6 27 30 66 2B 42 ST 4 6 10 22 33 4 12 {5 20 42 1 9 14 1% 20 24 29C 37 44 44 47 55 57 59 €4C 70 TG 7B 1 12 27
SAMPLE NUMBER IN GIVEN RUN
i . v I v I ’ Frapel N I v - P — - . . R
RUN RUN RUN RUN 11 RUN  RUN 14 RUN1S  RUN RUN 17 RUN 18 RUN 19 RUN 20
7 8 40 12 16 :
RUN NUMBER

Fig. 11.8. Ratio of ruthenium isotope activities for pump bowl samples.

.
L ¥

133

regions resulting in a limited retention rather than the
unlimited retention implied by an inventory value.
Although meaningful differences doubtless exist be-

tween different kinds of samples taken from the pump

bowl, their similarity clearly indicates that all are taken
from the same mobile pool, which loses material, but
slowly enough to have an average retention period of
several months.

Discussion. The data presented above represent
practically all the ratio data available for MSRE
samples. The data based on gamma spectrometer
surveys of various reactor regions, particularly after
runs 18 and 19, have not been examined in detail but in
cursory views appear not too inconsistent with values
given here,

It appears possible to summarize our findings about
fission product ruthenium in this way:

The off-gas deposits appear to have resulted from the
fairly steady accumulation of material which had been
retained elsewhere for periods of the order of several
months prior to deposition.

The deposits on surfaces also appear to have con-
tained material retained elsewhere prior to deposition,
though not to quite the same extent, so that an
appreciable part could have been deposited soon after
fission.

All materials taken from the pump bowl contain
ruthenium isotopes with a common attribute: they are
representative of an accumulation of several months.
Thus all samples from the pump bowl presumably get
their ruthenium from a common source.

Since it is reasonable to expect fission products to
enter salt first as ions or atoms, presumably these
rapidly deposit on surfaces or are agglomerated. The
agglomerated material is not dissolved in salt but is
fairly well dispersed and may deposit on surfaces to
some extent. It is believed that regions associated with

‘the pump bowl — the liquid surface, including bubbles,

the shed roof, mist shield, and overflow tank — are
effective in accumulating this agglomerated material.
Regions with highest ratio of salt surface to salt mass
(gas samples containing mist and surfaces exposed to
the gas-liquid interface) have been found to have the
highest quantities of these isotopes relative to the
amount of salt in the sample. So the agglomerate seeks
the surfaces. Since the subsurface salt samples, however,
never show amounts of ruthenium in excess of in-
ventory, it would appear that material entrained,
possibly with bubbles, is fairly well dispersed when in
salt.

Loss of the agglomerated material to one or more
permanent sinks at a rate of 1 or 2% per day is
indicated. In addition to the off-gas system and to some

extent the reactor surfaces, these sinks could include
the overflow tank and various nooks, crannies, and
crevices if they provided for a reasonably steady
irreversible loss. :

Without additional information the ratio method
cannot indicate how much material follows a particular
path to a particular sink, but it does serve to indicate
the paths and the transport lags along them for the
isotopes under consideration.

References

1. E. L. Compere and E. G. Bohlmann, “Examination
of Deposits from the Mist Shield in the MSRE Fuel
Pump Bowl,” MSR Program. Semiannu. Prog. Rep. Feb.
28, 1971, pp. 76—85.

2. A. Houtzeel, private communication.

3. P. N. Haubenreich, 4 Review of Production and
Observed Distribution of Tritium in the MSRE in the
Light of Recent Findings, ORNL-CF-71-8-34 (Aug. 23,
1971). (Internal document — No further dissemination
authorized.)

4. R. B. Briggs, “Tritium in Molten Salt Reactors,”
Reactor Technol. 14, 335—42 (Winter 1971—-1972).

5. H. E. McCoy and B. McNabb, Intergranular Crack-
ing of INOR-8 in the MSRE, ORNL-4829 (November
1972).

6. S. S. Kirslis et al., “Concentration Profiles of
Fission Products in Graphite,” MSR Program Semiannu.
Progr. Rep. Aug. 31, 1970, ORNL-4622, pp. 69-70;
Feb. 28, 1970, ORNL-4548, pp. 104—5; Feb. 28, 1967,
ORNL-4119, pp. 125—28; Aug. 31, 1966, ORNL-4037,
pp. 180—84; D. R. Cuneo et al., “Fission Product
Profiles in Three MSRE Graphite Surveillance Speci-
mens,” MSR Program Semignnu. Progr. Rep. Aug. 31,
1968, ORNL-4344, pp. 142, 144—47; Feb. 29, 1968,
ORNL-4254, pp. 116, 118-22.

7. R. J. Kedl, A Model for Computing the Migration
of Very Short Lived Noble Gases into MSRE Graphite,
ORNL-TM-1810 (July 1967). '

8. C. F. Baes, Jr., and R. B. Evans III, MSR Program
Semiannu. Progr. Rep. Aug. 31, 1966, ORNL-4037, pp.
158-65.

9. R. B. Briggs, “Estimate of the Afterheat by Decay
of Noble Metals in MSBR and Comparison with Data
from the MSRE,” MSR-68-138 (Nov. 4, 1968). (Inter-
nal document — No further dissemination authorized.)

10. R. B. Evans III, J. L. Rutherford, and A. P.
Malinauskas, Gas Transport in MSRE Moderator Graph-

ite, II. Effects of Impregnation, Ill. Variation of Flow
Properties, ORNL-4389 (May 1969).

11. J. Crank, p. 67 in The Mathematics of Diffusion, -
Oxford University Press, London, 1956.
12. H.7J. de Nordwall, private communication.

13. C. E. Milstead, “Sorption Characteristics of the
Cesium-Graphite System at Elevated Temperatures and
Low Cesium Pressure,” Carbon (Oxford) 7, 199200
(1969).

14, E. L. Compere and E. G. Bohlmann, MSR
Program Semiannu. Progr. Rep. Feb. 28, 1970, ORNL-

134

4548, pp. 111—-18; see also Sect. 6, this report,
especially Fig. 6.5.

15. A. Houtzeel and F. F. Dyer, A Study of Fission
Products in the Molten Salt Reactor Experiment by
Gamma Spectrometry, ORNL-TM-3151 (August 1972).

135

12. SUMMARY AND OVERVIEW

A detailed synthesis of all the factors known to affect
fission product behavior in this reactor is not possible
within the available space. Many of the comments
which follow are based on a recent summary report.'

Operation of the MSRE provided an opportunity for
studying the behavior of fission products in an
operating molten-salt reactor, and every effort was
made to maximize utilization of the facilities provided,
even though they were not originally designed for some
of the investigations which became of interest. Sig-
nificant difficulties stemmed from:

1. The salt spray system in the pump bowl could not
be turned off. Thus the generation of bubbles and
salt mist was ever present; moreover, the effects
were not constant, since they were affected by salt
level, which varied continuously.

2. The design of the sampler system severely limited
the geometry of the sampling devices.

3. A mist shield enclosing the sampling point provided
a special environment.

4. Lubricating oil from the pump bearings entered the
pump bowl at a rate of 1 to 3 cm?/day.

5. There was continuously varying flow and blowback
of fuel salt between the pump bowl and an over-
flow tank.

In spite of these problems, useful information con-
cerning fission product fates in the MSRE was gained.

12.1 Stable Salt-Soluble Fluorides

12.1.1 Salt samples. The fission products Rb, Cs,
Sr, Ba, the lanthanides and Y, and Zr all form
stable fluorides which are soluble in fuel salt. These
fluorides would thereby be expected to be
found completely in the fuel salt except in those cases
where there is a noble-gas precursor of sufficiently long
half-life to be appreciably stripped to off-gas. Table
12.1 summarizes data from salt samples obtained during
the 233U operation of the MSRE for fission products
with and without significant noble-gas precursors. As
expected, the isotopes with significant noble-gas pre-
cursors (3°Sr and '37Cs) show ratios to calculated
inventory appreciably lower than those without, which
generally scatter around or somewhat above 1.0.

12.1.2 Deposition. Stable fluorides showed little
tendancy to deposit on Hastelloy N or graphite.
Examinations of surveillance specimens exposed in the
core of the MSRE showed only 0.1 to 0.2% of the
isotopes without noble-gas precursors on graphite and
Hatselloy N. The bulk of the amount present stemmed
from fission recoils and was generally consistent with
the flux pattern. ‘

However, the examination of profiles and deposit
intensities indicated that nuclides with noble-gas pre-
cursors were deposited within the graphite by the decay
of the noble gas that had diffused into the relatively
porous graphite. Clear indication was noted of a further

Table 12.1, Stable fluoride fission product activity as a fraction of calculated

inventory in salt sampies from

233 .
U operation

Without significant noble-gas precursor

With noblegas precursor

Nuclide 957, 141, 144, 1474 89¢, 137 9ty 140p.
Weighted yield, g 6.01 6.43 4.60 1.99 5.65 6.57 5.43 5.43
Half-life, days 65 33 284 11.1 52 30 yr. 58.8 12.8
Noble-gas precursor 89kr 137x%e 1Ky 140xe
Precursor half-life 3.2 min 3.9 min 9.8 sec 16 sec
Activity in salt?

Runs 15-17 0.88-1.09 0.87-1.04 1.14-1.25 0.99-1,23 0.67-097 0.82-093 0.83-146 0.82--1.23

Run18 1.05-1.09 0.95-099 1.16-1.36 0.82-1.30 0.84-0.89 0.86-099 1.16-1.55 1.10-1.20

Runs 19-20 0.95-1.02 0.89-1.04 1.17-1.28 1.10-1.34 0.81-0.98 1.02-1.20

0.70-0.95

1.13-1.42

“Allocated fission yields: 93.2% 223U, 2.3% 235U, 4.5% 23°Pu.
bAs fraction of calculated inventory. Range shown is 25-75 percentile of sample; thus half the sample values fall within this

range.
diffusion of the relatively volatile cesium isotopes, and
possibly also of Rb, Sr, and Ba, after production within
graphite.

12.1.3 Gas samples. Gas samples obtained from the
gas space in the pump bowl mist shield were consistent
with the above results for the salt-seeking isotopes with
and without noble-gas precursors. Table 12.1 shows the
percentages of these isotopes which were estimated to
be in the pump bowl stripping gas, based on the
amounts found in gas samples. Agreement with ex-

pected amounts where there were strippable noble-gas’

precursors is satisfactory considering the mist shield,
contamination problems, and other experimental dif-
ficulties. Gamma spectrometer examination of the
off-gas line showed little activity due to salt-seeking
isotopes without noble-gas precursors. Examinations of
sections of the off-gas line also showed only small
amounts of these isotopes present.

12.2 Noble Metals

The so-called noble metals showed a tantalizingly
ubiquitous behavior in the MSRE, appearing as salt-
borne, gas-borne, and metal- and graphite-penetrating
species. Studies of these species included isotopes of
Nb, Mo, Tc, Ru, Ag, Sb, and Te.

12.2.1 Salt-borne. The concentrations of five of the
noble-metal nuclides found in salt samples ranged from
fractions to tens of percent of inventory from sample to
sample. Also, the proportionate composition of these
isotopes remained relatively constant from sample to
sample in spite of the widely varying amounts found.
Silver-111, which clearly would be a metal in the MSRE
salt and has no volatile fluorides, followed the pattern
quite well and also was consistent in the gas samples.
This strongly supports the contention that we were
dealing with metal species.

These results suggest the following about the noble
metals in the MSRE.

1. The bulk of the noble metals remain accessible in
the circulating loop but with widely varying
amounts in circulation at any particular time.

. In spite of this wide variation in the total amount
found in a particular sample, the proportional
composition is relatively constant, indicating that
the entire inventory is in substantial equilibrium
with the new material being produced.

The mobility of the pool of noble-metal material
suggests that deposits occur as an accumulation of
finely divided, well-mixed material rather than as a
“plate.”

136

No satisfactory correlation of noble-metal concen-
tration in the salt samples and any operating parameter
could be found.

In order to obtain further understanding of this
particulate pool, the transport paths and lags of
noble-metal fission products in the MSRE were exam-
ined using ‘all available data on the activity ratio of two
isotopes of the same element, 39.6-day '°?Ru and
367-day '°®Ru. Data from graphite and metal sur-
veillance specimens exposed for various periods and

- removed at various times, for material taken from the

off-gas system, and for salt and gas samples and other
materials exposed to pump bowl salt were compared
with appropriate inventory ratios and with values
calculated for indicated lags in a simple compartment
model. This model assumed that salt rapidly lost
ruthenium. fission product formed in it, some to
surfaces and most to a separate mobile “pool” of
noble-metal fission product, presumably particulate or
colloidal and located to an appreciable extent in pump
bowl regions. Some of this “pool” material deposited
on surfaces and also appears to be the source of the
off-gas deposits. All materials sampled from or exposed
in the pump bowl appear to receive their ruthenium
activity jointly from the pool of retained material and
from more direct deposition as produced. Adequate
agreement of observed data with indications of the
model resulted when holdup periods of 45 to 90 days
were assumed.

12.2.2 Niobium. Niobium is the most susceptible of
the noble metals to oxidation should the U*/U3" ratio
be allowed to get too high. Apparently this happened at
the start of the 232U operation,!as was indicated by a
relatively sharp rise in Cr?" concentration; it was also
noted that 60 to about 100% of the calculated **>Nb
inventory was present in the salt samples. Additions of
a reducing agent (beryllium metal), which inhibited the
Cr*" buildup, also resulted in the disappearance of the
®5Nb from the salt. Subsequently the ?*Nb reappeared
in the salt several times for not always ascertainable
reasons and was caused to leave the salt by further .
reducing additions. As the 2>3U operations continued,
the percentage of ?>Nb which reappeared decreased,
suggesting both reversible and irreversible sinks. The
?5Nb data did not correlate closely with the Mo-Ru-Te
data discussed, nor was there any observable correlation
of its behavior with amounts found in gas samples.

12.2.3 Gas-borne. Gas samples taken from the pump
bowl during the 235U operation indicated con-
centrations of noble metals that implied that substantial
percentages (30 to 100) of the noble metals being
produced in the MSRE fuel system were being carried
out in the 4 liters (STP)/min helium purge gas. The data
obtained in the 223U operation with substantially
improved sampling techniques indicated much lower
transfers to off-gas. In both cases it is assumed that the
noble-metal concentration in a gas sample obtained
inside the mist shield was the same as that in the gas
leaving the pump bowl proper. (The pump bowl was
designed to minimize the amount of mist in the
sampling area and also at the gas exit port.) It is our
belief that the 233U period data are representative and
that the concentrations indicated by the gas samples
taken during 2°5U operation are anomalously high
because of contamination. This is supported by direct
examination of a section of the off-gas system after
completion of the 235U operation. The large amounts
of noble metals that would be expected on the basis of
the gas sample indications were not present. Appre-
ciable (10 to 17%) amorphous carbon was found in dust
samples recovered from the line, and the amounts of
noble metals roughly correlated with the amounts of
carbon. This suggests the possibility of noble-metal
absorption during cracking of the oil.

In any event the gas transport of noble metals appears
to have been as constituents of particulates. Analysis of

137

the deposition of flowing aerosols in conduits de-

veloped relationships between observable deposits and
flowing concentrations or fractions of production to
off-gas for diffusion and thermophoresis mechanisms.
The thermophoretic mechanism was indicated to be
dominant; the fraction of noble-metal production car-
ried into off-gas, based on this mechanism, was slight
(much below 1%).

12.3 Deposition in Graphite and Hastelloy N

The results from core surveillance specimens and from
postoperation examination of components revealed
that differences in deposit intensity for noble metals
occurred as a result of flow conditions and that deposits
on metal were appreciably heavier than on graphite,
particularly for tellurium and its precursor antimony.

The final surveillance specimen array, exposed for the
last four months of MSRE operations, had graphite and
metal specimens matched as to configuration in varied
flow conditions. The relative deposition intensities (1.0
if the entire inventory was spread evenly over all
surfaces) were as shown in Table 12.2. '

The examination of some segments excised from
particular reactor components, including core metal and
graphite, pump bowl, and heat exchanger surfaces, one

Table 12.2. Relative deposition intensities for noble metals

Deposition intensity

Surface Flow regime 95Nb 99M0 99TC 103R].1 106Ru lszb 129mTe 132Te
Surveillance specimens
Graphite Laminar 0.2 0.2 0.06 0.16 0.15
Turbulent 0.2 0.04 0.10 0.07
Metal Laminar 0.3 0.5 0.1 0.3 0.9
Turbulent 0.3 1.3 0.1 0.3 2.0
Reactor components
Graphite
Core bar channel Turbulent
Bottom 0.54 0.07 0.25 0.65 0.46%
Middle 1.09 1.06 1.90 0.924
Top 0.23 0.29 0.78 0.62%.
Metal ,
Pump bowl Turbulent 0.26 0.73 0.27 0.38 2.85 . 0.89°
Heat exchanger shell Turbulent 0.33 1.0 0.10 0.19 2.62 1.35%
Heat exchanger tube Turbulent 0.27 1.2 0.11 0.54 4.35 2.57¢
Core
Rod thimble _
Bottom Turbulent 142 1.23 1.54 0.50 327 1.65%
Middle Turbulent 1.00 0.73 0.58 0.42 1.35 " 0.544

a12'7Te
138

year after shutdown also revealed appreciable accumu-
lation of these substances. The relative deposition

intensities at these locations are also shown in Table.

12.2.

It is evident that net deposition generally was more
intense on metal than on- graphite, and for metal was
more intense under more turbulent flow. Surface
roughness had no apparent effect.

Extension to all the metal and graphite areas of the
system would require knowledge of the effects of flow
conditions in each region and the fraction of total area
represented by the region. (Overall, metal area was 26%
of the total and graphite 74%.)

Flow effects have not been studied experimentally;
theoretical approaches based on atom mass transfer
through salt boundary layers, though a useful frame of
reference, do not in their usual form take into account
the formation, deposition, and release of fine particu-
late material such as that indicated to have been present
in the fuel system. Thus, much more must be learned
about the fates of noble metals in molten-salt reactors
before their effects on various operations can be
estimated reliably.

Although the noble metals are appreciably deposited
on graphite, they do not penetrate any more than the
salt-seeking fluorides without noble-gas precursors.

The more vigorous deposition of noble-metal nuclides
on Hastelloy N was indicated by postoperation
examination to include penetration into the metal to a
slight extent. Presumably this occurred along grain-
boundary cracks, a few mils deep, which had developed
during extended operation, possibly because of the
deposited fission product tellurium.

12.4 lodine

The salt samples indicated considerable '>!1 was not
present in the fuel, the middle quartiles of results
ranging from 45 to 71% of inventory with a median of

62%. The surveillance specimens and gas samples
accounted for less than 1% of the rest. The low

tellurium material balances suggest the remaining *3'1
was permanently removed from the fuel as !'3!Te
(half-life, 25 min). Gamma spectrometer studies indi-
cated the *2'I formed in contact with the fuel returned
to it; thus the losses must have been to a region or
regions not in contact with fuel: This strongly suggests
off-gas, but the iodine and tellurium data from gas
samples and examinations of off-gas components do not
support such a loss path. Thus, of the order of
one-fourth to one-third of the iodine has not been
adequately accounted for.

It is recognized that, as shown in gas-cooled reactor
studies, ~° fission product iodine may be at partial
pressures in off-gas helium that are too low for iodine
to be fixed by steel surfaces at temperatures above
about 400°C. However, various off-gas surfaces at or
downstream from the jumper line outlet were below
such temperatures and did not indicate appreciable
iodine deposition. Combined with low values in gas
samples, this indicates little iodine transport to off-gas.

12.5 Evaluation

The experience with the MSRE showed that the noble
gases and stable fluorides behaved as expected based on
their chemistry. The noble-metal behavior and fates,
however, are still in part a matter of conjecture. Except
for niobium under unusually oxidizing conditions, it
seems clear that these elements are present as metals
and that their ubiquitous properties stem from that fact
since metals are not wetted by, and have extremely low
solubilities in, molten-salt reactor fuels. Unfortunately
the MSRE observations probably were substantially
affected by the spray system, oil cracking products, and
flow to and from the overflow, all of which were
continuously changing, uncontrolled variables. The low
material balance on '3'I indicates appreciable unde-
termined loss from the MSRE, probably as a noble-
metal precursor (Te, Sb).

Table 12.3 shows the estimated distribution of the
various fission products in a molten-salt reactor, based
on the MSRE studies. Unfortunately the wide variance
and poor material balances for the data on noble metals
make it unrealistic to specify their fates more than
qualitatively. As a consequence, future reactor designs
must allow for encountering appreciable fractions of
the noble metals in all regions contacted by circulating
fuel. As indicated in the table, continuous chemical
processing and the processes finally chosen will sub-
stantially affect the fates of many of the fission
products.

References
1. M. W. Rosenthal, P. N. Haubenreich, and R. B.
Briggs, The Development Status of Molten Salt Breeder
Reactors, ORNL-4812 (August 1972), especially chap.
5, “Fuel and Coolant Chemistry,” by W. R. Grimes, E.

G. Bohlmann, et al., pp. 95—-173.
2. E. Hoinkis, A Review of the Adsorption of lodine
on Metal and Its Behavior in Loops, ORNL-TM-2916

- (May 1970).

3. C. E. Milstead, W. E. Bell, and J. H. Norman,
“Deposition of lodine on Low Chromium-Alloy
Steels”, Nucl. Appl. Technol 7,361—66 (1969).

L )

I

o

.
Table 12.3. Indicated distribution of fission products in molten-salt reactors
istributi
Fission product group Example isotopes Distribution (%)
In salt To metal To graphite To off-gas Other

Stable salt seekers Z21-95, Ce-144, Nd-147 99 Negligible << 1 (fission recoils) Negligible Processing?
Stable salt seekers (noble gas precursors) Sr-89, Cs-137, Ba-140, Y-91 Variable/T, 2 of gas Negligible Low Variable/T | ;5 of gas

Noble gases Kr-89, Kr-91, Xe-135, Xe-137 Low/T;;, of gas  Negligible Low High/T , of gas

Noble metals Nb-95, M0o-99, Ru-106, Ag-111 1-20 5-30 5-30 Negligible Proces_singb
Tellurium, antimony Te-129, Te-127, Sb-125 1-20 20-90 5-30 Negligible Processing®
Iodine I-131,1-135 50-175 <1 <1 Negligible Processing®

2For example, zirconium tends to accumulate with protactinium holdup in reductive extraction processing.
bParticuiate observations suggest appreciable percentages will appear in processing streams.

¢Substantial iodine could be removed if side-stream stripping is used to remove I-135.

6¢1
4. P. R. Rowland, W. E. Browning, and M. Carlyle,
Behavior of Iodine Isotopes in a High Temperature Gas
Reactor Coolant Circuit, D. P. Report 736 (November
1970).

140

5. E. L. Compere, M. F. Osborne, and H. J. deNord-
wall, lodine Behavior in an HTGR, ORNL-TM-4744
(February 1975).

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