ORNL-TM-2304.txt

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OAK RIDGE NATIONAL LABORATORY
operated by

UNION CARBIDE CORPORATION
NUCLEAR DIVISION
for the
U.S. ATOMIC ENERGY COMMISSION

ORNL- TM- 2304

 

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MSRE DESIGN AND OPERATIONS REPORT
Part XI-A
Test Program for 233y Operation

J. R. Engel

NOTICE This document contains information of a preliminary nature
and was prepared primarily for internal use at the Oak Ridge National
Laboratory. It is subject to revision or correction and therefore does

not represent a final report.

GISTRIBUTLION Qf THIS DCCUMENT I8 INOTMITER,

 
LEGAL NOTICE -—-—n e

 

 

This report was prepared as an account of Government sponsored work, Neither the United States,
nor the Commission, nor any persoa acting on behaolf of the Commission:
A, Makes any warranty or representation, expressed or implied, with respect to the accurecy,

completeness, or usefulness of the informotion contained in this report, or that the use of

any information, opparctus, method, or process disclosed in this report may not infringe
privately owned rights; or
B. Assumes any liabilities with respect to the use of, or for damages resulting from the use of
any information, apparatus, method, or process disclosed in this report.
As used in the cbove, ‘“‘person acting on beholf af the Commission’” includes any employese or
contractor of the Commission, or employee of such centractor, to the extent that such employee
or contractor of the Commission, or employee of such contractor prepares, disseminates, or

!
I
i
provides access to, any informatien pursuant to his employment or contract with the Commission,
or his employment with such contractor.

 
ORNL-TM-230k4

Reactor Division

MSRE DESIGN AND OPERATIONS REPORT

Part XI-A

Test Program for 37U Operation

J. R. Engel

SEPTEMBER 1968

LEGAL NOTICE

This report was prepared as an account of Government sponsored work, Neither the United
States, nor the Commission, nor any persou acting on behalf of the Commission:

A, Makes any warranty or representation, expressed or implied, with respect to the accu-
racy, completeness, or usefulness of the information contained in this report, or that the use
of any information, apparatus, method, or process disclosed in this report may not infringe
privately owned rights; or

B. Assumes any liabilities with respect to the use of, or for damages resulting from the
use of any information, apparatus, method, or process disclosed in this report.

As used in the above, ‘‘person acting on behalf of the Cominission® includes any em-
ployee or contractor of the Commission, or employee of such contractor, to the extent that
such employee or contractor of the Commission, or employee of such contractor prepares,
disseminates, or provides access to, any information pursuant to his employment or contract
with the Commission, or his employment with such contractor,

OAK RIDGE NATIONAL LABORATORY
Oak Ridge, Tennessee

Operated by

UNION CARBITE CORPORATION
for the
U. S. ATOMIC ENERGY COMMISSION

g =

o
iii

CONTENTS
PREFACE .
INTRODUCTION. v « v v v v o o o o v v v o o o o &
OBTECTIVES. « ¢ v v v v v v o e v o e e e e e e e v e u
BASIC NUCLEAR TESTS v v v « o o o v v o o o o v e e u
233y Critical Experiment . . v v v v v v v v 4 . . .

Preparations for Fuel Loading . . . . .

2337 Loading SEqUEnce . . .« + « v . o . . . .
Control-Rod Calibration
Other Basic Nuclear Parameters

233] Concentration Coefficient of Reactivity.

Isothermal Temperature Coefficient of Reactivity.

Power Coefficient of Reactivity . . . . . . . .

REACTOR OPERATION WITH 33U FUEL. . .
Power Calibration. . . . « . « . « « + « v « o . . .
Control Systems Tests. . . ¢« v ¢ ¢« o ¢ ¢ ¢« & ¢« o o«

Reactor Dynamics . ¢« « . ¢ ¢« & ¢ o o &

Reactivity Balance . . . . & ¢ ¢ ¢ v o« o o o o o o

Application of Noise Analysis. . . « + « ¢« « o &

Measurement of 222U Capture to Fission Ratio . .
CHEMICAL, AND MATERTAL STUDIES IN THE FUEL ILOOP. . . .

Surveillance of Corrosion and Salt Contamination .

Surveillance of Uranium Inventory.

Fission-Product Behavior . . ¢ « ¢« ¢« & ¢ ¢« o o « + &

Graphite and Hastelloy Surveillance. . . . . . . . .

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O WO WO O O w w N«

H o R E E R R OE R F R
oUW DR OO O
PREFACE

This report is one of a series that describes the design and opera-

tion of the Molten Salt Reactor Experiment. All the reports have been

issued with the exceptions noted.

ORNL-TM-T728

+*
ORNL-TM-T729

ORNL-TM-T730

ORNL~TM-T31" "

ORNL-TM-732

ORNL-TM-2111

ORNL-TM-T733

ORNL-TM-90T

MSRE Design and Operations Report, Part I,
Description of Reactor Design by
R. C. Robertson

MSRE Design and Operations Report, Part IIT,
Muclear and Process Instrumentation, by
J. R. Tallackson

MSRE Design and Operations Report, Part I1I,
Nuclear Analysis, by P. N. Haubenreich,
J. R. Engel, B. E. Prince, and H. C. Claiborne

MSRE Design and Operations Report, Part IV,
Chemistry and Materials, by F. F. Blankenship
and A. Taboada

MSRE Design and Operations Report, Part V,
Reactor Safety Analysis Report, by S. E. Beall,
P. N. Haubenreich, R. B. Lindauer, and
J. R. Tallackson

MSRE Design and Operations Report, Part V-4,
Safety Analysis of Operation with 27U, by
P. N. Haubenreich, J. R. Engel, C. H. Gabbard,
R. H. Guymon, and B, E. Prince

MSRE Design and Operations Report, Part VI,
Operating Limits, by S. E. Beall and
R. H. Guymon

MSRE Design and Operations Report, Part VII,
Fuel Handling and Processing Plant, by
R. B. Lindauer

 

*
Part of this report, IT-A, has been issued; the remainder is in

process.

¥%
These reports will not be issued.
ORNL-TM-908

ORNL-TM-909

ORNL-TM-910

ORNL-TM-911

ORNL-TM-2304

vi

MSRE Design and Operations Report, Part VIII,
Operating Procedures, by R. H. Guymon

MSRE Design and Operations Report, Part IX,
Safety Procedures and Emergency Plans, by
A, N. Smith

MSRE Design and Operations Report, Part X,
Maintenance Equipment and Procedures, by
E. C. Hise and R. Blumberg

MSRE Design and Operations Report, Part XI,
Test Program, by R. H. Guymon,
P. N. Haubenreich, and J. R. Engel

MSRE Design and Operations Report, Part XI-A,

Test Progrsm for =°°U Operation, by
J. R. Engel

MSRE Design and Operations Report, Part XIT,

Lists: Drawings, Specifications, Line Schedules,

Instrument Tabulations (Vol. 1 and 2)

 

These reports will not be issued.
INTRODUCTION

The initial critical operation of the MSRE with #25U occurred on
June 1, 1965. The reactor was subsequently operated at various powers
up to 8 Mw for sustained periods and accumulated a total of 9005 equiva-
lent full-power hours with that fuel. The reactor loop was drained on
March 29, 1968 and preparations were started to replace the 23°U-27gg
mixture in the fuel salt with #>2U. The test program that was conducted
with the ©35U ig described in Reference 1. The purpose of this memo is

to outline the program that is to be followed with the 2337 loading.
OBJECTIVES

Since the MSRE will be the first reactor to be fuelled completely
with 223U, there will be considerable interest in the initial critical
experiment. This experiment will provide additional data on the ade-
quacy of the calculational techniques used to predict the critical
uranium concentration in the MSRE. Comparison of the results with those
of the 23U critical experiment will also provide some indirect evidence
about the quality of the input nuclear data used for 33U, In addition
to the initial critical concentration, we will measure other basic nuclear
parameters of the system with 2337 fuel ~— temperature and uranium-
concentration coefficients of reactivity, reactivity effects of fuel
circulation, and control-rod reactivity worth. In each case comparisons
will be made with the predicted values.

After the zero-power experiments, we will continue our studies of
the overall nuclear performance of the MSRE. Some changes, due to the
233U, are expected in the long-term reactivity behavior and in the dy-
namic response of the reactor. Extensive investigations will be carried

out in both these areas to compare the predicted and observed behavior.

 

1R. H. Guymon, P. N, Haubenreich, and J. R. Engel, MSRE Design and
Operations Report, Part XI, Test Program, USAEC Report ORNL-TM-911,
Oak Ridge National ILaboratory, November 1966,
In addition, we plan to use neutron fluctuation spectra as an operational
diagnostic aid. Some useful correlations were developed during the 235y
operation and it may be practical to use similar correlations to monitor
reactor performance. Of particular interest in the area of performance
tests will be a special experiment to measure the effective neutron yield
for 33U in a molten salt reactor neutron spectrum.

Studies of reactor chemistry and materials behavior will be con-
tinued throughout the operation of the reactor system. Data will be
gathered on our ability to accurately monitor uranium inventory at low
concentrations as well as on the behavior of fission and corrosion pro-
ducts. The studies of the effects of exposing graphite and metal to fuel
salt, fission products, and radiation will also be continued.

The =3?U fuel mixture will probably be used for all the remaining
operation of the MSRE. However, consideration is currently being given
to an interruption of that operation to permit substitution of a less
expensive secondary salt (sodium fluoroborate) for the LiF-BeF- mixture
in the system to demonstrate its operating characteristics. Since this
change is still being studied, the test program for that phase of opera-

tion will be defined later.
BASTC NUCLEAR TESTS

In addition to providing a check on the calculational techniques
used to predict the properties of the MSRE with 233y fuel, measurements
of these basic properties will supply much of the data that is required
for monitoring the subsequent behavior of the reactor. The on-line
reactivity-balance calculation requires, as input information, data on
control-rod worth, and various coefficients of reactivity. Where possible,
results of direct measurements of these properties will be used. In
other cases, (e.g. fission-product effects) calculated values will be
employed., Direct comparisons of calculated and observed values will be

useful in establishing confidence in quantities that cannot be measured.
2337 Critical Experiment

Essentially all of the original uranium will be removed from that
portion of the fuel salt that is fluorinated. However, a small heel of
fuel salt (containing about 1.2 kg of the total U) will be left in a
fuel drain tank when the salt is transferred to the fuel storage tank for
processing. This uranium will be mixed with the fuel carrier salt before
the #23y critical experiment is started. Most of the plutonium and many
of the non-volatile fission products (notably samarium) that were produced
in the 235y operation will remain in the salt for the 22U operation.
Thus, the £33y critical experiment will not be ''clean'" and corrections
for the effects of these contaminants will have to be made when the re-
sults are evaluated.

Except for a practice addition of about 0.8 kg of 238U,* all of the
uranium that is added in the critical experiment will have the isotopic
composition listed in Table 1. This uranium is available as the eutectic
salt mixture LiF-UFy (73 - 27 mole %) in special cans containing up to
7 kg of total U. As in the initial %75U critical experiment, most of the
uranium will be added to the fuel salt in a drain tank (FD-2). At ap-
propriate intervals the reactor will be filled with the salt mixture
to follow the subcritical multiplication as the uranium concentration is
increased. After the fourth fill, the concentration will be close to the
critical value and subsequent uranium additions will be made with capsules

through the sampler-enricher to make the reactor critical.
Preparations for Fuel Loading

Since the 237U mixture is heavily contaminated with ©>2U and the
last chemical purification of the uranium occurred some L4 years ago, the
enriching salt contains substantial amounts of the daughter products of

238y decay. Several of these daughters are strong emitters of both alphas

 

*
The 28U is added to adjust the isotopic composition of the mixture

for convenience in evaluating "alpha" for 233y later in the operation of
the reactor.
Table 1

Isotopic Composition of £>°Feed Material

Abundance
U_Isotope (atom %)
232 0.022
233 91.49
234 7.6
235 0.7
236 0.05
238 0.1k

and gammas, making the cans of salt strong neutron (fiom a,n reactions

in fluorine and lithium) and gamma radiation sources. 1In addition, the
carrier salt, to which the uranium must be added, is highly radiocactive,
although it will contain essentially no volatile fission products. These
considerations require that shielded equipment be used for the uranium
additions to the drain tanks. Special charging equipment, using shielded,
remote-maintenance components, has been built and installed above fuel-
drain-tank No, 2 (FD-2), as shown in Fig. 1. This equipment permits the
transfer of single cans of enriching salt (containing no more than 7 kg
of U) from the shielded transport cask into the drain tank under shielded,
controlled-ventilation conditions. The empty cans are stored on a
turntable within the equipment for removal as & group at the end of the
drain-tank loading operations.

The nuclear reactivity of a drain tank containing 2330 fuel will be
somewhat higher than the same drain tank containing the 2357238y mixture.
Although the drain tank is expected to be far suberitical under all normal
storage conditions, careful observations will be made during the fuel ad-
ditions to ensure that criticality is not approached in the tank. To
accomplish this, two neutron-sensitive chambers — a sensitive BFs cham-
ber and an insensitive fission chamber to cover a wide range of counting

rates — will be installed just outside the drain tank for the loading
ORNL-DWG 68-967

   
     
          

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Fig. 1. Arrangement for Adding 233y Enriching Salt to Fuel Drain
Tank.
operations. Since the fuel itself is an intense ((-n) neutron source,
no external source will be required for neutron monitoring.

Neutron counting to observe the progress of subcritical multipli-
cation with the fuel in the reactor will be accomplished with the normal
reactor instrumentation in the nuclear-instrument penetration. For the
suberitical conditions we will use the high-sensitivity BF=s chamber and
the two movable fission chambers. Since the reactor cell will be covered
during the entire experiment, it will not be possible to install extra
chambers around the core. For the same reason the external neutron source
will remain fixed in the thermal shield throughout the experiment.
However, the intense internal neutron source will completely overshadow
the external source so that a movable source is of little value in this

experiment.

233 Loading Sequence

 

The bulk of the =33y enriching salt will be added to the fuel carrier
salt through the special equipment attached to FD-2. We anticipate adding
about 3k kg of total U in L4 major steps. The first two rounds of ad-
ditions will consist of 21 and 7 kg U, respectively, and the subsequent
additions will be based on extrapolations of count-rate ratios obtained
from the preceding additions with the salt in the reactor. The objective
is to bring the uranium loading to within 1/2 kg of critical in this
manner. The enriching salt is available in cans of various sizes so that
arbitrary amounts can be added in 1/2-kg increments. The final uranium
additions will be made to the circulating loop in 98-gram increments
through the sampler-enricher.

The initial charging operation will require the addition of three
T-kg cans of uranium to FD-2. These cans will be delivered to the re-
actor site individually, in a shielded cask and inserted into the charging
equipment., The cans are designed to be remotely suspended in the gas
space of FD-2 above the liquid carrier salt. 1In this position the en-
riching salt will slowly melt and drip into the carrier salt below it.
During this time the can will be suspended from a weighing device so that

progress of the melting can be followed. (Gross and tare weights of each
can of salt will be available in advance.) At the same time we will
observe the increase in neutron count rate as the neutron source and sub-
critical multiplication increase. If the count rate increases too rapidly,
the can can be withdrawn to slow or stop the uranium addition. After the
addition, the empty can will be weighed more accurately, to ensure that

it is empty, and stored on the turntable for later disposal.

Prior to the addition of each can of enriching salt, one-half of
the carrier salt will be transferred to the adjacent fuel drain tank (FD-1)
to provide room for suspending the cans without contacting the salt. After
the addition of each can, the remaining salt will be returned to FD-2 for
mixing and to provide neutron count-rate data on the full tank. Extrapo-
lations of ratios of these count rates will be used in conjunction with
observations during additions to ensure that the drain tank remains sub-
critical. After the addition of the first can of U, the transfers to
FD-1 will also remove some uranium to keep X pp Very low during subse-
gquent additions.

After three cans of enriching salt (2l-kg U) have been added in this
manner, the mixed fuel salt will be loaded into the reactor. Neutron
count rates will be measured with the salt at several levels in the re-
actor vessel to ensure that criticality is not attained before the vessel
is full. (During salt additions, as in all filling operations, the three
control rods will be partly withdrawn so that they can suppress any pre-
mature criticality and initiate a salt drain,) Additional count-rate
data will be obtained before, during, and after circulation of the salt
in the loop. These data will be used for extrapolations to the pro-
Jjected critical loading.

Since the 22U mixture provides such an intense internal neutron
source, count rates with the external source and no fuel are of little
value in this critical experiment. Therefore, the data obtained from
the first loop fill with uranium-bearing salt will be used as a basis
for the usual inverse-count-rate plots. Thus two loadings of predeter-
mined size (21-kg and T-kg) are required before extrapolations can be

made to establish the size of subsequent loadings.
The procedures described above will be carried out three more times
(with different numbers and sizes of E33y-salt cans) to bring the reactor
system within l/2—kg of the critical loading. After the fourth loop
fill, the salt will remain in the loop and 98-gram additions of uranium
will be made through the sampler-enricher to make the reactor just criti-
cal at 1200°F with the fuel stationary and all control rods fully with-
drawvn. The fuel salt will then be left in the loop for the remainder of

the zero-power and low-power tests.
Control-Rod Calibration

Although there will be no change in the basic configuration of the
reactor, the control rods in the MSRE will have about 30% more reactivity
worth with Z3%U fuel than with the original loading. Therefore, a com-
plete recalibration of the rods will be required before the reactor can
be returned to full operation. The basic approach to be used for this
work will be the same as that used for the original calibration.®,”

The fundamental measurements to be made are the differential worth
of one rod as a function of position with the other two rods withdrawn
to their upper limits. These measurements will use the rod-bump period
technique with the fuel salt stationary. The various critical positions
of the control rod in question will be obtained by adding uranium to the
salt to increase the amount of rod insertion., Since uranium can only
be added in increments of 98 grams, each capsule will increase the fuel
reactivity by about 0.12% 5k/k, so about 24 capsules will be needed to
produce full insertion of one rod. As a conseguence, rod sensitivity

measurements will be made after at least every second capsule addition.

 

ZB. E. Prince et al., Zero-Power Physics Experiments on the Molten-
Salt Reactor Experiment, USAEC Report ORNL-4233, Oak Ridge National
Laboratory, February 1968, pp 11 - 37.

°B. E. Prince, Period Measurements on the Molten Salt Reactor Experi-
ment during Fuel Circulation: Theory and Experiment, USAEC Report
ORNIL-TM-1626, Oak Ridge National Laboratory, October 1966
Measurements may be made after each addition in regions where the rod
worth is changing rapidly. Integration of the differential-worth data
will then provide a curve of reactivity as a function of position for one
control rod.

The bagic data will be supplemented by differential-worth measure-
ments with the fuel circulating, rod-shadowing measurements, and rod-drop
experiments to provide information on the reactivity worth of all three
rods as a function of configuration. All of the data will be used to
evaluate coefficients in a theoretically derived expression of rod worth
that is amenable to evaluation by a digital computer. This expression
will then be used (as it was during the £35y operation) by the on-line
computer to calculate control-rod poisoning from the positions of the

three rods.

Other Basic Nuclear Parameters

233 Concentration Coefficient of Reactivity

As excess 237U is added to the loop to bring the concentration to
the operating value and calibrate the control rods, data will be col-
lected to evaluate the reactivity effect of the excess uranium. These
data will be reduced to a uranium-concentration coefficiént of reactivity
to be used in evaluating the effects of burnup and subsequent fuel

additions.
Tsothermal. Temperature Coefficient of Reactivity

When the initial set of fuel additions has been completed, an experi-
ment will be performed to measure the isothermal temperature coefficient
of reactivity of the reactor. In this experiment the fuel loop tempera-
ture will be slowly varied between about 1150°F and 1225°F while the
control-rod configuration required to keep the reactor just critical 1is
recorded. The observed reactivity change will be corrected for any ef-
fects due to changes in the circulating void fraction with temperature
to obtain the total (fuel + graphite) temperature coefficient of reac-

tivity.
10

Power Coefficient of Reactivity

In the MSRE, a power coefficient of reactivity is used to describe
the reactivity effect of the change in steady-state temperature distri-
bution in the core that accompanies a change in power level, The value
of this coefficient depends on the mode of temperature control (the re-
actor outlet temperature is held constant on this reactor) and the magni-
tudes of the separate fuel and graphite temperature coefficients of re-
activity. Since the detailed temperature distribution in the core cannot
be measured directly, the power coefficient will be inferred from the
observed change in control-rod configuration with power after steady-
state temperatures are achieved and before there has been a significant
change in fission-product poisons. Observations will be made at several

levels during the approach to full power to obtain a best value.
REACTOR OPERATION WITH 33U FUEL

After the initial tests to investigate the physics of the MSRE with
233U, the reactor wlll be operated at power to continue the studies of
its long-term behavior. Some special tests will be required to prepare
for extended operation and others will be used to demonstrate the con-

tinuing satisfactory performance.
Power Calibration

The instantaneous indication of reactor power for the servo control
and safety instruments is derived from neutron-sensitive chambers in the
nuclear instrument penetration. Since the ratio of power level to neu-
tron flux in the penetration may change with the new fuel, all the chanm-
bers will be repositioned as required to eliminate any inconsistencies.
The reference standard for repositioning the chambers will be the heat-
power of the reactor calculated by the on-line computer from overall
system heat balances. FEach of the compensated and uncompensated ion
chambers will be individually positioned to give a direct readout of re-
actor thermal power. In the case of the wide-range counting channels it

may be necessary to modify the function generators that operate on
1l

chamber position to obtain consistency over the entire power range.
After all the chambers have been shifted, they will be rechecked to elimi-
nate any mutual shadowing effects.

Special precautions will be observed in moving the uncompensated
chambers that serve as inputs to the flux safety system. Only one chamber
at a time will be moved under strict administrative control and the other
two chambers will be watched carefully to ensure that they are not ad-

versely affected by the one that is moved.
Control Systems Tests

The reactor control systems (flux, temperature, and load) were tested
under a variety of conditions with 235U fuel to demonstrate their ade-
quacy.? A similar series of tests will be performed with the 37U fuel.
However, this time the main emphasis will be on the reactor servo con-
troller, flux servo at low power and temperature servo at powers above
1 Mw. (The load-control system will not be affected by the change in
fuel.) Both the steady-state behavior and the response to perturbations
will be examined. Calculations, analog-simulator studies, and measure-
ments on the MSRE indicated that, at most, the high~frequency gain of the
flux servo may have to be adjusted slightly for satisfactory performance
with 23%U. (Reference 5) If such changes are made, the pertinent tests

will be repeated to demonstrate the adequacy of the final system.

 

4R. H. Guymon, P, N. Haubenreich, J. R. Engel, MSRE Design and
Operations Report, Part XI, Test Program, USAEC Report ORNL-TM-911,
Oak Ridge National Laboratory, November 1966, pp. 5-2 to 5-3.

S0ak Ridge National Laboratory, MSRP Semiann. Progr. Rept.
February 1968, USAEC Report ORNL-425k4, in preparation.
12

Reactor Dynamics

The dynamic behavior of the MSRE with #°°U fuel was the subject of
extensive theoretical and experimental investigation.®»7 A comparable
theoretical analysis of the dynamics with 23] has been performed and
the results indicate that the reactor will be inherently stable at all
powers.® The purpose of the dynamics tests is to provide another veri-
fication of the calculational techniques.

As with the ®7°U operation, various dynamic tests will be performed
at zero power and during the approach to full power, Follow-up tests at
power will be performed periodically to prove the persistence of proper
behavior. In general, the tests will include pulse and step reactivity
perturbations with a control rod as well as pseudorandom binary and

ternary perturbations of control-rod position and fiux demand.
Reactivity Ralance

Reactor operation with 235 demonstrated the utility of an on-line
reactivity balance as an operating guide.® The same type of calculation,
with appropriate changes in coefficients, will be used during the £33
operation.

Detailed experiments during the last operation with 235y showed
that the current calculation does not adequately treat the xenon poison-
ing. Veariations in system temperature and pressure induce changes in

xenon poisoning by affecting the effectiveness of the gas stripper.

 

©5. J. Ball and T. W. Kerlin, Stability Analysis of the Molten-Salt
Reactor Experiment, USAEC Report ORNL-TM-1070, Oak Ridge National Labora-
tory, December 1965.

7T. W. Kerlin and S. J. Ball, Experimental Dynemic Analysis of the
Molten-Salt Reactor Experiment, USAEC Report ORNL-TM-1647, Oak Ridge
National Iaboratory, October 13, 1966.

S0ak Ridge National Laboratory, MSRP Semiann. Progr. Rept. Aug. 31,
1967, USAEC Report ORNL-5191, pp 61 - 62,

°J. R. Fngel and B. E. Prince, The Reactivity Balance in the MSRE,
USAEC Report ORNL-TM-1796, Oak Ridge National Laboratory, March 10, 1967.
13

These changes are of secondary interest in the MSRE because most of the
reactor operation is at a fixed temperature and pressure. However, since
a detailed understanding of the xenon behavior is important to the
breeder programs, an attempt will be made during the 23U operation to
modifly the mathematical model to incorporate temperature and pressure
effects. It may be necessary to conduct additional experiments to supple-

ment the available data in order to accomplish this improvement.
Application of Noise Analysis

In the course of operating the MSRE with £75U, a large amount of
data was collected on the spectral density of the inherent neutron-level
fluctuations in the reactor. Evaluation of these data indicates that
changes in the flux 'moise™ spectrum are a good qualitative indication
of changes in the circulating void fraction in the fluid fuel. Neutron
noise data will be collected routinely to monitor this aspect of reactor
operation and to look for any other changes in system performance. At-
tempts will also be made to make the void indication more quantitative.

Techniques have been developed to use the BR-340 computer at the
reactor site to collect neutron-noise data (with all other computer
functions inhibited) and process it immediately thereafter (with all
other functions active). Thus, spectral-density results can be made
availlable with very little delay for use as an operating guide if
satisfactory correlations are developed. The same techniques may be
used to monitor the vibration spectra from mechanical components if

adequate data samples can be obtained.
Measurement of 227y Capture to Fission Ratio

An important factor in the breeding performance of molten salt re-
actors is the ratio of parasitic neutron captures to fissions ("alpha')
in ®37U in the reactor neutron spectrum. In current reactor design cal-
culations, this ratio is obtained, in effect, by integrating differ-
ential cross sections over the neutron-energy spectrum. A precise
measurement of the integral value of "alpha' in a neutron spectrum typi-

cal of moliten-salt reactors could reduce the uncertainty in the breeding
1h

performance of new core designs., Since the neutron spectrum in the MSRE A
with 223U fuel is very similar to that in the proposed breeders, a pre-

cise measurement of "alpha' will be made during the 233y operation. The ?
measurement consists of precise uranium isotopic assays before and after

substantial 223U burnup. The value of "alpha" is derived from the build-

up of 234U relative to the depletion of =>7U.

CHEMICAL AND MATERIAL STUDIES IN THE FUEL LOOP

A major objective in the operation of the MSRE is to study the be-
havior of the basic materials — molten salt, graphite, and Hastelloy-N —
in combination in a radiation enviromment. These studles are accomplished
primarily through the examination of samples of the appropriate materials.
A high level of effort will be maintained in this area throughout the

operation of the reactor.
Surveillance of Corrosion and Salt Contamination

The most direct monitor of Hastelloy-N corrosion that is readily
available in the MSRE is the level of chromium in the fuel salt, Chro-
mium, leached from the surface of the metal remains in solution in the
salt where its concentration can be measured in samples. In the first
three years of reactor operation, the chromium in the fuel salt increased
from ~ 38 ppm to ~ 80 ppm. If this is interpreted in terms of uniform
attack on the fuel loop the increase represents leaching from less than j
0.3 mil of metal. This low rate of attack is expected to continue but |
salt samples will be analyzed regularly to detect any changes.

The processing of the fuel salt to remove the #°°U will require the
establishment of a new baseline of chromium concentration. Fluorination
of the salt in the fuel storage tank will substantially increase the
Cr (also Fe and Ni) level. However, most of this will be reduced to free
metal and filtered out in subsequent steps before the processed fuel is
returned to the drain tanks. The resultant chromium concentration will

be measured in samples taken at the start of the Z°7U operation.
15

Another salt contaminant that is carefully monitored is oxygen. The
oxide tolerance of the MSRE fuel mixture is ~ 700 parts per million but
observed concentrations have been around 50 - 60 ppm. Samples will be
analyzed regularly to ensure that oxygen intrusion remains at a low

level.
Surveillance of Uranium Inventory

The chemical concentration of uranium in the fuel salt for the 23°U
operation was about L4.6% by weight. At this level the standard deviation
of all the uranium chemical analyses was 0.02 wt.%.* The average ana-
lytical concentration at the end of that operation was such that the
apparent uranium inventory was within 200 gm (out of 222 kg) of the book
inventory derived from known additions and depletions.

In the 237Q operation, the chemical uranium concentration will be
only about 0.7 Wt%. It appears that average values of analytical results
will provide the necessary data for precise, long-term comparisons of
"book" and observed uranium inventory. However, improvements in pre-
cision are under development and will be required to permit short-term
comparisons using individual results. In general, reactivity-balance
results will be used in conjunction with the uranium analytical results

to monitor the short-term behavior of the system.
Fission-Product Behavior

Several aspects of fission-product behavior are of particular
interest in the MSRE. These include the deposition of noble-metal species
on graphite and Hastelloy surfaces, the escape of some noble metals as
"smoke'" or "dust" in the reactor offgas, and the escape of other volatile
species (Xe, Kr, Te, etc.).

Much information has been collected about the behavior of fission
products in this system but more is required for a thorough understanding

and evaluation, Therefore, considerable effort will be expended during

 

*
The relative precision, AC/c is %% or 0.4,
16

the remaining operation of the reactor in analyzing the fuel salt and
reactor offgas to elucidate the fisslon product behavior. To aid in
this effort special samples of salt and cover gas will be withdrawn
through the fuel sampler. In addition, use will be made of the offgas
sampler and of special instrumentation and collection devices installed
in the reactor offgas line at the fuel pump. Selected specimens of
metal and graphite that are exposed in the reactor core will be examined
to provide more data on fission-product plateout on surfaces and in-
trusion into graphite.

The circumstances under which various samples must be obtained will,
to some extent, influence the power operation of the reactor. For ex-
ample, it may be desirable to take some samples while the fuel salt is
stationary, in which case the reactor must be at zero power., In other
cases, prolonged operation at power will be required to reach steady-
state conditions with regard to particular fission products. A complete
shutdown will be required to remove specimens from the core or the off-

gas line,
Graphite and Hastelloy Surveillance

Graphite and Hastelloy surveillance specimens are exposed to fuel
salt in the MSRE core and other Hastelloy specimens are suspended just
outside the reactor vessel., The primary purpose of these specimens is
to study radiation and salt-exposure effects under reactor conditions;
however, they have also been used in connection with some of the fission-
product studies. Selected specimens are removed for examination at
approximately six-month intervals and replaced with new ones. Since such
changes are major operations, they are normally scheduled to coincide
with other major reactor shutdowns.

The first sets of samples installed in the MSRE were representative
of materials actually used in the construction of the reactor system.
Since that time continued development effort has led to other graphites of
interest and to minor changes in the composition of Hastelloy-N, Conse-
quently, subsequent arrays have included samples of some of these materials,
This surveillance program is expected to continue for the operating life

of the reactor.
O -3 OV W P
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17

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Bohlmann
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Corbin
Cottrell
Crowley
Culler, Jr.
Davis
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Doss
Eatherly
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Hudson
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Kirslis
Krakoviak
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Lindauer
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MacPherson
Martin
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18

Internal Distribution

 

(continued)
85. R. C. Robertson 102. R. C. Steffy
86, J (. Robinson 103, H. H. Stone
87-91., M. W. Rosenthal 104, J. R. Tallackson

92. A. W. Savolainen 105. R. E., Thoma
93, Dunlap Scott 106, D. G. Trauger
94, J. H. Shaffer 107. C. 8. Walker
95, E. G. Silver 108. B. H. Webster
96, M. J. Skinner 109. A. M. Weinberg
97. A. N. Smith 110. J. R. Welr

98. 0. L. Smith 111, K. W. West

99. P. G. Smith 112. M. E, Whatley
100. I. Spiewak 113. J. C. White
101. D. A, Sundberg 114. G. D. Whitman

115. Gale Young

116-117. Central Research Library (CRL)

118-119. Y-12 Document Reference Section (DRS)

120-149. ILaboratory Records Department (LRD)
150. ILaboratory Records, (LRD-RC)

Fxternal Distribution

 

151-152. T. W. McIntosh, AEC, RDT, Washington, D. C. 20545
153, H. M. Roth, Laboratory and University Division, ORO
15h, T. G. Schleiter, AEC, RDT, Washington, D. C. 20545
155, Milton Shaw, AEC, RDT, Washington, D. C. 20545

156-170. Division of Technical Information Extension, (DTIE)