ORNL-3215.txt

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ORNL-3215
UC-80 — Reactor Technology

MOLTEN-SALT REACTOR PROGRAM
PROGRESS REPORT
FOR PERIOD FROM MARCH 1 TO AUGUST 31, 1961

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

L oxrel

T=

$2.75

Office of Technical Services

Printed in USA. Price . Available from the

Department of Commerce
Washington 25, D.C.

LEGAL NOTICE

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

A. Mokes any warranty or represéntation, expressed or implied, with respect to the accuracy,
s of the information contained in this report, or that the use of

completeness, or usefuln

ony information, apparatus, method, or process disclosed in this report moy not infringe

privately owned rights; or
e with respect 1o the use of, or for domages resulting from the use of

B. Assumes any liabi
any information, opporatus, method, or process disclosed in this report.
As used in the above, “‘person acting on behalf of the Commission includes any employee or

contractor of the Commizsion, or employee of such contractor, 1o the extent that such employee
or contractor of the Commission, or employee of such contractor propares, disseminates, or
provides occess to, any information pursuant 1o his employment or controct with the Commission,
or his employment with such contractor.

ORNL-3215
UC-80 — Reactor Technology
TID-4500 (16th ed.)

Contract No. W-TLO5-eng-26

MOLTEN-SALT REACTOR PROGRAM

PROGRESS REPORT

FOR PERIOD FROM MARCH 1 TO AUGUST 31, 1961

R. B. Briggs, Program Director

Date Issued

(At J N T A ‘@2
B | g
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OAK RIDGE NATIONAL LABORATORY
Oak Ridge, Tennessee
operated by
UNION CARBIDE CORPORATION
for the
U.S. ATOMIC ENERGY COMMISSION

MARTIN MARIETTA ENERGY §

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3 Y456 O3LYyypg 1

RS

SUMMARY

PART I. MSRE DESIGN, COMPONENT DEVELOPMENT, AND ENGINEERING ANALYSIS

1. MSRE Design

In order to facilitate graphite sampling and to make possible the use of
solid control rods, the reactor layout was reviewed, and the pump location was
moved from the top of the reactor to a position which left the top of the reactor
accessible from above. This change, while not of a basic nature, necessitated a
considerable amount of new design work.

A "tee" section on top of the reactor vessel was designed as a port of entry
for control rods and graphite samples and as an outlet for fuel. A frozen salt
seal and a gasketed flange type of gas seal were designed for the entry port.
Also thimbles and rods together with rod drives and a cooling system had to be
designed for the control system.

Moving the pump required designing a pump support structure which would
allow movement of the pump as the suction and discharge lines changed length with
thermal cycling of the system., OStress analysis of the entire piping system was
redone,

Auxiliary systems for the pump, off-gas handling system, electrical heating
system layout, and all penetration designs were modified somewhat as a result of
the cell piping changes.

Containment cell penetrations were designed and checked. The design of the
top plugs was modified to increase the minimum design pressure for the contain-
ment to 40 psig.

The drain-tank cooling system was simplified and improved with respect to
mechanical complexity and steam flows.

Major building modification design was completed and sent to prospective
bidders.

Component designs were completed, and a design report for these components
was 1ssued. A design package consisting of the reactor vessel, storage tanks,
radiator, and heat exchanger was sent to prospective bidders,

Thirteen instrument-application drawings were issued for comment. Prepara-
tion of these drawings is approximately 85% complete. Preparation of instrument-
application tabulations is approximately 60% complete. Twelve thermocouple-
locations drawings were issued for approval, Control circuit design is in a pre-
liminary stage., Layout of the instrumentation and controls system is proceeding
as layout of the building and equipment become firm and as instrumentation
requirements become known., A front-elevation layout of the main control board
and a proposal for layout of the main control area have been prepared.

iii
A study of MSRE data-handling requirements 1is nearing completion.

A study is under way to determine the most reliable and least expensive
method of providing single-~point temperature alarm channels for monitoring the
operating status of freeze flanges and valves.

There are no areas of design which pose unresolved problems, and the design
of the MSRE is essentially on schedule.

2. Component Development

Design of a 5-in. freeze flange incorporating a buffered ring-joint gas seal
was completed and procurement initiated for INOR-8 forgings and soft nickel rings.
A 6-in, freeze flange of Inconel was fabricated and installed in the thermal-
cycle and gas-seal test facility, using a L0O-kc induction coil in the bore to
produce the temperature distribution. A study of fabrication and inspection
methods for the flanges and rings was initiated. Thermal cycle and seal tests
for a 3-1/2-in. Inconel flange and gold-plated Inconel gasket using a spring
clamp indicated that successful seals can be made by this method.

Freeze valves opened by Calrod heaters and incorporating siphon breaks were
installed in the engineering test loop and successfully tested under extreme
operating conditions.

A two-plece pipe heater insulation unit was tested and discarded because of
excessive heat loss. A modified heater, better adapted to maintenance from above
and designed for less heat loss, was accepted for testing.

Tests conducted on the core-heater prototype verified the calculated heat
loss; however, insulation shrinkage caused some trouble. The system was operated
for 3000 hr without difficulty and with no apparent deterioration of the metal
components.

A firm proposal for the sampler enricher was accepted for testing and detall
design was started.

The solder freeze valve was less reliable than had been predicted and was
replaced in the sampler design with a buffered-seat gate valve. The sample cap-
sule development progressed to the stage of a hot-cell crushing and unloading
study. A sample transport container was designed and constructed for testing a
system of atmospheric control during shipment of the deoxidized capsule to the
site and return of the sample to the hot cell,

The drain-tank cooler was mocked up and testing was started.

Flow tests in the lower head of the one~fifth-scale core model indicate a
shift in the flow characteristics with reduced flow. Fuel age and heat transfer
coefficient and characteristics of solid distribution were measured in this area.
Efforts to produce oscillatory disturbances in the core were unsuccessful,

Cormponents for the full-scale-core-model test loop were fabricated and in-
stalled in preparation for the arrival of the core internmals.

Helium purification tests, using titanium sponge as the oxygen getter, were
used as the basis for the design of a full-scale MSRE helium purification system.

The engineering test loop was operated continuously for 1300 hr; the oxide
bulldup history was observed during a startup-flushing operation. Revisions made
to the bubble-type level indicator made the system very reliable under engineering-
test loop conditions. The design of the graphite sectlon of the salt loop to-
gether with the graphite-handling dry box was completed and fabrication started.

A maintenance plan using the most appropriate maintenance method for each
operation was evolved. The one-twelfth-scale model was completed to the existing
design stage. A full-scale maintenance mockup is being constructed for study of
special problems such as freeze-flange operation.

The program to develop a remotely brazed Jjoint for l~l/2-in. pipe progressed
to the bench-demonstration stage, and design for tools to produce remotely brazed
Joints was started. OSeveral mechanical disconnects were demonstrated which could
be used in auxiliary lines,

A program was initiated for the study of the steam generator problems.

Development of a continuous-level element, for use in measurement of molten-
salt level in the MSRE fuel and coolant salt pump bowl, is continuing. A
graphite float assembly was fabricated, and a test stand was constructed. A
recent transformer design for the level element was operated for more than two
weeks at temperatures in excess of 1200°F without evidence of insulation failure.
Performance data obtained from this transformer at temperature is very encouraging.

A new design concept for a single-point conductivity-type level probe was
developed and tested. A conceptual design was developed for a two-point level
probe, based on the new design concept and compatible with the physical geometry
of the MSRE storage tanks.

Extensive measurements were made of the resistivity of a molten salt at
various voltages, frequencies, and temperatures.

A temperature-scanning system is being developed to a present profile dis-
play of approximately 250 thermocouples attached to the reactor pipes and
components.

The development of techniques and procedures for attaching sheathed thermo-
couples to INOR-8 pipes and vessels is continuing. Sample quantities of sheathed
thermocouple wire have been procured, and several methods of attachment are being
investigated.

3. Reactor Engineering Analysis

The MSRE temperature coefficients of reactivity were calculated and found to
be -2.8 x 1077 Ak./°F for the fuel salt, and -6.0 x 10 ° Ake/°F for the graphite.
The reactivity worth of all three MSRE control rods was calculated to be 6.T7%
Mke. The worth of individual rods varied between 2.8 and 2.9% Ake, while the
reactivity worth of pairs of rods varied between 4.9 and 5.3% JAY: 9N

The addition of rod thimbles (no control rods) lowered the peak-to-average
power density ratio by about 8% of the homogeneous reactor value and moved the
position of peak power density to a point sbout 6 in. away from the reactor
center line,

The peak gamma-ray heating rate in the core was calculated to be 2.5 w/ce
with the reactor at 10 Mw; the maximum fast-neutron heating in the rod thimbles
(rods out) was sbout 0.1 w/cc.
Vi

The critical fuel concentration with the use of 4LO-mil-thick INOR-8 fuel
tubes in the MSRE was calculated to be nearly double that associated with no
cladding present.

An improved estimate of the activity acssociated with the Flg(n,a)N16 reaction
was obtained. The N'© activity in the fuel salt ranged from 0.27 X 101° ais sec™t
cc™l of fuel at the pressure-vessel inlet to 0.65 x 10'° at the exit from the
core proper. The activity associated with the pump bowl was calculated on the
basis that the daughter products of xenon and krypton plate out in the bowl. The
associated dose rate 10 ft from the pump bowl was 10° r/hr after 10 days' cooling
time following 1 year's operation at 10 Mw. For the same conditions, the resid-
ual activity in the heat exchanger gave a dose rate of 2 x 10% r/hr, based on the
assumption that all the isotopes which might plate out on INOR-8 do so in the
heat exchanger.

The gamma-ray dose rates above the top shield (3.5 ft of barytes concrete :
plus 3.5 ft of ordinary concrete) during 10-Mw operation were calculated to be
about 15 mr/hr for a solid shield and sbout 80 mr/hr if the ordinary concrete
has 1/2-in.-thick slits; the neutron dose rates for these conditions were about .
2 mr/hr and 4.6 r/hr, respectively. Filler material and additional shielding
will reduce the dose rates below the tolerance value.

Estimates were made of the dose rates outside the side shield from individual
sources within the reactor cell. The primary radiation source during 10-Mw power
operation was the neutron-capture gammas from the iron in the thermal shield.

With 7 £t of ordinary concrete the total dose rate was about 45 mr/hr; addition
of 1 ft of barytes concrete block reduced the dose rate to about 1 mr/hr,

PART II. MATERTALS STUDIES

4, Metallurgy

Examination of the final eight INOR-8 forced-convection loops was completed,
and summary information for the program is reported. In general, maximum corro-
sion rates of INOR-8 by fused salts at 1300°F ranged from 1/2 to 1 mil in 20,000
hr for both LiF-BeFs; and NaF-BeFo systems. The attack was in the form of a
pitted surface layer.

The compatibility of molybdenum sheet with the materials in the MSRE system
has been tested in thermal convection loops. No significant attack was observed
on the metal parts of the system; however, a deterioration of the mechanical
properties of the molybdenum was noted.

Corrosion studies were started in order to test the effect of the oxidizing
impurities in fused-salt mixtures. Tests designed to establish the effect of
moisture were completed and are being examined.

Work was continued in order to determine the solubility limits of chromium
plus lron in nickel«base alloys contalning 18% molybdenum over the temperature
range of 900 to 2000°F. A phase boundary for this metal system has been
established.

The 'temperature range of melting was investigated for various heats of INOR-3,
and data are presented that show solidus temperatures to be higher than nil duc-
tility temperatures reported by Rensselaer Polytechnic Institute,

Specific heat of annealed INOR-8 was determined by direct calorimetric
measurements and the data reported. An anomalous rise of about 20% was observed
at approximately 600°C.

The total hemispherical emittance (t.h.e.) was determined for INOR-8 in the
bright-finished, matte, and oxidized conditions. INOR-8 in the oxidized condition
was found to have a t.h.e. at 600°C of 0.76, compared with 0.24 for an unoxidized
surface value.

Studies were conducted to circumvent the problem of cracking and microfissur-
ing observed in the welds of certain heats of INOR-8 material. The problem is
associated with improper melting practices used by vendors in pouring the initial
ingots. By using base metal and weld metal originally poured under staisfactory
conditions, sound, crack-free welds possessing good mechanical properties can be
made.

An investigation is under way to develop brazing procedures suitable for
remote fabrication operations. Brazed joints possessing good shear strengths at
elevated temperatures were made, and a joint design suitable for remote operation
was developed.

A program has been started to determine the strain fatigue behavior of INOR-8.
Data at 1300 and 1500°F are reported with a plot of Coffin's equation.

Molten-salt permeation tests with different grades of graphite indicated
that increasing the diameter of a graphite rod or fabricating it in the shape of
a pipe can decrease its resistance to impregnation by molten salts.

Tests showed that oxygen contamination can be removed from a moderately
permeable grade of graphite by exposing it for 20 hr to the thermal decomposition
products of NH4F:HF at temperatures as low as 930°F. The tensile specimens of
0.040~in.-thick INOR-8 exposed to this same oxygen-purging atmosphere developed
a reaction layer <0.0005-in. thick. The reaction layer did not alter the prop-
erties of the INOR-8.

5. In-Pile Tests

A molten-salt-fuel capsule experiment, ORNL-MIR-4T-3, has been operating at
the Materials Testing Reactor from May 5 to July 24 and is now at Battelle
Memorial Institute for postirradiation examination., The four capsules contained
fuel in AGOT, fuel impregnated, or R~-0025 unimpregnated graphite "boats". Samples
of molybdenum, pyrolytic carbon, and INOR-8 were irradiated in contact with the
fuel to temperatures to 900°C.

6. Chemistry

The phase diagram for the ternary system NaF-ThF4-UF4 has been finished;
this completes the systems limiting the quaternary Nal'-BeF,~-ThF4-UF4 and provides
interesting comparisons with LiFF-BeF,-ThF4-UF4, in which solid solutions arising
from the interchangeability of UF4 and ThF4 are much less cormon.

Studies of the crystallization of the MSRE fuel show that with fast cooling
a nonequilibrium path is followed along which the equilibrium primary phase,
6LiF«BeFo«ZrF4, fails to nucleate. Slow cooling leads to considerable segregation
of the MSRE fuel (LiF-BeFp-ZrF4-ThF4-UFy, 70-23-5-1-1 mole %), with UF4 concen-
trated in the last liquid to freeze. Since the first phases that freeze out are
viii

rich in ZrF4, the depleted residual liquid can, under some conditions, deposit
UO0s. Small segregated regions are produced vwhen frozen plugs of fuel are used
as freeze valves.

Much difficulty has been encountered with fuel studies that involve the
sampling and the measuring of oxide content in the range 100 to 1000 ppm. A
suitable resolution of the problem of oxide analyses is being sought.

An spparatus for measuring the surface tension of fluoride melts has been
constructed for use in studying the wetting behavior of salts with respect to
graphite. Graphite withstood at least a mild exposure to cesium vapor without
noticeable alteration of its interfacial behavior toward fuel.

A large-scale move from laboratories in the Y-12 Plant to new quarters at
ORNI: interrupted much work on MSRE problems in both the Analytical Chemistry and
Reactor Chemistry Division. Methods for analyses of the MSRE cover gas are under
development, as are improved treatments for fuel and coolant purificationm.

T. Engineering Research

The enthalpy of the coolant mixture LiF-BeF, (68-32 mole %) was determined
over the range 50 to 820°C. For the liquid, the heat capacity varied from 0.48
cal/ge°C at 500°C to 0.66 cal/g+°C at 800°C. The solid-liquid transition was not
sha;ply defined; the heat of fusion, evaluated between 360 and 480°C, was 151.k
cal/z.

In order to further clarify the heat-balance discrepancy noted in the heat-
transfer studies with the LiF-BeFo-ThF4-UF4 (67-18.5-14-0.5 mole %) mixture, the
enthalpy of the liquid was redetermined, using a sample of circulated salt.
Despite significant differences in composition, the enthalpies of circulated and
uncirculated salt samples were equal within 3.5%. In contrast, the heat capaci-
ties showed a deviation of as much as 10% at the extremes of the temperature
range (550 to 800°C); mean values of the heat capacity were identical (Eb =
0.335 cal/g-°C).

Interpretation of the data obtained in the study of heat transfer with the
LiF-BeF o=UF4-ThFg (67-18.5-0.5-1L mole %) salt mixture was continued, with pri-
mary emphasis on the evaluation of the abnormal heat balances observed. Exami-
nation of the data and the analytical procedures suggests that additional measure-
ments of the heat capacity and density of this salt mixture are needed to resolve
possible errors in the convective heat gain.

8. Tuel Processing

Compounds of SbFs with KF, AgF, and SrF, were prepared by reacting the com-
ponents in anhydrous HF. Products were AgSbFg, KSbFg, and a compound that may
not have been stoichiometric in the case of SrFo.

The material prepared by reacting NaF with MoFg in HF had the approximate
composition MoFg+5NaF plus some HF; attempts to remove the HF by evacuating the
container appeared to remove some of the MoFg as well,

Reactions of LiF, NaF, and KF with UFg, all in HF solutions, yielded yellow
or orange solids on evaporation of the solvent, The solids contained HF and much
less UFg than the anticipated complexes, probably because of evaporation of UFg
during the HF evaporation.
A single experiment to test the possibility of separating rare earths from
ThF4 in MSBR blanket salt by dissolving the rare earths in HF contalning SbhFs
was unsuccessful; the HF-5bFs solution dissolved neither rare earths nor thorium.
CONTENTS
SUMMARY vovuvveesonaseannsanee teseecans D 111
PART T. MSRE DESIGN, COMPONENT DEVELCOPMENT, AND ENGINEERING ANALYSIS
1. MORE DESIGN suveveceososesoaosasossstosonsosssossssssscssnsnsosesssansnsaas
Introduction s.veevvriecicscrsanenssssrsnonsnns ceseane vesesasones

1l
L
Reactor Core and Vessel ...civveene cessans Cetsssreresrasrensans e 2
Primary Heat Exchanger ........ Gt eesresessessstesessaarseraorenas A
A
4

PFHHRRRHREHERF
O =3 W Fw o

Radiator ...veevveenns e

Fuel-52l1t Drain Tanks iteveeessvsaessvosestonssnssssssossaasssssas
Equipment Layoul sueveeeesecssocsessavsosnessssssasssascososssssnses 8
Cover-Gas SYSTEIM .eesrsreevisosssstsosssesectosesnssnsscnsssssns s 11
System Heaters tveevertesecsnssscsossotascccsnnsss tesserevacsseanses 12
Design Status of Remote-Maintenance Systems ..cevecectirsecncreosss 13
1.9.1 Maintenance-Design Systems ....... tasecssserserrecsannans 13
1.9.2 Remote Maintenance by ManipulatorsS ..eeeeesecoessocsssces 13
1.9.3 Remote Maintenance by Manual Operations .eeeevseccecssos 14
1.9.4 Assembly Jigs and FixXBUTesS .eeeeverevosasocorasenossonss 15
1.9.5 Graphite-Sample Removal .uvveeeerornoscssrcsssoscsssnssonsse 15
1.10 Reactor-Control DesSigil teeeeressesccrescerescosscascsasossassocnsess 16
1.11 Decsign Status of Building and Site .vieeeceveessesocasecscnncsess 19
1.12 Reactor Procurement and Installation ..eeeeseseessevencveossveose 19

1.12,1 Demolition and Minor Alteration Work

to Bullding 7503 teeeeesasessctssescsensetscasenancnes 19
Major Modifications to Building 7503 ..... cesecsacseaa .o 20
Procurement of Materials .......... tevecennasarscasrsacs 20
Procurement of Components ...... tessessescesseraneserens 20
Instrumentation and Controls Desigll veeseevrescsescocoone 21
Instrument Application Diagram ..eeesvecocsrseosceronanes 21
Electrical Control Circuitry s.veeserereecasosscaccnsras 2L
LayOUL tivreeeneorsvsonnsseosessscsosssscosossssssnsssnsnes 21
Main Control Board ...eeeesesseossosassssossssssssssnasas 22
MSRE Data-Handling Study seesesesccceassssarsssnossscsss 24
Single-Point Temperature Alarm SYyStemM ceeeesesrssscacssse 25

s
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PRRHRPRRFO
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wWwwwww et oD
OV FWNRR WD

1.13

2. COMPONENT DEVELOPMENT aeeeevsscosocosessssssscessossassssnssasascssassca 28
2.1 Freeze-Flange DeveloPmMENt .veieeesrsesncncsasscsssesvococesasssnassne 28
2.1.1 MSRE 5-in. Flanges tueevsvescccscacsosssssevsscossnacanas 28

2.1.2 Freeze-Flange-Seal Test Facility ceeecececeencessosoasan 28

2.2 Freeze ValVeS sueeesserscscscsstosnsesasstsssancsssssstssccasavancs 32
2.3 Heater TeStS ctesreearterrascocssensesssesosssnsacsocsssssscsnssossssas 33
2.3.1 Pipe Heaters ..cveeeseresarsseossnens fetecesesereserennnase 33
2.3.2 Core Heaters teveesesessessssrssoscosssssssassoscsnsananas 33
2.4 Drain-Tank COOLET'S «eeteecrsessorsosesotsosossoecssososnssosassences 37
2.5 Sampler-Enricher Development ......... tereaeaene tesasensncaseases 37
2.5.1 Sampler-Enricher CONCEDPt vuveverrcrrsosessoscosccasasonas 37
2.5.2 So0lder-Freeze VAlVE .ceeeeetcssresrsosvsosoccscsacsesasacacacas 37
2.5.3 Sample CapSUle .uieesscescsosveraorrasssosnasonssnsans ces 40
2.5.4 Sample-Transport Container and Removal Seal ....eeeees.. 0
2.5.5 Sampler-Enricher DeSifll seeveetesrsrserecessasassoasssascs 40

x|

3.

,

2.6

2,11

xit

MSRE Core Development .c.ccceeveesecenrsacsanss cessesercsedsnsranns

2.6.1

2.6.2

2.6.3

2.6.4
2.6.5
Helium Purification ....... e eseas s s s es st trasat e st us et s ea
Pump Development ....coceeeseevetssscsssssesrscsvnacsons ceseesnan

2
2
2
T

8.
8.2
8

1

.3

Heat Transfer Coefficients in the Lower

Head of the Reactor Vessel ........ cecsccsresarenrsenne
Fuel-Age Measurements in the Lower Head

of the Reactor Vessel ...ceeecitervanscsssssscense cees
Studies on the Disposition of Fine Particles

in the Lower Head of the Reactor Vessel ........ PN
Fluid-Dynamic-Induced Power Oscillations ...... cevesaane
Full-Scale MSRE Core Model ...iciieevrsacassescosossancss

MSRE F‘uelPLm]P.'l....l.l.l.l.l.l.l.'.t....l-..I....Clll
Water Test of MSRE Coolant PUMD ceeeecvoressccscacecanes
Advanced Molten-salt P‘-lmps e d & & 4 & 8 B B &0 & PSS AP R F R C SRS

Graphite-Molybdenum Compatibility Test ....... e
MSRE Engineering-Test LOOD ceeececceacerroressascscsssssaaceonsros
.10.1 Oxide-Flush Run s.cvececcecresoncscesrscsnacens seseesnas
.10.2 Freeze-Valve Operation ...ceceecececesscssccaas ceveesnae

2.11.4

.10.3 Level Indicator for Molten Salt (Test TOOD) eceveesseanns
.10.4 Graphite-Handling Facility .cecevveceecennnronensocacns ‘e

Maintenance Development ....... Cetetseserentasnasasneerens -
1.1 Maintenance Plan ...c.ccc... cessisessssrsessesessnrenans
1.2 Remote-Maintenance Mockup ...ccevevecscrornsnsrnsrcncsnns
2.11.3 MORE MOd€l ..vevevesncccssecscanasasesoscsrsssossnsscsssas .

Portable Maintenance Shield ....ccceeoveenscososcssnersss

Brazed-Joint Development ..eecvcecn. Cieeesectsasatsseseannans e
Mechanical-Joint Development ...ceccceecacse et esesssenss s n s
Steam Generator tececssrtsseseesssssscsssesossssssssassssenssssese
MSRE Instrument Development ...eeeevsrececscccsascrscssssscseacasna
.15.1 Pump Bowl Level Indicator ..ecceciesscasssnscsscrsnnsonns
.15.2 Single-Point Level Indicaltor ..ccecececssscsransosocssss
5.3 Temperature SCANNEY ..c.ieveseeerrosrosssansvrsssscsssssacss

2
2
2
2

.1
.1

5.k

Thermocouple Attachments and End Seals ...coveecencccsns

REACTOR :ENG]:NEERING ANALYSIS ¢ * 8 80800 ® 2 0 " 8 B &k 2 PN O SN S PEE SRS EN S S e esbe
ReaCtOr Physics 'E N EEREEEEE RN NN NI NI I B R R IR IR R I IR R IR I I I I A B
3.1.1

3.1

-F"L.AJI\)

}EIIAIJH]RGY ® 8 ¢ % & 068 080 LB P T LR PSS PR S S

b1

3.1.2

3.1.3
3.1.4
Nitrogen-16 Activity in Fuel Salt ...cecvvveneneen. eceiasasariae
Residual Activity in Pump Bowl and Heat Exchanger ......... ceesas
Reactor-Cell Shielding ..... Gt eceteesser et asus s et nses s s tane b
3.4.1
3.4.2

Analysis of MSRE Temperature Coefficient

Of Reactivity ceeeesererrencsesscscssosncsrscansnansens
Rod Worth, Flux, and Power Density Distributions

in MOSRE s.tiviesesenseneccccnnssncesanans cesasaeranse
Gamma -Ray and Fast-Neutron Heating in Thimbles .........
Reactors with INOR-8 Fuel TUDES sesscvaseosansscssoscnns

TOP Shield N N NI N R A B R I R Y R I R I R Y R I T I I I I IR R I R B L
Side Shield- * % & 8 & B OB F O B E S S AP A PRSP O NS SES eSO EPSE SR

PART II. MATERTALS STUDIES

mnamic-corrosion Stud.ies P I B B B B I I T T I R R BN Y R R Y I RN B N RN B BN BN N BN E S

Examination of INOR-8 Forced-Convection LoOPS +vivsseess
Molybdenum-Graphite Compatibility Tests .eeeecevsvansans
Compatibility of INOR-8 — 2% Nb and

Molten Fluorides ..cececeicescccscccseocorosnerennnssanns
Fluoride-~Salt Contamination Studies ......iccevevenseess

77

95

5.

6.

7.

8.

L.,2

=&
=

-
—3 O\

IN-PILE TESTS

5.1

INOR-8 Development .........

® s 8 9 v 0 9O r e e s 4 2 0 8 8 s 4 s 4 a8

h.2.1 Structural Stability of Nlckel Based 18%

Molybdenum Alloys

h.2.2 Temperature Range of Melting for INOR-8 ...ieeeveeersene
4,2.3 Specific Heat of TNOR-8 ..evereronnrorrrorosonosnrenenons
Total Hemispherical Emittance of TNOR-8 ..veeerrveerrnrresonensns
Welding and Brazing STudies ..ceevievrerenesescnsrennsssssserasnns
Lh.li,l  Welding of INOR-8 tuvuivienoroenconssssosensassasaacsnnes

L.k
b,
L.k

=

.2 Metallographic Examination and Bend Tests

ON WEldS suvevevesvesnencsasosscaconseascesosesssasnanns
Transverse Tensile and Creep Tests .vieierersvcnsncncass
Welding of INOR-8 to Stainless Steel

and Inconel ...cieveineenanancsons tesesesarssasnaseana

L,k,5 Remote BrazZiNg ceeeeeeeseeseoceroosoresstonssenesersaoess

Mechanical Properties of TNOR=8 .iiveveereronrnsrersesaonsrsnonsas

Impregnation of Graphite by Molten Salts ...iiivieoennnterennonees
Ammonium Bifluoride as an Oxygen-Purging

Agent for Graphite .......

h,7.1 Removal of Oxygen from Graphite ....eeeevvescsrcsoseanss
h.7.2 Effects of the Thermal-Decomposition
Products of Ammonium Bifluoride on INOR-8 ......cuve.s

Graphite-Fuel Capsule Experiments ...vieeeesessscesssasossnsenene

CHEMISTRY [ R N A A DR B I I B N I O A LT A I I T I I B I N Y I R Y BN B B R B I B I Y B B I R I R Y I I B B I I IR ]

6.1

6.2

6.3

6.k
6.5
6.6

ENGINEERING RESEARCH

T.1
7.2

FUEL PROCESSING

Phase-Equilibrium Studies ..

6.1.1 Systems Involving NaF with ZrF,, ThF4, and

UFg and Other Constituents ..cvesevevrrsvrocsersrenens
6£.1.2 The System NaF-ThF4-UF4 . veeeeeeeeneresososnasensnnnosoes
6.1.3 Stannous Fluoride as a Component of Reactor Fuels ......

Oxide Behavior in Fuels ....

¢ # o o 0 s OO b e bbb SO E S PSS #8008 09 00e

6.2.1 Behavior of MSRE Fuel on Freezing and Effect
of Segregation on Oxide Precipitation ........cecuenne
6.2.2 Oxide Content of Fluoride MeltsS ....eeevececreercnnncans
Physical and Chemical PropertiesS .s.veecereereroscososessarsconsss
6.3.1 Surface Tension Apparatus for Molten Fluorides .........
6.3.2 Volatilization of Iodine from the MSRE Fuel .....vevev0.
Graphite Compatibility seeeserersvesrsereerseascesorssosrescacsanenss
6.4.1 Effect of Cesium Vapor on Graphite ...e..eeveeerensnnsse
Fuel ProdUCtion seeveesreecstressasessscsosnsossossrsssssacssncsnssss
6.5.1 Purification Treatments ....eeeeeeeeeseecececaccaeacanas

Analytical Chemistry .......

6.6.1 Analyses of MSRE Fuel MixXture .eeeeeviecscasscorsnncnsss
6.6.2 Analyses of MSRE COVEr GB5 +cvvevevsorcsacsnncancananans

LI B I B I Y I B I B I B I B R I R BN BN R R R BN AN BN I R N Y B DN A B L B B B RN B R B R AR B A

Physical-Property Measurementis .ieeceiscerascsssocssorssssscssensns
Heat—TranSfer Studies ® B & & 3 8 4 & 0 & 8 S0P P ORI R ERNSEEEE S ER SR

0 8 0 8 & 4 8 0 0 8 F PR PSSR PRSP P AN PSSR SRS S s

Preparation and Analysis of Complex Fluorides of SbFs

with KF, AgF, and SrFpo seeeecetcevsorescscosssnoascssssssesanss
Preparation and Analysis of the NaF-MoFg CompleX ..cceveersceccecs
Preparation and Analysis of Complexes of UFg with LiF,

NaF, and KF seveveveecses .

0 & @ & 0 P 2 & & v é P B AR RSN Rr e

Attempt to Separate Rare Earth Iluorides from MSBR

Blanket Salt by SbFs in HF

® & 8 0 P8PSR T DN N LSRN e

136

136

PART I. MSRE DESIGN, COMPONENT DEVELOPMENT,
AND ENGINEERING ANALYSIS

1. MSRE DESIGN

1.1 TINTRODUCTION

Just as the semiannual report for the period ending February 28, 1961 was
being issued, it was decided to review the fuel-circuit design with regard to
two guestions: (1) should the reactor vessel have a flanged top head, and (2)
should the pump be removed from the top of the reactor.

The additional design studies indicated that it was not practical to flange
the reactor head but that it was probably worth the additional complexity in
pump mounting to remove the pump from the top of the reactor vessel. It was
believed that the only advantage provided by flanging the reactor top was in the
ability to remove the core independent of the reactor vessel. This was thought
not to be of sufficient Importance to warrant the additional complexity of a
large flanged seal.

Moving the pump off the reactor, however, was thought to be quite desirable
in order to facilitate graphite sampling and control-rod operation. The com-
plexity of a more involved pump mount was thought to be justified by the
accessibility to graphite and rods which it provided. Accordingly, the design
was changed to include the new fuel-pump location.

The liquid-poison tubes described in the last report were eliminated. With
the pump moved off the reactor and with the top exposed, the more conventional
and tested solid control rods were preferable.

In all other basic concepts the MSRE system has remalned the same, and
detailing of building alterations and component design has proceeded. The
result is that the building-alterations designs and component designs, together
with specifiications for these designs, have been issued to prospective bidders.

The electrical heater design is not yet finished, but work is proceeding
without difficulty. Instrumentation is also incomplete but is being carried
forward as rapidly as necessary to meet construction and installation schedules.
Design of both these phases can be accomplished during the building-alteration
period and the component-fabrication period.

There will be a decline in design manpower effort as of October 1, 1961,
and for the remainder of the fiscal year the design manpower will be carried at
a reduced rate. This effort will consist in effecting necessary minor changes
and completing auxiliary-systems designs.
Major items of design are discussed in more detail in the paragraphs which
follow.

1.2 REACTOR CORE AND VESSEL

The general configuration of the core and contalner vessel has not been
changed, but some changes have resulted because of moving the pump off the top
of the reactor. These changes involve: (1) the reactor discharge line, {2) the
upper -head neutron shield, (3) the graphite-sampling access, and {4) the control
rod penetrations. The reactor i1s shown in Fig. 1.1.

The reactor discharge is through a 10-in. pipe, rising vertically from the
center of the top head. This 10-in. pipe has a 5-1in. slde outlet leading off to
the pump suction. The 10-in.-diam vertical section terminates in a flanged top
with a metallic O-ring gasket. Into the 10-in. pipe, and extending down to a
point above the 5-in. discharge tee, is a hollow plug which is welded to the
mating flange. The purpose of this plug is to provide removable penetrations
for control rod thimbles and a graphite-sample port. There is an annulus between
the plug and the vertical pipe and fuel entering this annulus can be frozen to
form a salt seal. In this manner the ring-seal flange 1s never required to hold
molten salt but becomes a gas seal only. Air 1s circulated on both sides of the
annular space. When the air is unheated it serves as a coolant to establish
this freeze plug. The air passes through an electrical Turnace so that it can
be heated, when desired, to thaw the plug after the reactor has been drained so
that the plug can be removed.

Removing the pump off the top of the reactor vessel made it unnecessary to
put the massive INOR-8 shield plug in the top plenum of the reactor vessel.
This plug has been eliminated completely, and the resulting space will be filled
with fuel.

Provision has now been made for control rods and graphite samples in the
center of the reactor core. The four graphite stringer positions at the corners
of the exact center of the core are omitted from the core matrix. INOR-8
thimbles are inserted in three of these positions. The upper ends of the
thimbles are welded in the bottom of the plug which fills the 10-in.-diam
vertical pipe on top of the reactor. These thimbles then provide penetrations
into the core into which solid contrel rods can be inserted for control of the
reactor.

The fourth position in the core is arranged to accommodate four 7/8-in.-diam
graphite rods. Any one of these rods (which extend only down to the midpoint of
the reactor, the lower half containing a standard stringer of half length) can
be withdrawn from the reactor for examination.

In order to remove the sample graphite a 3-in.-diam access tube extends up-
ward from the base of the flanged plug. A bolted flange seal on top of this tube
can be removed for access to the core. This 3-in. riser tube has a plug,
similar in principle to the one used on the 10-in.-diam pipe, with a similar
frozen-salt annular seal.

The plug in the graphite-sampling port provides a holddown of the four
samples. When the flange is unbolted and the plug is removed, the samples are
free and can be removed by the graphite-sampling device. In the event that the
condition of the sample warrants more extensive examination, main graphite
stringers can be remcved by opening the 10-in. flange and removing the entire
control-rod-thimble assembly.

3

UNCLASSIFIED
ORNL-LR-DWG 61097

AIR QUTLET AR INLET
_ FLEXIBLE CONDUIT TO
: CONTROL ROD DRIVE {3)

SMALL GRAPHITE SAMPLE ACCESS PORT
CONTROL ROD COGCLING AIR INLETS

CONTROL ROD COOLING AIR OQUTLETS

COOLING JAGKET AIR INLETS
COOLING JACKET AIR QUTLETS

AGCESS PORT COOLING JACKETS

FUEL OUTLET REACTOR ACCESS PCRT

SMALL GRAPHITE SAMPLES

CONTROL ROD THIMBLES (3) HOLD DOWN ROD

LARGE GRAPHITE SAMPLES (5)

CORE CENTERING GRID

FLOW DISTRIBUTOR

STRINGER

GRAPHITE-MODERATOR \

FUEL INLET / 5

REAGTOR GORE GAN — |

REACTOR VESSEL — ]

I

“'
T i,l
# l

(=)

111 |
TE l ,fi’l
L

| . ?‘;I!

/ > | L I 4 6
_ = ' 20 ‘

ANTI-SWIRL VANES
MODERATOR

VESSEL DRAIN LINE SUPPORT GRiD

Fig. 1.1. Cutaway Drawing of MSRE Core and Core Vessel,
The tubes for liquid poison, previously considered for reactor control, were
abandoned in favor of rods similar in design to those used successfully in the
Aircraft Reactor Experiment. The control design is discussed in Sec. 1.10.

1.3 PRIMARY HEAT EXCHANGER

Heat transfer and pressure-drop calculations for this component, along with
preliminary stress analysis, have been reported in the component design report.l

Drawings showing configuration and support and tiepoint locations have
Leen prepared and were included with the component package released for bidding.

1.4 RADIATOR ]

The design of the air-cooled radiator c0il? and enclosure was approved, and
drawings and specifications were released for bidding. -

The door drive mechanism and superstructure, which were not included in the
coil-and-enclosure bid package, were redesigned. The single counterweight for
the doors was eliminated. Each door was suspended from a drive shaft by means
of a wire rope and sheave assembly. A Tflywheel was mounted on each drive shaft
to prevent damage to the door by causing it to fall slowly when released. An
over-running clutch was provided for each flywheel to allow the flywheel energy
to be dissipated through friction. A magnetic clutch and a magnetic brake were
placed on each drive shaft to permit individual raising, lowering, and positioning
of the radiator doors. A chain drive system was retained in a modified form.
The redesigned radiator assembly is shown in Fig. 1.2.

1.5 FUEL-SALT DRAIN TANKS

Several changes were made in the design of the fuel-salt drain tanks. The
most significant one was the method of connecting the steam and water lines from
the bayonet heat-removal units to the steam dome and water supply.

The number of bayonets was reduced from 40 to 32, and the steam dome diame-
ter wag increased to 48 in. The steam lines enter the steam dome through the
lower head, and the bayonet water supply tubes are concentric with the steam
lines except at their inlets inside the steam dome. The water supply tube in-
lets are through the wall of the steam outlet nozzles. The steam dome serves as
an intermediate water reservoir during operation. The water inlets are staggered
on two elevations to permit use of only half the bayonets, if desirable, by
controlling the condensate return rate. The condensate returns to the steam
dome from a reservoir located with the condenser outside the drain-tank cell.

The bayonets rest on a support plate which is attached to the steam dome.
This permits removal of the steam dome and bayonets without removal of the tank.

The steam dome has a penetration through its center, through which passes
a nozzle from the salt tank. This nozzle provides access for inspection and
sampling and for inserting level probes in the salt tank.

Figures 1.3 and 1.4 depict the new arrangement of the bayonet heat-exchanger
units and the complete fuel-salt drain tank, respectively. This arrangement *

DRIVE CHAIN

[/

DOOR DRIVE MOTOR AND

GEAR REDUCER —._ ', ‘

oo
WIRE ROPE SHEAVEj—w\\
s T s i‘

. LY

SUPPORTING STEEL 7
WIRE

DOOR CAM GUIDE ——T

BLDG. 7503, FIRST FLOOR
{ELEV. 852 ft - Qin.)

-7 | INLET DOOR
|

AIR BAFFLES
/ \
/ \
AIR INLET DUCT |
AN

/AR DUCT FLANGE—//f &
| -

P

MAIN AIR BYPASS DUCT
/

/
£

[~

Fig. 1.2. MSRE Radiator Coil and Enclosure.

UNCLASSIFIED
ORNL-LR-DWG 55841R

PENTHQUSE

.« T~AIR OUTLET DUCT

RADIATOR ENCLOSURE

B

RADIATOR TUBES

UNCLASSIFIED
ORNL-LR-DWG &0B3BA

!' STEAM OUTLET
h

WATER INLET —— /

N

STEAM DOME FLEXIBLE HOSE

LOWER HEAD

BAYONET SUPPORT PLATE

DRAIN TANK HEAD

Fig. 1.3. Bayonet Cooling Thimble.

UNCLASSIFIED
ORNL-LR-DWG 61719

INSPECTION, SAMPLER, AND
LEVEL PROBE ACCESS

STEAM QUTLET
STEAM DOME

CONDENSATE RETURN

WATER DOWNGOMER INLETS

BAYONET SUPPORT PLATE CORRUGATED FLEXIBLE HOSE

STEAM RISER

0

0 T

STRIP WOUND FLEXIBLE

BAYONET SUPPORT PLATE
HOSE WATER DOWNCOMER

HANGER CABLE

GAS PRESSURIZATION gar

AND VENT LINES INSTRUMENT THIMBLE

FUEL SALT SYSTEM
FILL AND DRAIN LiNE

SUPPORT RING

1 |
|
FUEL SALT DRAIN TANK
BAYONET HEAT EXCHANGER ‘ '| |
THIMBLES (32) TANK FILL LINE
3 |
|
2q |
'
8
ON
Q \\\\r
lN 8 L .
CH ™y
E i
S g

THIMBLE POSITIONING RINGS

FUEL SALT SYSTEM
FILL AND DRAIN LINE TANK FILL LINE

Fig. 1.4. Primary Drain and Fill Tank for MSRE,
simplifies the unit, lowers its fabrication cost, improves the water supply sys-
tem, facilitates maintenance, and reduces maintenance cost.

Other minor dimensional changes were made.

1.6 EQUIPMENT LAYOUT

One maJjor change in layout design has been made. The fuel pump has been
moved off the top of the reactor, and, therefore, the pump mount must provide
Tlexibility in order to allow for movement imposed by thermal expansion of the
lines. Figures 1.5 and 1.6 show the plan and elevation of the layout of the
salt circuits.

The fuel system now permits the removal of any component, although, to
remove the reactor or pump bowl, shifting of some other components will be
required.

The piping system for the revised layout has been analyzed for stresses
and has been found to be conservatively loaded by a good margin. The coolant
piping system has also been analyzed by the same computer code and was found
to te satisfactory.

A satisfactory layout has been accomplished for the secondary piping (for
0il, gas, and cooling water) for the fuel pump. The final design makes use of
a ring-joint flange for each line. These flanges are grouped around the pump
in a circular pattern to facilitate remote make and break of these containers.

A satisfactory layout of service disconnects plus spares has been finished
for the electrical heaters and the thermocouples.

Details of containment penetrations for all services (gas, electric power,
lubrication, and instrumentation) have been designed. The reinforcement of the
vessel wall for these penetrations was designed and analyzed for stresses.

The nuclear instrumentation penetrations have been reduced to one large
tube in which will be housed all the radiation counters to be used on the equip-
ment .

The cooling air which is used to establish and maintain the freeze valves
and which alsc cools the control rods and top surface of the fuel pump is
supplied from positive-displacement air pumps located in the special equipment
cell. This air is recirculated, and the heat is removed by the space coolers
within the reactor contalmment vessel.

A1) structural members for component mounting within the containment vessel
have been designed. This includes the reactor thermal shield, which is used as
the structural support for the reactor.
AUX. EQUIP ROOM

FUEL PUMP LUBE OIL PK.
NEUTRON INST. TUBES
AIR DUCT

SPECIAL EQUIP. ROOM
FUEL PUMP

HEAT EXCHANGER
REACTOR

COOLANT PUMP
COOLANT DRAIN TANK
RADIATOR

hSRLE L -any. )
. X L

STACK
BLOWER HOUSE

WASTE TREATMENT CELL
. EQUIP, STORAGE CELL x

DECON. CELL
FUEL TRANS. CELL

SPENT FUEL STORAGE TANK{;‘E

DRAIN TANK NO. 1

. DRAIN TANK NO. 2

FLUSH TANK

AUX. INST. ROOM
| L A i P LT A |

o

UNCLASSIFIED
ORNL-LR-DWG 60680

Fig. 1.5. MSRE Plant Layout, Plan,
UNCLASSIFIED
ORNL-LR-DWG 60681

1. 30 TON CRANE 8. WATER/SAND ANNULUS  15. HOT STORAGE
2. 7 & 3 TON CRANE 9. CONTAINMENT VESSEL  16. DECON. CELL
3. MAINTENANCE CONTROL ROOM 10, FUEL DRAIN LINE 17. SHIPPING CASK
4. COOLANT PUMP 11. RADIATOR 18. FUEL STORAGE TANK
5. FUEL PUMP 12, BY-PASS DUCT 19. DRAIN TANK NO. 1
6. HEAT EXCHANGER 13, COOLANT DRAIN TANK  20. DRAIN TANK NO, 2 D)
7. REACTOR VESSEL 14. STACK 21. FLUSH TANK
N ||
a ® |
I 1l
L.J -
o -
"~ '
: : :
:, ) .~ —
=, In o

Fig. 1.6. MSRE Plant Layout, Elevation.

(I RN YIRS TR S
P AR .
s N et At

o 7 EEa
Y A

TR AT
T e
S g

ot
11

1.7 COVER-GAS SYSTEM

The original design of the cover-gas system provided the alternatives of
recycling or discarding contaminated helium at the option of the operator.
Since the equipment for cleaning and recycling the gas requires some additional
development as well as additional expense, installation of a recycle system has
been deferred.

The conceptual design of the components of the supply-and-discard system
was completed, and the construction drawings were started for the charcoal beds,
the cell blowers, and the helium dryers. A drawing showing the general location
of components and the arrangement of interconnecting piping was issued for review
and comment. Design memoranda on the charcoal beds and the leak detector system
were issued. The only basic change made in the off-gas system was in the xenon
holdup time provided by the charcoal beds. The design now calls for a xenon
holdup time of 9C days at a sweep-gas flow rate of 4.2 liters/min (6000 liters/
day).

The leak-detector system will be used to monitor flanged joints in the fuel-
circulating system and in various auxiliary lines. Helium pressure, higher than
system pressure, will be applied against joint-sealing surfaces. Leakage, if
any, will be into the salt system, and leaks will be detected by loss of helium
pressure. (See Figs. 1.7 and 1.8).

UNCLASSIFIED
ORNL-LR-DWG 63241

LEAK-DETECTOR LINE
FROM LEAK-DETECTOR
STATION

O-RING GASKET
w /% Z Z
& ?_,

SE’F‘{%*NG ‘\LEAK DETECTOR LINE
SURFACE TO NEXT FLANGE, OR
MAY BE CAPPED OFF.

l
TR
.

FLANGES

ANY LINE REQUIRING
LEAK-DETECTOR SERVICE

REACTOR
COMPONENT

REMOTE-DISCONNECT FLANGE

]

I - -

i 3 FROM

L ————w— | EAK-DETECTOR
STATION

Fig. 1.7. Leak-Detector Flange and Method for Using One Leak-Detector Line
to Serve Two Flanges in Series.

12

UNCLASSIFIED
ORNL-LR-DWG 63242

TO PRESSURE INDICATOR
AND PRESSURE ALARM
T IN CONTROL ROOM

jw—— 250-psig HELIUM HEADER
|

PE -l
O MINIATURE BELLOWS-SEALED VALVE,
SIMILAR TO HOKE 480 SERIERS OR

EQUAL
TUBING, ¥g-in.0D x
’//’ooss—m.meL

HIGH-SENSITIVITY {0.05 psi)
D/p CELL (LOCAL)

REFERENCE-PRESSURE LEG
= 1/2 in. SCHED 40 AVERAGE OF 50 ft
—{><}~
EQUALIZING VALVE RANGE ; 0-5C cm3/min
MAX AP; Spsig
—D<—
{50 psi
/ P9 —D><~ TO LEAK-DETECTOR
S~ - < JOINTS
— 1/4 in. SCHED 40 | ] |
VOLUME OF SYSTEM DOWNSTREAM — <~
OF THIS VALVE TO BE KEPT AS
SMALL AS POSSIBLE TO INCREASE [

SENSITIVITY

— TO OTHER LEAK-DETECTOR STATIONS

L

Fig. 1.8. Leak-Detector Station.

1.8 SYSTEM HEATERS

The design of a typical line-heating unit has been changed from that previ-
ously reported. The present design is shown in Fig. 1.9; remote-maintenance
practices dictated this modification in the unit. The heater as shown can be
removed from the pipe by breaking the electrical disconnect and engaging the
lifting eye with a tool. A simple lifting motion will then remove the unit.

The unit consists of heating elements of resistance coils (Nichrome V) embedded
in fused alumina, thermal insulation, and a metal container or housing. The
thermal insulation is a light-weight felt of alumina and silica. Reflective
foils are placed at intervals through the insulation. The horizontal piece does
not contain heating elements. It serves only as thermal insulation and as a
support for the top unit. Process-piping supports penetrate it.

Heating elements are provided in the radiator enclosure. Flat, ceramic-
embedded heating elements are mounted where possible around the radiator coil,
These elements have a combined capability of 30 kw and will be used for pre-
heating and during periods of zero-power operation. At all other times they will
be energized at a reduced voltage. Tubular heating units are placed parallel to
the tubes and are located between the tubes and the radiator-enclosure doors.
The tubular elements will have their full-rated voltage applied when the doors
are closed or dropped. The surface temperature of these units will reach 100C°F
in 1.5 min, and the heaters are installed to prevent freezing of the outer row
of tubes in a lossg-of-flow incident. The tubular heaters can also be used in
preheating and during barren-salt operation.

UNCLASSIFIED
ORNL-LR—-DWG 63243

LIFTING BAIL-——~__
Y

HIGH-TEMPERATURE
THERMAL INSULATION >

HEATING
ELEMENT ——

\ SECTION A-A

Fig. 1.9. Typical Pipe-Heater Unit.

ELEVATION

1.9 DESIGN STATUS OF REMOTE-MATNTENANCE SYSTEMS

1.9.1 Maintenance-Design Systems

Two distinct systems of maintenance are provided, one of which permits
working directly through the shielding roof plugs for replacement of small com-
ponents or the preparation for removal of large components, and the second
requires operation of the crane and a manipulator from a remote control room.
Although it is theoretically possible to do all maintenance tasks by the fully
remote method, the use of semidirect methods will greatly reduce the time
required to perform many small jobs. Furthermore, the capital cost of Building
503 shielding will be greatly reduced since it will not be necessary to provide
protection to control room and office areas for long periocds of time while all
roof plugs are removed from above the reactor cell for remote maintenance.

1.9.2 Remote Maintenance by Manipulators

The completely remote maintenance system, to be used for the replacement of
those major components which require removal of major porticns of the cell
shielding, requires that personnel, operating in a shielded area, can see and
manipulate the cell shielding, the component disconnects, and the components.
The elements of this system have been designed or specified. The maintenance
control room, above and to the right of the reactor cell (see Fig. 2.28), pro-
vides a shielded work area with windows for direct vision of the reactor cell,
drain cell, and crane bay. Located in it are controls for the 7.5-ton and 30-
ton cranes, the manipulator, air-operated tools, and the television system,

The 7.5- and 50-ton cranes will be modified to permit control of five speeds
and three speeds, respectively, in all directions, from the maintenance control
room. The 30-ton crane will be further modified by the addition of a motorized,
360°-rotating hook and a remote-reading load cell.

The manipulator and bridge have been specif'ied and the rails designed to
permit the manipulator to be used over the reactor cell and the maintenance-
practice cell.
14

The television viewing system provides two pairs of stereo cameras, each
pair with its own pan-and-tilt mechanism and vehicle to carry it around the
track.

Specifications have been written for automatic 1ift tongs (Fig. 1.10) that
will engage, l1ift, lower, and disengage the lower shield beams and shield support
beams, using only the remotely operated crane.

UNCLASSIFIED
ORNL-LR-DWG 61193

A\ / MOTORIZED CRANE HOOK

SHIELDING PLUG—"

Fig. 1.10. Remotely Operated Lifting Tongs for Shielding Plugs.

1.9.3% Remote Maintenance by Manual Operations

This semidirect maintenance system permits personnel to use manually operated
tools through the cell shielding and to view directly through small windows or
with optical aids. Conceptual design of and procurement for the portable mainte-
nance shield has been completed. The shield covers the opening left by the
removal of two lower shield beams from the reactor or drain cell and provides a
movable opening through which tools, windows, and lights can be inserted.
Several hand tools have been designed for use with the shield. -

15

Several special problem areas have had to be considered separately. A
fixture has been designed for the removal and precise replacement of the rela-
tively heavy sampler-enricher spoocl piece with expansion bellows. The mechanical
tools for making a remotely brazed joint have been designed and detailed for
construction. These include machines for holding the pipe, cutting it, taper-
machining the male stub, assembling the Joint, and holding it during the brazing
cycle,

1.9.4 Assembly Jigs and Fixtures

The design of the jigs and fixtures for the assembly of the reactor, fuel
circulating pump, heat exchanger, and associated piping is progressing. The
conceptual work is almost complete, and detail design has been started. A small
model has been built. The jig (Fig. 1.11) will consist of four basic parts plus
additional supports and bracing. After the completion of the fabrication of the
initial system, the jig can be broken down into four parts for storage. If it
becomes necessary to fabricate replacement components, only two parts of the jig
will be required for the duplication of any given component .

UNCLASSIFIED
ORNL-LR-DWG 611524

FREEZE FLANGE (TYPICAL) MOTOR

;! COOLANT PIPE

= \\\<\ PUMP BOWL
SECONDARY FLANGE aS AT
CLAMPING SECTION YA )
{ DETACHABLE) \ = L
\ Ne

COOLANT PIPE

L &— PUMP ASSEMBLY SECTION
(DETACHABLE )

+_—— THERMAL SHIELD

HEAT EXCHANGER
ASSEMBLY SECTION
{ DETACHABLE)

%

REACTOR VESSEL

BASIC FRAME

REACTOR SUPPORT
{FOR FINAL JIGGING)

NOTES: DRAIN LINE

1. JIGGING FOR AUXILIARY LINES, CLAMPS FOR PIPE AND PIPE FLANGES,
LEVELING DEVICES AND CERTAIN FRAME BRACES ARE NOT SHOWN.

2. STRESS RELIEF AND THERMAL CYCLING ARE DONE BEFORE
REMOVING FINISHED COMPONENTS.

Fig. 1.11. Concept of Fabrication Jig for Primary System,

1.9.5 Graphite-Sample Removal

The preliminary design of a proposed method for removing the small (7/8-in.-
diam) graphite core samples has been completed. The graphite sampler consists
of':
16

1. An eccentric steel plug 32 in. in diameter and 18 in. thick [Fig. 1.12
(3)] fitted with seals and buffer gas to prevent the escape of activity
or the introduction of air to the reactor system while samples are being
removed and replaced.

2. A standpipe [Fig. 1.12 (0)] bolted to the sample access flange and to
the lower shield plug for containment purposes. An exhaust line will
be attached to the lower end of the standpipe, and nitrogen will be
introduced at the upper end [Fig. 1.12 (5)].

Tnside the standpipe will be located two containers for the graphite samples,
one containing the new graphite sample to be installed and one for the graphite
sample that is to be removed. In addition, a bracket [Fig. 1.12 (8)] for holding
the sample-access flange and holddown assembly [Fig. 1.12 (7}] while samples are
being changed will be mounted on the side of the standpipe.

Viewing will be through a lead glass plug, and general lighting will be pro-
vided by a floodlight [Fig. 1.12 (17)] inserted in the eccentric steel plug. A
zirconia light will be used to project a light beam inside the sample-access
pipe for handling the graphite samples.

1.10 REACTOR-CONTROL DESIGN

After a preliminary assessment of MGRE hazardsbr it was concluded that neither
the amount nor the rate of addition of excess reactivity from any source would
create a hazard. Therefore, the need for a fast-acting, multichannel safety
system is eliminated.

Control rods have been designed to shim for reactivity changes as follows:

1. ZXenon 0.013%
2. Fuel-pump speed 0.002
3., Power coefficient 0.002
4. Burnup between fuel additions 0.002

5. Temperature control (~300°F)
and fuel penetration 0.027
Total reactivity C.0LO

Three control rods are provided. Each contrecl rod has a maximum worth,
when inserted with all other rods withdrawn, of 0.025 8k/k. Their combined
worth is 0.0LO 8k/k. The maximum rate of withdrawal for a single rod is
0.0002 8k/(k-sec). With three rods inserting as a group, the maximum rate
of poisoning is 0.0005 &k/(k-sec) .

All rods are identical, and any one, but only one at a time, may be used
as a servo-operated shim for automatic control purposes.

Control rod design has been altered by substituting solid, mechanically
driven rods for the liquid poison tubes originally considered.

Fach poison rod consists of a series of short tubes of B,C which are
sheathed with INOR-C and mounted concentrically on a flexible drive cable
(Fig. 1.13). The resulting configuration is the same as a so0lid rod 1 in. in
diameter. These rods move inside of vertical thimbles located centrally in the
core (Fig. 1.14). The assembly of three rods is part of a larger flange-
mounted assembly mounted on top of the reactor vessel. The rods are air cooled,
but loss of air cooling will not render the rods inoperative.
UNCLASSIFIED
ORNL-LR-DWG 61027A

FLOOD LIGHT INSERT

Fig. 1.12. Schematic of Graphite-Sample Removal System.

]

18

fmmt—— 4,360 in. DIA (REF) ———=

5ft 5in.

|

1% in. ——————————=]

L i e ] Wi,

Y N
Y
v

N

\

W

NZZ

N 7777777 77 77 777777 7 77 7 7 7 A N 7 7 AN A

PC 7 NUT

Y

Fig. 1.13. Vertical Section of Poison Control Rod.

UNCLASSIFIED
ORNL-LR-DWG 6i152A

| —— PC 8 SUPPORT SPIDER

GUIDE TUBE
1% -in. OD x 0.065-in, WALL

——— PC 10 SEAL RING

| —s— THIMBLE 2-in. OD x 0.065-in. WALL

INOR-8

/PC 9 FLEXIBLE INCONEL HOSE

L —— AIR FLOW: 3.3 cfm

| __—— PC { B4C CONTAINED IN INOR-8
CONTAINER—1.040-in. OD

—— PC 8 SUPPORT SPIDER

|_—GUIDE BARS, 4 AT 90°

— PC 6 FLEXIBLE DRIVE CABLE

19

UNCLASSIFIED
RNL-LR-DWG A
F\’EACTOF\’ q:_ ORNL-LR-D 60845

f .\ ‘Q§§\ GUIDE TUBE
N \ - GUIE BAR
FuETLYEEA“NLNEU /'\\ \i///
. \‘ \\\\\

el 7t

AN

4 GRAPHITE lRRADlATION V/\\\“’/////

SAMPLES ( 7/g-in.DIA) > J

g in. TYPICAL

//

-t 2.00in —»=

Fig. 1.14. MSRE Contro!l Rod Arrangement.

Since ultimate reactor safety will not be vested in the control rods, the
number of nuclear penetrations has been reduced from three to one. This re-
maining penetration, a large water-filled tube, will house two fission chambers
and two neutron-sensitive ion chambers. These are sufficient to cover the full
range of reactor operation and will serve as control system input sensors.

A preliminary list of those situations calling for either automatic shut-
down by draining or for contreol rod insertion has been prepared.

1.11 DESIGN STATUS OF BUILDING AND SITE

All major design on the building has been completed, and drawings and
specifications were sent out to prospective bidders in July. The packages
designated A and B (construction in reactor, radiator, and containment areas,
drain-tank cells and maintenance control room) will be bid on September 6, 1961.
These packages, now combined and currently referred to as Major Building Modi-
fication, have had three addenda issued. The addenda were made in time to be
included in the September & bid date.

The remaining site modification design, consisting of Minor Building Modi-
fication (office ventilation, service-air compressors, etc.) and Exterior Site
(cooling tower, filter house, charcoal filter columns, etc.) will be completed
in October.

1.12 REACTOR PROCUREMENT AND INSTALLATTION

1.12.,1 Demclition and Minor Alteration Work to Building 7503

The demclition and minor alteration work performed on a cost-plus-fixed-
fee contract by the H. K. Ferguson Company was completed on schedule in June.
The work consisted of stripping structural steel from the 24-ft-diam containment
20 )

vessel, partial excavation of the drain-tank cell, rehabilitation of the storage
cells, stripping of obsolete conduit and piping which interfered with MSRE
installation, excavation for the nuclear instrumentation tube, removal of a
section of the penthouse wall, construction of an outside stairway for an
emergency exit, painting of offices, control room and hallways, and general
cleanup.

1.12.2 Major Modifications to Building 7503

Drawings and specifications for the first bid package of work, Modifiication
of Building 7503, have been sent to prospective bidders, and bids for performing
this work are to be opened September 6. This package includes all work associated
with the modification of the 2L4-ft-diam reactor containment vessel, the radiator
cell, drain-tank cells, the secondary containment walls, and the remote mainte-
nance control room. Involved are structural, piping, and electrical work, &all
of which will be accomplished by lump-sum contract.

Other site-preparation work will be advertised for lump-sum contracts in
October, when design work is scheduled to be completed. All site-modification
work is scheduled for completion prior to June 30, 1962.

1.12.% Procurement of Materials

Contracts were awarded on a formal bid basis for the fabrication of INOR-8
plate, sheet, rod, and weld wire for the MSRE. Promised delivery date for this
material is September 15.

Bids received for fabricating pipe, tubing, and fittings of INOR-8 are
being reviewed and evaluated. It is expected that contracts will be awarded to
successful bidders in August.

Graphite manufacturers have been invited toc submit bids for furnishing the -
MSRE moderator graphite, completely machined to specifications and tolerances.
Bid closing date is August 14, and the purchase contract is expected to be
awarded tc the successful bidder by September.

1.12.4 Procurement of Components

A request for bids on a package of MSRE major components has been sent to
prospective fabricators. This package consists of the reactor vessel, including
the internal support structure, the radiator, the primary heat exchanger, and
the fill-and-drain tanks for both the fuel and coolant systems. All INOR-8
mgterial will be furnished by the Project.

A pre-bid conference with fabricators interested in bidding for this work
is scheduled for August 29. Because INOR-8 is a new alloy it is expected that
scme fabricators may not be familiar with its working properties, therefore
sample material produced to MSRE specifications will be furnished to each
qualified prospective fabricator requesting it in order that he may become
familiar with the materigl before bidding. Closing date for bids to fabricate
these components is October 6. .
21

1.15 RBEACTOR INSTRUMENTATION AND CONTROLS DESIGN

1.15.1 TInstrument Application Diagrams

During the past report period the major effort was directed toward the
completion of an acceptable set of instrument application flow diagrams. A
total of 15 drawings is scheduled. Of this number, the 13 listed below were
issued for comment at least once.

D-AA-B-40500 Fuel-Salt Circuit

D-AA-B-L0501 Coolant-Salt System

D-AA-B-L0502 Fuel, Flush, and Drain-Tank System
D-AA-B-40503 Cover-Gas System

D-AA-B-4050L Fuel-Salt-Pump Lubricating 0il System
D-AA-B-L0505 Fuel-Sampler-Enricher System
D-AA-B-40506 Liquid-Waste System

D-AA-B-L0508 Coolant-Salt-Pump Lubricating 0il System
D-AA-B-L0509 Water System

D-AA-B-L40510 Off-Gas System

D-AA-B-40513% Fuel Fill and Transfer System
D-AA-B-40514 Instrument-Air and Service-Air Systems
D-AA-B-L0515 Containment Air

Revisions are now being made to all these drawings, bringing them up to date with
current thought. It is estimated that this basic part of the instrumentation
ef'fort is 85% ccmplete. A tabulation of all instruments shown on the flow dia-
grams, giving identifying numbers, location, function in the process, and a brief
description is also being developed along with the application drawings. This
tabulation is approximately 60% complete. Twelve thermocouple-location drawings
were issued "For Approval," but some additional revisions will be required before
final approval is obtained. A tabulation suggesting the method of readout for
all thermocouples was also completed.

1.1%.2 ZElectrical Control Circuitry

Control-circuit design is in a preliminary stage. The first series of
conferenceg to establish control-circuit requirements was completed. A function-
al listing” of the most important interlocks was prepared. This listing is
preliminary and many control 1limits have not yet been established.

1.15.3 Layout

Layout of the Instrumentation and Controls (I and C) system is proceeding
as layout of the building and equipment become firm and as instrumentation
requirements become known. Two major instrument areas have been designated. A
main control area will be located at elevation £52 in the northeast corner of
the building, and a transmitter room will be located on the 840 level, adjacent
to the reactor. Some instrumentation will be located on auxiliary panels outside
these areas. However, an effort i1s being made to centralize instrumentation,
and field panels will be used only where necessary or where the nature of the
operation dictates that controls and instruments be located in the field.

The T and C system layout is being designed to permit all routine operations
to be performed in the main control-room area. A proposed arrangement of the
main control area is shown in Fig. 1.15. All thermocouple leads will be brought
through a patch panel located in the auxiliary area adjacent to the main control
rocm. Safety-control-circuit relays will be mounted in a special cabinet also
22

located in the auxiliary control area. Data-logging equipment will be located
in a room adjacent to the main control area.

UNCLASSIFIED
ORNL-LR-DWG 63244
1 OBSERVATION DECK

- A f'/\
S NEW GLASS WA MAIN CONTROL THERMOCOQUPLE ‘
@ LL BOARD . PATCH PANEL —- . o
~ i
CONSOLE@ t[2]3]a] |

i
FRONT T
AUXILIARY CONTROL
ROOM

INSTRUMENT SHOP
AREA 18 x17 ft (306 112)

DATA AND SUPERVISORY
ROCOM AREA 18 x14 ft

(252 f12) REL AY

apEsobnaen

NEW GLASS WALL = -- 18 % 49 ft (BA2 £12)

AUXILIARY CONTROL BOARDS
I e

ELEVATION 852 ft = = ]
C 2 4 6 8 10
SCALE IN FEET

Fig. 1.15. Proposed Operational Area Layout.

A1l electrical and pneumatic signal lines originating within the contained
areas will be brought out through specially designed lesktight penetrations in
the wall of the containment vessel. .Thermocouple and electrical signal leads
will be terminated in pressurized junction boxes located in a tunnel adjacent
to the reactor. Standard control cable and thermocouple lead wire will be run
in open trays from the tunnel to the main and auxiliary control areas. Pneu-
matic signal lines will be brought through a tunnel to the transmitter room
adjacent to the reactor. Some of these lines will terminate at equipment such
as weigh-system control panels, solencid valves, and pneumatic receivers mounted
on auxiliary panels in this area. Other lines will continue through the trans-
mitter room to the main control area. Detail layout of equipment in these areas
will begin soon.

A survey of requirements for pneumatic tubing, thermocouple cable, and
electrical signal cable for the contained areas was made in order to establish
the number and size of penetrations needed. A preliminary drawing showing
proposed routing of cable, trays, conduits, and tubing throughout the 7503 area
was prepared and issued "For Comment." Existing facilities will be utilized
wherever possible.

1.13.4 Main Contrcl Board

Figure 1.16 shows a front elevation layout of the proposed MSRE control
panel. This design is based on the principle that only that instrumentation
necessary for operation of the reactor system will be located on the main con-
trol board (MCB). Other information will be read out either on a data logger
or on instruments located in an auxiliary area adjacent to the main control
room. Those controls and instruments associated with routine adjustment of
reactor power will be located on a console. Although the layout shown is
incomplete with respect to details and the number of instruments, comments
indicate that it is acceptable from the standpoint of general arrangement,
method of presentation, and systems shown.
a4 —

s g—

1

T4 -

[ HHEEEE ]

IEEGGEEEN

HELEEE ||| M

23

UNCLASSIFIED

ORNL-LR-DWG 63245

[ BEHETE |

EEEEE

[ REFEDE

[ EREENE |

EEEER

| BEROE |

JUMPER BHOARD

JUMPELR BOARD

JUMPER BOARD

AUKILIARY PUNMP &
COMPREDSOR
P UH-ESUTTOMNS

RAULTIPONIT
TEMPERATURE
(NDICATOR

ADLILIARY PUMP &
COMPRESS,OR
PUSHBUTTOM®S

PAMNEL MYM.CB.

PANEL N% M.CB.2

; i e bi -
L 3N
- 3 i LR FP-A,
Dt (D IprC 3224
&

4

—

YA AT ]

PANEL MN*MC B3I

PANEL NOM.C.B 4 i

KU I POIRIT
TE W APERATURE
CECORDER

| PAEL NTMCBS DANEL N'MCH &

PANLL N*IC BT

PAMEL NEMCEB S

PAMEL NOMCBI

PAMNEL. N MEBG

PAMEL 2 MC By | |

2eto’

Fig. 1.16. Composite Control Board Layout.

24

1.15.5 MSRE Data-Handling Study

A study of the MSRE data-handling requirements is nearing completion. The
objective 1s to determine those requirements and to propose one or more data
systems to meet them.

Studies have indicated that approximately 612 input signals, excluding
those transmitted to field-mounted instruments and indicator lamps, are to be
collected by a data display system. A summary of these inputs is shown below.

Radiation 37
Pressure ) 2l
Flow 17
Level L
Speed 1
Weight 5
Tlectrical (voltage, current, frequency) 7
Temperature 517

Of the 517 thermocouple inputs, 245 are used primarily during the reactor
heating and coocling operation and will be handled by a separate temperature-
scanning system.

All information necessary for reactor operation will be displayed on the
reactor conscle and on main and auxiliary panel boards by conventional recorders
and indicators. An automatic data-handling system would implement this system
by recording data in a convenient form. Using this approach to information dis-
play, a data system should have a capacity to handle 250 to 500 inputs.

The major problem associated with the conventicnal display system is the
long-term storage and retrieval of information. This problem results from the
storage of information on numerous single-point recorder charts and the subsequent
job of retrieving and correlating information from several charts at a common
time base.

Another major disadvantage of the conventional data-display system is the
production on charts of large guantities of static data. This static data
recording is necessitated by the requirement to record transients should they
occur. Using strip charts, the data are continually recorded, even during static
system operation.

A data-handling system would alleviate these problems and provide further
improvement in data handling by:

1. providing high- and/or low-1limit alarm detection for each input and
providing printout of the signal value when an alarm occurs;

2. logging data on a typewriter at periodic intervals or by ocperator demand,
in a format such that interaction between variables can be seen;

%. 7providing for long-term data storage, by recording on magnetic tape in
a format for entry into an external computer;

4., displaying data directly in engineering units;

5. providing calculations necessary to perform some on-line date analysis;
and
25

6. providing flexibility so that the input-scanning sequence and output
logging can be altered merely by a program change.

A1l these features will aid in the collection and handling of data. The
data will be more accurate, and analytical results will be available much
guicker. This will result in reducing manpower requirements for data processing
and in reducing over-all experiment costs by providing information in time to
influence the planning of subsequent experiments.

The complete results of the data study will be contained in the report to
be issued to the MSRE proJect in the latter part of August cr early September.

1.1%3.5 Single-Point Temperature Alarm System

A study is under way to determine the most reliable and least expensive
method of monitoring the operating status of freeze flanges and freeze valves.

Both the flanges and valves use temperature measurements to indicate their
operating status. The flanges are equipped with four thermocouples and four
spares and the valves with three thermocouples and three spares. Low tempera-
ture of the flanges indicates normal operation, and low temperature of the valves
indicates closed position. High temperature of the freeze flanges indicates
possible seal leaskage, and high temperature of the valves indicates valve-open
position. Definite operating temperature ranges are designated for both the
flanges and valves. The operating status of both these components must be known
during the fill and drain and normal reactor operation.

The freeze-flange status will be determined by monitoring two thermocouples,
by single-channel devices. An alarm will be produced when either thermocouple
exceeds a preset high limit. Single-channel monitors are required to produce
the required reliability. In addition to these monitors, one thermocouple will
be logged to provide an operating history.

The freeze-valve operation is determined by monitoring three thermocouples
by single-channel devices and logging one spare thermocouple for operating
history. Two of the single-channel monitors will produce a visual indication
when the temperature reaches either high or low setpoints. The visual indication
will indicate valve position. A visual signal on low temperature indicates valve
closed, and on high temperature, valve open. Control circuits will be utilized
with these monitors to produce audible and visual alarm signals should operating
malfunctions occur.

The third single-channel monitor will be used to provide control interlocks
or permissives during certailn phases of reactor operation. This unit produces
high- or low-limit signals which are used for this purpose.

One of the systems being considered for this application utilizes bistable
magnetic amplifiers equipped with high/low alarm detectors. The units are com-
pletely solid-state devices. A block diagram of the system is shown in Figs. 1.17
and 1.18.

The control module furnishes power to operate up to 22 slarm modules. It
contains a circuit common to all alarm modules to provide central indication of
high- or low~alarm condition when any module in the system is triggered.

The input signal from a thermocouple is fed to an isolated winding on the
alarm-module magnetic amplifier. The magnetic amplifier is adjusted so that an

26

UNCLASSIFIED
ORNL-LR-DWG 63246

LIMIT REMOTE QUTPUT
SET
MAGNETIC

AMPLIFIER
HIGH-LIMIT REDOH[GH-LIMiT
THERMOCOUPLE DETECTOR INDICATOR LAMP
< 113
_j SET
CALIBRATE
SIGNAL
LOW-LIMIT AMBERO LOW-LIMIT
DETECTOR INDICATOR LAMP

od 1]

TEST METER

REMOTE QUTPUT

Fig. 1.17. Block Diagram of Alarm Module ond Single-Channel Temperature
Alarm System.

UNCLASSIFIED
ORNL-LR-DWG 63247

[ METER
METER
LINES FROM D MASTER
ALARM MODULES RED LIGHTS o L RED
- OR
AMBER LIGHTS
»| CATES MASTER
AMBER
CARRIER POWER CARRIER
-

TO MODULES OSCILLATOR

AND POWER

AMPLIFIER

DC ' ‘
POWER
SUPPLY POWER SPEAKER
FAILURE
DETECTOR w1  LIGHT
] \
CALIBRATOR

i l BATTERY

Fig. 1.18. Block Diagram of Control Module and Single-Channel Temperature
Alarm System.

input in its operating range produces a differential output into the limit-
detecting circuits. If this cutput exceeds the levels set by the high- or low-
limit set controls, the magnetic amplifier is triggered into a "high" or "low"
operating condition and drives the proper alarm light or relay. The magnetic
amplifier is reset by the "reset" button, but if the off-normal condition still
exists the magnetic amplifier will revert to its alarm condition after reset.

27

The system is designed for fail-safe operation and is equipped with power-
failure-detection circuits. The design of the alarm modules is such that mal-
functions can be detected by the alarm circuits.

To provide the best reliability, the system would be operated with two
control modules, each module supplying one alarm module at each freeze flange

and valve. In this manner the failure of one power supply would not cause the
logs of the complete information or control action.

REFERENCES

1. E. S. Bettis, MSRE Component Design Report, MSR-01-0( (June 20, 1961).

2. MSRP Prog. Rep. Feb. 28, 1961, ORNL-3122, Table 1.%, p O.

%, Ibid., p 16.

4L, S. E. Beall, W. L. Breazeale, and B. W. Kinyon, MSRE Preliminary Hazards
Report, ORNL CF-61-2-46 (Feb. 28, 1901).

5. Addendum to MSRE Preliminary Hazards Report, ORNL CP-01-2-46, (Aug. 14,
1961), p 29.

2. COMPONENT DEVELOPMENT

2.1 FREEZE-FLANGE DEVELOPMENT

2.1.1 MSRE 5-in. Ilanges

The freeze flanges for use in the primary and secondary salt systems are now
sized for 5-in. sched-40 pipe connections. The over-all dimensions, with the ex-
ception of the pipe connection size, are the same as those previously reported.l
The gas seal was changed from a buffered double-conical-gasket type to a buffered
ring-joint type because of the following considerations:

1. The conical-gasket type of seal is believed to be susceptible to
gasket damage if excessive angular misalignment exists during
assembly. It is believed that the ring-joint type of seal is
less sensitive to misalignment as the mating flanges are belng
brought together.

2. Additional confidence in the ability of ring-joint seals to meet
the leaktightness requirements has been gained through the
freeze-flange thermal cycling testing described in Sec. Z2.1.2
and the remote-disconnect testing described in another I'eport.2
The revised flange design is shown in Fig. Z2.1.

Two sets of clamps were purchased for developmental use with the 5-in.
flanges. However, they were returned to the vendor for heat treatment, which
nhad been overlooked during fabrication. Two sets of rough-machined INOR-8
flange forgings were ordered for development use. Four nickel ring-joint gaskets
are also on order. These flanges and gaskets will be used for testing the full-
size gas seals and for thermal cycling to check the dimensional stability and
thermal-fatigue life of the flanges.

A system for installing and removing the clamps was designed and is described
in Sec. Z.11.

A study is under way to determine means of obtaining consistent over-all
thicknesses of the flange-gasket assemblies. Methods of closely controlling the
effective height of the ring gasket and the effective depth of the flange grooves
will be reguired. This requirement is imposed by the steep load-deflection
characteristic of the spring clamp.

2.1.2 Freeze-Flange-Seal Test Facility
The equipment for testing freeze-flange seals was completed and placed in
operation. The facility, with a 3-1/2-in. and a 6-in. freeze flange installed,
is shown in Fig. 2.2.
In heating the 3—1/2-in. flange, it was necessary to install a hollow copper

heat collector that eacily clears the inside of the pipe but closely fits the
bore of the flange. This collector receives heat from the entire length of the

28

29

UNCLASSIFIED
ORNL-LR-DWG 63248

¢ ~ \
FLANGE / fin. —~=0,030-in. \

CLAMP ———» * GAP WIDTH
\ l
{in.
(TYP) | 2-in. TAPER
/ 0400in./in. .
5 ey, (TYP) ks
BUFFER / \ %
CONNECTION {a=in. TAPER
0.050 in./in.
MODIFIED R-68 (TYP)
RING GASKET —] .
_ Y
g

L !

t%-in-R (TYP)

9-in. SCHED-40C PIPE SIZE

Fig. 2.1. Five-inch Molten-Salt Freeze Flange.

silicon carbide heater and conducts a large part of the heat to the flange. By
appropriate adjustment of the heater power and suitable insulation thickness on
the pipe stubs, the desired temperature distribution was obtained.

The 3-1/2-in. flange was assembled, using a gold~plated Inconel gasket and
a resilient clamp, and each of the two bolts was torqued to 300 ft-1b. The
gasket was lubricated with a coating of graphite-alcohol suspension before in-
stallation. During and after torqueing, each clamp block was rapped sharply on
each side several times with a small hammer, with noticeable immediate reduction
in leak rate. The flange was then cycled 21 times between 200 and 1300°F.
Results of leak-rate readings made during the cycling are listed in Table 2.1.
No measurable leakage occurred after the fifth cycle.

Fig. 2.2. Freeze Flange Seal Test Facility Showing 6-in. Flange (left) and 3‘/2-in. Flange.

SRR UnCLASSIFIED
. PHOTO 36959
- L 9939

5

0€
31

Table 2.1 Helium Leak Rates (std cc/sec) of 3—1/2—in. I'reeze Flange
with Gold-Plated Inconel Ring-Gasket During Cycling
Between 200 and 1300°F

Cycle No. Male Qutside Male Inside Female Outside Female Inside
Initial Makeup NDL* 3.5 x 10'8 NDL 5.6 x 107
B-1 Hot 1 x 10'8 1.5 x 1078 NDL NDL
B-1 Cold NDL NDL NDL NDL
B-2 Hot 7 X 1077 6 x 1077 NDL NDL
B-2 Cold NDL 5 x 1077 NDL NDL
B-3 Hot NDL 2 x 1077 NDL NDL
B-3 Cold 2 x 1077 1.3 x 1070 NDL NDL
B-4 Hot NDL b x 1077 NDL NDL
B-4 Cold NDL 6 x 1077 NDL NDL
B-5 Hot NDL b x 1077 NDL NDL
B-5 Cold NDL NDL DL NDL
B-6 Hot NDL NDL DL NDL
B-11 Hot NDL NDL NDL NDL
B-15 Hot NDL NDL NDL NDL
B-19 Cold NDL NDL NDL NDL
B-21 Hot NDL NDL NDL NDL
B-21 Cold NDL NDL NDL NDL
Disassembled, then Reassembled with Same Gasket

Initial Makeup 1.4 x 1077 NDL NDL NDL
B-22 Hot 5 x 107 NDL NDL NDL
B-22 Cold 1.5 x 1077 NDL DL NDL
B-26 Hot NDL NDL NDL NDL
B-26 Cold 1.8 x 1077 NDL NDL NDL

*No detectable leak.

The 3—1/2-in. flange was then disassembled, the gasket was again coated with
the graphite-alcohol suspension, and the assembly was remade with hammering. As
shown in Table Z.1, all seals were nearly perfect during the hot part of thermal
cycling, but one of the seals persisted in leaking when cold and showed no im-
provement after five thermal cycles.

A solid-nickel ring gasket has been installed and will undergo similar test-
ing. Additional nickel rings, two of which are grooved on each "nose" to in-
crease resiliency, are on hand for testing. It is expected that the nickel
rings will form tight seals without requiring the use of gold plate.

32

The 6-in. flange-pair in the seal-test facility is essentially the
flange design except that the bore is slightly greater and Inconel is used as
the flange material stead of INOR- Efforts to heat the bore of this flange
to 1400°F with the silicon carbide bar heater at the center line and with an
additional Nichrome wire heater at the bore were uccessful. Resistance heaters
were installed on the outside of the pipe stubs adjacent to the flange, and an
induction coil was placed in the bore of the flange to supply the required heat
1 distribution. Temperature distributions along the gap between the flanges and

on the outside surface will be determined. It is intended, then, to thermal cycle
the flanges and check their dimensional stability.

2.2 FREEZE VALVES

Two Calrod-heated valves,3 of the type proposed for MSRE operutional valves,
were operated in the RE engineering-test loop (Sec. 2.10). A photograph of
one of the valves is own in Fig. 2.3. eir arrangement in the loop was shown
previously (see Fi 5 in ORNL- ). The valves are located in the
test loop 5o that salt may be diverted to either of two drain tank; cservoirs
are placed on the vertical risers to ensure thot the freeze zones will not run
dry following a drain.

Helium was permitted to flow through one valve and reservoir at the rate of
2.2 cfm to check the operation of the reservoir after a system dump. The other
valve was closed during this operation. The gas pressure was then equalized

7 UNCLASSIFIED
PHOTO " 36713

PIPE HEATER

FREEZE ZONE

I

Fig. 2.3. Freeze Valve.

33

between the drain tank and circulating system, the valve was frozen, and x-ray
photographs were taken of the valve assembly. The photographs indicated that
sufficient salt remained in the valve to make a seal. The valve proved to be
gas tight when refrozen. The valve to the flush tank was operated with cooclant
salt through 40 cycles without difficulty. The melt time varied from about

3 min at 1.5 to 2 kw of input to about 15 min when the only heat provided was
from losses from the loop.

2.3 HEATER TESTS

2.3.1 Pipe Heaters

A prototype, clamshell heater-insulation unit for the MSRE piping was built
and tested (see pp 28 and 29 in ORNL-3122). When the pipe wall was maintained
at the design temperature of 12000F, there was a heat loss of 550 w from a 1-f%
length of 4-in. pipe. The average surface temperature of the insulation was
2LOOF except at the latch closure, where the temperature was 450C0F., Heaters of
this type were eliminated from the MSRE design because of their high heat loss
and the difficulty in their removal and installation with remote maintenance
tools.

A heater of the type shown in Fig. 1.9 is being buillt for testing.

2.3.2 Core Heaters

The reactor vessel is to be surrounded by a furnace containing 63 hairpin
resistance heaters.u The heaters are nested in nine banks of 2-in. stainless
steel tubes; each bank of tubes is replaceable separately. An Inconel-sheet
reflector and ceramic insulation are provided to minimize heat losses from the
furnace.

A test mockup of a full-size bank of heaters was set up in order to evaluate
heater and furnace performance and replaceability. Figure 2.4 shows a photo-
graph of one of the test Inconel heaters, which has a resistance of 0.2 ohm.

Both Inconel and stainless steel heaters were used. TFigures 2.5 and 2.6 show the
test facility which contains nine heaters, containment tubes, reflector high-
temperature insulation, and a heat sink replacing the core vessel.

The system was operated for 3000 hr without difficulty. Visual inspection
was made of the heaters after 576 hr and again at 3000 hr. There was little
change in the appearance of the equipment. The entire inside of the furnace,
heaters included, had oxidized to a soft black, but there was no indication of
severe attack in the form of flaking. The Inconel heaters appeared in better
condition than the stainless steel heaters. The stainless steel containment
tubes had a brassy-to-black mottled appearance, compared with a uniform black on
the Inconel. The reflector and heat sink faces were also uniformly blackened.

The organic binder in the Cerafelt (Johns-Manville Co.) insulation had been
burned out to a depth of 1—1/2 to 2 in. at the hot face. The insulation had
little remaining structural strength, although the insulating qualities were un-
affected. Shrinkage of the insulating material was 1.5 to 2%. Shrinkage of the
insulation on a system of this size (9 ft high and 4 ft wide) causes serious
heat losses unless some provision is made to avold air leakage. Leakage on the
test stand was stopped by plugging the cracks with Fiberfrax, a loose-wool high-
temperature lnsulation (product of Carborundum Company ).

1"

ft

34

UNCLASSIFIED
PHOTO 36550

‘/%—in, INCONEL ROD ELECTRICAL
CONNECTOR

INSULATION FOR THERMAL
NEUTRON COVER SHIELD
PENETRATION

[
E 'fi,/%—in. ROD TO ¥g-in.- TUBE
TRANSITION
\%~in.-on, 0.035-in.-WALL

INCONEL TUBE HAIRPIN

HEATING ELEMENT

' INSULATORS

Fig. 2.4. Core Heater Element.
35

UNCLASSIFIED
ORNL-LR—DWG 63249

0.031-in. INCONEL REFLECTOR— 4|

f"fl 7in.
:

|1
TUBE SHEET — 777

INCONEL TUBE HAIRPIN HEATER
IE/ (3/g-in.— 0D, 0.035—in.—~WALL)
- «—HEAT SINK (0. 425 —in- INCONEL)

b\

— COOLING TUBES

4-in, CERAFELT INSULATION
(JOHNS-MANVILLE CO) — ™

304L SS CONTAINMENT TUBE
(2-in.- 0D, 0.0GS—in.—WALL)/U:U:
3in——» }<—

INSULATION (HI-TEMP)

REFLECTOR CERAFELT INSULATION

HEAT SINK
AND COOLING TUBES

CONTAINMENT
TUBE SHEET

INSUL ATION ¥ _INSULATION

INSULATION

T0P

Fig. 2.5. Profile of Core Heater Test, a Full-Scale Mockup of One-Seventh of the Core Heaters.
UNCLASSIFIED
PHOTO 36551

Fig. 2.6. Core-Heater Test Assembly.
37

Tests of commercial high-temperature insulations are in progress; at least
one of the samples does not shrink at operating temperature. [The Fiberfrax
ceramic material without organic binder (2300°F)].

When the experimental apparatus was designed, the total heat loss was esti-
mated to be 5 kw with the heat-sink at 1000°F. Heat losses caused by insulation
shrinkage, warping of the test stand, etc., made it necessary initially to in-
crease the input power to 14.6 kw to achieve 1000°F operation of the heat sink.
After the air leaks were stopped, the average input power was 5.1 kw for 1000°F
at the heat sink, with no heat being removed by the coolant tubes in the sink.
Runs were made alsc with various rates of heat withdrawal, but these data have
not yet been analyzed. Preliminary estimates indicate 15 kw can be supplied with
the heaters operating at a maximum temperature of 1400°F and the heat sink at

975°F.

2.4 DRAIN TANK COOLERS

The system for removing after-heat from each MSRE drain tank has 32 thimbles
projecting intc the molten salt. Inside each 1.5-in. thimble is a l-in. bayonet
boiler tube in which steam is generated at low pressure. There is air in the gap
between the thimble and the bayonet tube.

Three full-scale thimbles (Fig. 2.7) were installed into a tank to be filled
with a mixture of carbonate salts for testing. Preliminary tests without salt
are in progress to obtain the system heat losses and the radiant-heat transfer
coefficients in the cooling tubes. After salt is added, the cooling capacity of
each tube will be determined, control of the steaming rate will be investigated,
and the ability of the system to withstand thermal shocks will be tested.

2.5 SAMPLER-ENRICHER DEVELOPMENT
2.5.1 Sampler-Enricher Concept

A schematic layout of the MSRKRE sampler-enricher system for the primary loop
is shown in Fig. 2.8. The components and operating techniques are similar to
those reported previously.5;6 When the location of the MSRE primary pump Wwas
changed, the angle of inclination of the transfer tube which connects the pump
bowl with the dry box was increased from 45° to 54-1/29; this aids in inserting
the capsule. The location of the dry box was also changed from the south side
of the reactor pit to the east side.

2.5.2 Sclder-Freeze Valve

In order to repair the sampler-enricher system components located outside
the reactor cell, a reliable maintenance valve must be provided. A solder freeze

valve was being considered because of its potential reliability. In initial tests,

a 1l-in. valve of this type appeared to perform as expected, with an acceptable
leak rate of < 108 ce of helium per second.! A full-sized l-l/z-in. single-seal
INOR-E valve using the same solder was then constructed and thermal cycled 22
times. The valve was at the operating temperature of lEOOOF for a total of

29.5 hr. Three cycles failed completely to seal, and the remaining cycles were
not consistent in the measured leak rates. This apparent unreliability has not
been explained, but examination is continuing.
UNCLASSIFIED
PHOTO 37231

STEAM CHEST

N
.
-
1-in. COOLING BAYONETS
(PARTLY WITHDRAWN)
N\
=
\ N\
N\ )
| N
B ~
.

1Yo ~in. CONTAINMENT
THIMBLES

Fig. 2.7. Drain-Tank Cooler Test Assembly.
39

ORNL-LR-Ivg. 55006

CONTAINMENT/

VESSEL SHELL

Unclassified
PLUG
EXHAUS] HOOD \P” TRANSPORT CASK
$$0
AREA 4 3
TO YACUUM PUMP | /
: (S stuie TRANSPORT CONTAINER
wr i' QEMOVAL VA\LV/\E
(ByH ACCESS DoO
CELLILLUMINATOR
‘ B WELDED ACCESS
BI(R ] 7 2 A / N
e b ;“e.l —
ZINT eAPSULE ,
B L% £ pepiscope pocess Br s
v T - |
T0 vacuum -F’_liMP [INAARRRRT fi@ AREA]  T—LATCH
S 1c L — CAPSULE
p =,
MANIPULETO o SAMPLE TRANSPORT — D=
% AREA 3 CONTAINER 70 VACUUM
- Jic—a [M]) ['SEAL PUMP
To Watuum puMp =—=p2/ =
AREA 2¢
DISCONNECT FlTANGE ACCESS Dol
[ Yo
OPERATIONAL VALVE —/WANGE
LANGE FOR REMOVAL rragr
EONTAINII:!ENT SHELL or\_ r'
[ AREA 7b MAINEENANCE
|"’"1 AQEA Ib VALY
—r 3 AC
il proéfu?s
HEATER A ELEV. 852
| BLDG, 7503
A A A AranAi ey A
" . Pttty g?;{ Ll
INDICATES BUFFER ZONE
DISCONNECT
FLANGES
ouTeR STEEL sHELL
EXPANSION TR STEEL S
JOINT
LATCH STOP
T CAPSULE GUIDE
PUMP BOWL

Fig. 2.8. Schematic Layout of MSRE Sampler-Enricher System.

40

For reactor use, a buffered double-sealing valve would be required. DBecause
of the apparent unreliability noted above and the complications produced by the
gas controls necessary to make the double-solder seal, an alternative design is
being prepared. This uses a buffered double-sealing gate valve.

2.5.3 Sample Capsule

In order to prevent oxygen contamination of the sample being isolated and of
the circulating fuel, the oxide film on the copper sample capsule must be removed.
A pretreatment consisting of 4 hr of hydrogen firing at 14759F removed the oxide
film from the capsule.

On the basis that molten fluoride salt does not wet the surface of oxide-
free copper metal, an attempt was made to prepare a re-usable capsule from which
a single slug of salt would be removed. However, the salt adheres to the metal
surface tightly enough to prevent its removal without damage to the capsule.

A capsule was then designed from which the salt can be completely removed
by crushing the capsule. The salt is then removed in the form of a coarse powder.
Capsules of two diameters (1/2 in. and 3/& in.) have been tested. Preliminary
crushing tests in a hot cell indicated that the salt would crush out of the
larger capsule more easily. The upper portion of the smaller one tended to col-
lapse, trapping salt in the lower part.

2.5.4 Sample-Transport Container and Removal Seal

A transport container was designed which will seal the hydrogen-fired cap-
sule 1in a dry, inert atmosphere for shipment to the sampler dry box and for trans-
porting the salt-filled capsule to the hot cells. The transport container is
inserted into the dry box through a buffered double-sealing section located
above the removal valve (see Fig. 2.8). The seal provides a purge zone to pre-
vent the contamination of the dry box with oxygen and the release of fission-
product gases from the dry box to the atmosphere. The leak rate of the container
and the seal section will be tested after each installation has been completed.

2.5.5 Sampler-Enricher Design

Detail design of the various components of the sampler-enricher system is in
progress. Preliminary design of the transfer tube, the maintenance valve, and
the operational valve are complete. A preliminary instrumentation flow sheet 1is
complete.

2.6 MSRE CORE DEVELOPMENT

Additional tests were conducted in the cne-fifth-scale MSRE core model.8
A report was issued summarizing the results of this program.

2.6.1 Heat Transfer Coefficients in Lower Head
of the Reactor Vessel

Heat transfer coefficients were determined experimentally in the lower head
of the one-fifth-scale core model at the reactor center line. Figure 2.9 is a
plot of these coefficients predicted for the MSRKRE as a function of the reactor
flow rate. The lines in Fig. 2.9 represent the theoretical slopes for laminar
and turbulent heat transfer. Irom this it can be seen that the character of
flow in the lower head changes from turbulent at the rated flow rate to laminar
at a reduced value of the flow rate.

4]

UNCLASSIFIED
CRNL—LR— DWG 63250

—
&
S 600 1 | T T T T T
h SLOPE = 0.8 FOR TURBULENT HEAT TRANSFER
: D
[
o« ~ 400 o
el
(£ 'GL a /é/
- B < o
29 L
fran) (o]
T, 7
W £ 200 e =7 S
jm}
a a—] o V/ . |
5 /O |
O o
w
x
n

’<\SLOPE = ’/3 FOR LAMINAR HEAT TRANSFER
I

100 200 500 1000 2000 3000

REACTOR FUEL FLOW RATE (gpm)

100

Fig. 2.9. Predicted Wall-Heat Transfer Coefficients at the Center of the MSRE
Reactor Vessel Head, Based on Measurements in the One-Fifth-Scale Model.

2.6.2 Tuel-Age Measurements in the Lower Head
of the Reactor Vessel

Fuel-age measurements were determined throughout the entire lower head of
the one-fifth-scale model; however, most of these measurements were made with a
reactor configuration which has since been modified slightly. From these
ecarlier measurements it was concluded that the most critical region, that is to
say, the region with the highest combirnation of fuel age, power dengity, and
importance was at the reactor center line.

For the present core configuration, the fuel ages were redetermined along
the center line. Figure 2.10 is a plot of these age measurements as a function
of the distance from the wall and with parameters of reactor flow rate. It i1s
not apparent why the fuel age is at a maximum at about 2 in. from the wall. It
may indicate a zone of recirculation of the fuel. Another possibility is that
the mere existence of the probe in the model has changed the flow characteristics.
At any rate, the point of greatest interest is that next to the wall. Figure 2.1l
is a cross-plot of the reciprocal of the flow rate vs the fuel age at the wall.
One would expect that the fuel age is related to the flow rate as follows:

constant {units of volume)
volumetric flow rate

fuel age =

Figure 2.11 secems to substantiate this except at low flow rates, where the age
increases less rapidly than would be predicted. 1f, however, the character of
flow has changed from turbulent to laminar when the flow rate is reduced as pro-
posed above, then it seems reasonable to expect the constant in the above equation
to change. This could also explain the radical change of shape of the curve of
lowest flow rate in Fig. 2.10.

Based on heat transfer coefficients, fuel ages and power-density distributions
in the fuel and vessel wall, the outside-wall temperature of the lower head of
the reactor vessel at the center line position has been estimated to be 1215°F,
which is 40°F above the iniet temperature of the reactor. This wall temperature
rises to 1226°F at 50% of design flow rate. This estimate is with the reactor
operating at design conditions and with no solids deposited at the wall.

42

UNCLASSIFIED
ORNL—LR-DWG 63251

140 ‘
FLOW RATE OF
REACTOR FUEL (gpm)
120 A 322
~

100 LAMINAR FLOW >
\

80‘ /U,
|

‘\\\fi 561

60

FUEL AGE AT POSITION INDICATED (sec)

40

o ~7Y 1120 —
° TURBULENT FLOW T—4
20 pD—m 1400
0
0 1 2 3 4 5 (3]
DISTANCE FROM WALL IN LOWER HEAD OF REACTOR VESSEL (in.)
Fig. 2.10. Fuel Age Predicted at the Center Line of the Lower Head of the -

Reactor Vessel.

It should be noted that after the above measurements were made, the design
of the reactor drain line was modified. The drain line now projects above the
wall at the center line of the head, and some minor changes will probably be pro-
duced in the flow distributions reported above.

2.6.3 Studies on the Disposition of Fine Particles
in the Lower Head of the Reactor Vessel

Experiments were conducted with the one-fifth-scale model to determine the
disposition of particulate matter in the core. The particles could result from
at least two sources: corrosion of the contaimment material and oxidaticn of the
fuel. For these tests, molybdenum powder with an average diameter of about 5 p
was used. This simulated 25-p particles with a specific gravity of about 10 in
the fuel salt. 1In one typical run, 70 g of molybdenum powder was added to the
lower head of the model, and the model was operated at the design flow rate for
a total of 4O hr. After the first 21 hr the model was disassembled and it was
found that about two-thirds of the powder had been swept out and that the rest -

UNCLASSIFIED
ORNL—LR—DWG 63252

FLOW RATE (gpm)

1400 1120 842 561 322
80 .
T ]
A
70 -
A
c0 !
|
|
!
3 50| -
i
—J
-
<X .
= |
= 40 | e -
L
1]
a
-
Lo
T 30 : i
20 —— - -
!
10 | - .
|
o . L
0 0.001 0.002 0.003
RECIPROCAL OF THE FLOW RATE (gpm)~*
Fig. 2.11. Fuel Age at the Wall of the Lower Head of the Reactor Vessel af

VYarious Flow Rates.

had piled up in the center of the lower head. Again, after a total of 40 hr,

about two-thirds of the remaining powder had been swept out. Indications were
that these particular particles would have eventually been completely removed

from the lower head of the model.

Tt should be noted however that one major difference between the model and
the reactor is that the model is operated in a once-through system, and as the
solids are swept out of the lower head they are carried to the drain. In the
reactor the solids remain in the system and may settle out again. If a drain
line were simulated in the model it would probably be a worse case than that
tested because it would provide a stagnant corner in which solids might accumu-
late. These same particles were tested at 60% of the design flow rate, and it
appeared that very few if any of the particles were swept out of the lower head.

Tests were also conducted in which 200~ to 500-u particles of specific
gravity 10 were simulated. These particles definitely settled in the lower head,
and there were no indications that they would be swept out even at the design flow
rate.

44

2.6.4 Fluid-Dynamic-Induced Power Oscillations

One proposed: cause for fluid-dynamic-induced power oscillations in the MSRE
core is due to the character of flow in the fuel channels. The Reynolds number
of the fuel in the center most channels is about 3500; therefore, the flow is in
the transition region, a region often thought of as oscillating between laminar
and turbulent flow. ©Such oscillating is a likely cause for power oscillations in
the reactor. An almost full-scale model of four parallel fuel channels was built
and tested in two different ways. ¥First, a visual method was used. With a double
refracting solution of Milling Yellow dye and two mutually perpendicular polarized
plates, it was possible to see clearly the transition between laminar and turbulent
flow. The transition was smooth and continuous, and no significant oscillatory
action was detected. The second method of detecting oscillations was by measuring
the fluid pressure drop along the length of the channels and between the channels.
A manometer that would react significantly to at least 5-cps pressure oscillations
was used for these measurements. The model was operated from Reynolds numbers of
1000 to L4000, and again no oscillations that could be regarded as significant were
detected. If oscillations faster than 5 cps occurred, then at any one time there
would be about 12 oscillations in the same fuel channel, and the effects would
probably cancel.

Another possible source of fluid-dynamic-induced power oscillations is from
short circuiting of the fuel through the lower head. This is possible because
the lower head is basically a cavity, not a channeled system. The "swirl killers"
provide partial but not complete channels. If a fraction of the fuel does short
circuit through the lower head, then a corresponding amount of fuel must reside
there a little longer. This results in the short-circuiting fraction entering the
core at a cooler temperature than the other fraction. Tests were conducted in the
one-fifth-scale model by injecting salt into the lower head at a constant rate and
measuring the oscillations of the fluid conductivity at the outlet. Based on these
tests, the power level in the MSRE should not oscillate on either side of the mean
by more than 0.03% as a result of short circuiting of the fuel in the lower head
of the reactor vessel,

2.6.5 Full-Scale MSRE Core Model

The loop for testing the full-scale core model was completed. The pressure
vessel of the model (Fig. 2.12) was received and is being installed in the loop.
The core internals that is, the core blocks, core support, and core shell are
scheduled to be delivered in September.

2. ( HELIUM PURIFICATION

Laboratory tests were made to determine the efficiency of titanium sponge as
a getter for removing oxygen from a helium stream. The test unit consisted of a
titanium bed, 3 cm in diameter and 25 to 50 cm long. Runs were made at flow
rates up to 4 liters/min and at bed temperatures of 600 to 1200°F. By using an
inlet concentration of 100 ppm of 02 by volume, outlet concentrations of well
below 1 ppm of Op were obtained at 1200CF. Similar results were obtained at
1000°F, provided that the flow rate was maintained at 2 liters/min or less. Effi-
ciency increased with temperature and decreased as the flow rate was increased.
Oxygen analyses of the effluent gas were made by gas chromatography. The results
were somewhat erratic, and confirming tests will be made, using an electrolytic
type of analyzer.

oince the test results were encouraging, a full-scale oxygen removal unit
was designed for testing under the maximum operating conditions of the cover-gas
supply system (250 psig, 10 liters/min).

UNCLASSIFIED
PHOTO 37215

Fig. 2.12. Pressure Vessel of the Full-Scale MSRE Model.
46

2.8 PUMP DEVELOPMENT

'The design and water testing of the fuel pump were completed, and the
thermal analysis was essentially completed. The variable-frequency supply for
the fuel-pump drive motor was received, and the contract was placed for the pump
drive motors. ZFProject review of the pump design was initiated, and fabrication
of the prototype pump and hot-test stand was restarted after satisfactory weld
repairs were made to the INOR-8 volute casting.

2.8.1 MSRE Fuel Pump

Water Tests

Water testinglo of the fuel pump was completed, and a reportll was issued.
Tests were performed to obtaln the pump performance, priming, coastdown and
fountain-flow characteristics, relative effectiveness of the xenon-removal con-
figuration to strip dissolved 002,12 and to prove baffle configurations that con-
trol the behavior of liquid in the pump tank.

Initially, at the lower flows, the pump head was found to be 6 to 10 ft lower
than the data published by the vendor of the impeller and volute. After insertion
of a baffle in the inlet to reduce prerotation and to increase the head at lower
flows, the maximum deviations in head were still negative but had been reduced to
1 ft. at 1300 rpm for flow range 800 - 2000 gpm, 1 1/2 ft. at 1150 rpm for flow
range of 600 - 1600 gpm, and 3 ft. at 860 rpm for flow range of LOO - 1200 gpm
[see p 1& in ORNL CF-61-8-43 (August 1961)].

PK-P Pump Hot Test

The test of an existing PK-P pumpl3 circulating LiF-BeFs-ThF), -UF),
(65-30-4-1 mole %) at 1225°F, 1950 rpm, and 510 gpm was terminated after 5436 hr
of operation as a result of mechanical-electrical failure of the drive motor.
There was no evidence of salt collecting in the off-gas line or spark-plug risers
in the pump tank. Inventory of the lubricant showed a discrepancy between the
measured loss from the reservoir and the measured leakages. The pump-tank offgas
was sampled for possible oil leakage into the fuel system. OSpectrographic analy-
sis of the gas samples did not show any trace of elements which would result from
thermal decomposition of oil in the pump tank. The offgas was run continuously
through a liquid-nitrogen cold trap for 24 hr during each of two tests, with no
evidence of oil collecting in the trap. In a third test, the offgas was passed
through the cold trap for five days without oil collecting in the trap. This
pump test was not restarted because back-diffusion tests will be performed in
the prototype fuel-pump test, and the plugging of gas lines with condensed vapors
of molten salt will be further investigated in a static test.

Back-Diffusion Test

A back-diffusion experimentlu was conducted in the PK~P pump test to deter-
mine the rate of back diffusion of Kr9% from the pump tank up the shaft annulus
and into the catch basin against the downward stream of helium, and to determine
the flow rate of helium required to prevent excessive radiation damage to the oil
leakage in the catch basin. The results of the experiment are shown in Fig. 2.13,
which shows the ratio of the catch-basin to the pump-tank concentrations of Krd5
plotted against the helium sweep-gds flow down the shaft labyrinth. The values

UNCLASSIFIED
ORNL—-LR-DWG 63253

10
N - I
e . )
v\ .
\ \
\ N
\ \
2 o \\
\
1072 \ \‘\
\ \
\ \
\\
sl \\ \
| e
g LN
o \ \ PUMP OFF
g 2f \
\
L‘;J 1072 \\
3 PUMP ON \ \'\ 7
: r—
S 0 ° \
& )
\A
: \ N
. \\
1074 —— . \‘
\ \
5 ! \ \
\ |\
2 YEAN
-5
0 o 1000 2000 3000 4000

SWEEP-GAS FLOW (std cm®/min)

5000

Fig.2.13. Concentration Ratio vs Sweep-Gas Flow. Catch-basin flow, 250 std

cc/min.

48

shown are calculated directly from the experimental data by the following equa-
tions:

Spp *+ 9y

2

Kpp F, (RE-1) o, (1)

where
KPT = krypton concentration in pump tank, curies/cmB,

K., = krypton concentraticn in catch basin, curies/cmS,

RE-1 = count rate of pump tank offgas minus background,
counts/min,

RE-2 = count rate of catch basin offgas minus background,
counts/min,

. 3
cell calibration factor curles/cm
2 counts/min

o

',_I
o
=
ol
r
il

s

QPT = pump tank offgas flow, std. cmg/min)

Q,, = nitrogen dilution flow, std. cm3/min.

The experimental data were taken with a pump-tank pressure of 10 psig. A lower
pressure would probably decrease the concentration ratio due to increased sweep-
gas velocities down the shaft.

The permissible radiation dose rate to the catch-basin oil leakage has been
assumed to be 107 r/day, based on the premise that any incremental volume of oil
will drain from the catch basin within a 2h-hr period and will retain its fluid
properties up to a dosage of 107 rads. The maximum permissible concentration
ratios for various sweep gas flows can be calculated from the experimental data,
the permissible dose rate, and the pump-tank activity level at the particular
sweep gas flow. Maximum permissible concentration ratios of 8.7 x 10-4 and
1.26 x 10-3 were calculated for the pump-tank activity levels resulting from
sweep-gas flow rates of 694 and 3470 standard em3/min. (1000 and 5000 liters/day),
respectively. As shown in Fig. 2.13, the permissible concentration ratio of
8.7 x 10-4 can be obtained on the PK-P pump with a sweep-gas flow of approximately
2000 std. em3/min. (2880 liters/day).

| The back-diffusion equipment is being modified to conduct experiments on the
prototype fuel pump.

Prototype Fuel Pump

rication of the rotary element is about 75% complete, and the pump tank is about
25% complete.

Four fuel-pump volute castings and two fuel-pump impeller castings were

\
|
The design of the prototype fuel pumpl® was completed (see Fig. 2.14). Fab-
received from the founder. Two of the volutes have been weld repaired and a
|
|

UNCLASSIFIED
ORNL-LR-DWG-56043-4A

— —]
SHAFT WATER
COUPLING T COOLED
- A s MOTOR
SHAFT SEAL gfi

LEAK DETECTOR

LUBE OIL IN
LUBE OIL BREATHER
SHAFT SEAL
GAS IN
LUBE OIL OuT N
A
LEAK DETECTOR D
BUBBLER TYPE AN GAS VENT
LEVEL INDICATOR E\§§g‘ !
B A . , XENON STRIPPER
n ‘-
A
BUOYANCY i e vt
LEVEL o 7. ; o i
INDICATOR— . . ; - . .
i 7
—-»(7 = o N i . i

Fig. 2.14. Prototype MSRE Fuel Pump.

third is ready for weld repair, which provides one volute for the prototype fuel
pump and two for the MSRE fuel pump. The leading edges of the vanes of one
impeller casting were built up with weld metal, and a second impeller is satis-
factory as cast. The third impeller will be cast later.

The variable-frequency power supply for the fuel-pump drive motor was deliv-
ered and is being installed for test on the fuel-pump prototype. The contract
for the drive motor was let, and delivery is expected January 1, 1962.

Prototype Fuel-Pump-Test Facility

The fabrication of the prototype test loop16 was completed except for the
pump tank. Assembly is about 50% complete. The instrumentation and controls

S0

panels were completed and installed. Gas and oil lines were run from the control
station to convenient stopping points on the loop, and installation will be resumed
as soon as the pump tank is available.

Reactor Fuel Pump

Drawings of the rotary portion of the reactor fuel pump have been submitted
for project review and approval prior to placing an order for fabrication. Fab-
rication will be started immediately after approval.

The drawings for the stationary element of the pump will be submitted for
project review and approval as soon as the thermal analysis is complete. Fabri-
cation will be initiated following hot test of the prototype fuel pump. Drawing
and vessel calculations of the pump components, comprising primary nuclear contain-
ment, namely, the drive-motor vessel, bearing housing, and pump tank, are being
submitted to Inspection Engineering for review,

Thermal Analysis of MSRE Fuel Pump

As a result of temperature distribution obtained by the GHT code,lr:18 the
design of the volute support cylinder was modified to reduce window width by a
half and to center the windows on the midplane of the volute support cylinder.
The effect of the revised configuration on temperature distribution is shown in
Fig. 2.15; reducing window width in the original position was relatively in-

UNCLASSIFIED
ORNL-LR-DWG 63254

1500

|~ ORIGINAL WINDOW POSITION

1300

1\

/

<
~

(100 # A

ORIGINAL REVISED

o WINDOW WINDOW

w POSITION  POSITION

x |

2 | | |

< 900 ; -+

L REVISED WINDOW POSITION I

S I {1

ul I ]

= SPHERICAL

SHELL

700 N\
500

0 2 4 6 8 10 12 14 16
AXIAL POSITION (in.)

Fig. 2.15. Effect of Window Position on Axial Temperature Distribution.

51

effective. Temperature distributions for fuel and ccolant pumps are shown in
Figs. 2.16 and 2.17 for the following operating conditions:

1. Fuel pump, 10 Mw with 200 acfm (actual cubic feet per minute) air
cooling.

2. IMuel pump, 10 Mw without external cooling.

3. PFuel pump, zero-power with 55 acfm air cooling.

4. TFuel and coolant pumps, zero-power without external cooling.
5. Coolant pump, 10 Mw without external cooling.

UNCLASSIFIED
CRNL-LR—DWG 63255

1700 y
CYLINDER JUNCTION
P — T ——
/ = —
o
/<:; ~
1500 /l_ FUEL PUMP (10 Mw, NO EXTERNAL COOLING) \\ ~
/ N
7/ \\\
£ 1300 / FUEL PUMP (10Mw, 200 cfm AIR COOLING) \
I 0 |
Lul
% /(FUEL. AND COOLANT PUMPS (ZERO POWER, NO EXTERNAL COOLlNG)] x
IE L__ ] — — — — ——
o il -
w L™
L 4100 / i —
'_U;J L et
//——"__—f- ] i e — —
// ,/"”“'
900 , ’_/ COOLANT PUMP (10 Mw, NO EXTERNAL COOLING)
|

/ FUEL PUMP (ZERC POWER, 55 c¢fm AIR COOLING}

| |

0 2 4 6 8 10 12 14 16 i8
MERIDIONAL POSITION {in.)

Fig. 2.16. Meridional Temperature Distribution of Torispherical Shell.

Calculations to determine the effect of varying the flow rate of cooling air
on the surface temperature of the spherical shell have been completed, and the
results are shown in Fig. 2.18. Since the calculations are based on no meridional
heat flow, the values are useful as a guide ornly.

Mechanical and thermal stresses have been calculated, using both an ORACLE
code and an IBM /090 code. The maximum stresses due to internal pressure and

axial mechanical load exist at the suction nozzle of the pump tank and are given
by the following equations, respectively:

S = 122.28 x pressure (psi), (3)
g = 1.766 x axial load (1b). (&)

The thermal stress profile for 10-Mw reactor operation and approximately Z00 acim
of cooling air is shown in Figs. 2.19 and 2.20.
TEMPERATURE (°F)

t400

1200

1000

800

600

400

52

UNCL ASSIF IED

ORNL-LR—DWG 63256

AX1AL POSITION (in.)

VOLUTE
— SPHERE
_— T \\ JUNCT ION
N
N "\// FUEL PUMP (10 Mw, NO EXTERNAL COOLING)
— ———— N FUEL PUMP (10 Mw, 200 cfm AIR COOLING)
— \\ AES —
\\ \ \
N ] \
N \\\
. \ \\\
P N\
COOLANT PUMP (10 Mw, NC EXTERNAL COOL ING) NN Y
NN
| | | N
FUEL PUMP (ZERO POWER, 55c¢fm AIR COOLING)~ - \}\\\\\ TOP FLANGET
! \\
FUEL AND COOLANT PUMP (ZERO POWER, NO EXTERNAL COOLING)/':\-.\.___
1 1 | | I t
0 2 4 6 8 10 12 14 16

Fig. 2.17. Axial Temperature Distribution of Volute Support and Bowl| Neck.

UNCLASSIFIED
ORNL-LR-DWG 63257

1400
U}
o 1200 \\
Ld
v
-
—
<I
x
4 1000 " 10-Mw NORMAL - POWER OPERATION
i \
—
wl
s \
& 800 N\ | —~
2 \
¥ \ ZERC - POWER OPERATION (1300°F
I SALT TEMPERATURE)
=
3 600
= ‘ \R
ZERO -~ POWER OPERATION (1200°F SALT TEMPERATURE)F
0 100 200 300 400 500 600

AIR FLOW {acfm)

Fig. 2.18. Nominal Surface Temperature vs Cooling-Air Flow.

700
(x 10%)

STRESS (psi)

—20

—25

53

UNCLASSIFIED

ORNL—LR-DWG £3258

(x10%)
20 N
\%
15 o(//\
®0
&
10
c
_ 5\0\
= 5 |— oV —
2 Gtk
- ” T —
w) — e Ap— — —
3 /- c\DE o
- ~ \N
» W g
-5
<</ /
-0 fo\Q
>/
-15 6-
/ O’8 = CIRCUMFERENTIAL PRINCIPAL STRESS
-20 / O‘¢ = MERIDIONAL PRINCIPAL STRESS
-25
0 1.87 3.74 5.62 7.4%

MERIDIONAL POSITION (in.)

9.36

Fig.2.19. Thermal Stress Profile — Sphere. Ten-Mw operation; about 200 acfm

of cooling air.

UNCLASSIFIED

ORNL—LR—DWG 63259

-

SPHERE
JUNCIHON TOP FLANGE
- VOLUTE l
Ox INSIDE\\// - 1
/ o, INSIDE
g, OUTSIDE v e
—— L - -
[’k ~d 7
ol S -~ —
- -..-‘

I g QUTSIDE /

o, = AXIAL PRINCIPAL STRESS
oy = CIRCUMFERENTIAL PRINCIPAL STRESS
oy OUTSIDE = —o, INSIDE

Ox

INSIDE
|

x
-6

AXITAL

2

Fig. 2.20. Thermal Stress Profile — Cylinder.

acfm of cooling air.

POSITION (in.)

Ten-Mw operation; about 200
54

Thermal stress calculations indicated that a small quantity of air cooling
will be required on the fuel-pump tank during zero-power operation to avoid ex-
cessive strain cycling during reactor power changes. The calculations also indi-
cated that forced air cooling of the coolant pump will not be required since the
stress range between zero-power and 1l0-Mw coperation is relatively small.

Calculations have been initiated to determine the necessity for variable
control of cooling air flow for the fuel-pump tank under all operating conditions
from zero to full power. The stresses shown on Figs. 2.19 and 2.20 were calcu-
lated by using an ORACLE code which assumes the spherical shell to be at constant
temperature. The effect of the thermal gradient on the thermal stresses in the
sphere are being determined with the IBM 7090 code which uses tangent cone
geometry in place of the sphere.

2.8.2 Water Test of MSRE Coolant Pump

Tests to determine the feasibility of conducting the coolant pumpl9 hot tests
in the fuel-pump volute have been completed. The tests indicated that the radial
hydraulic imbalance would cause rubbing between rotary and stationary parts. It
is planned to conduct the hot-proof test of the rotary element for the MSRE coolant
pumps in the hot-test stand for the fuel pump, using a fuel-pump impeller.

The water-test rig is being modified to permit hydraulic tests of the coolant
pump impeller-and-volute combination.

2.8.3 Advanced Molten-Salt Pumps

Pump Equipped with One Molten-Salt-Lubricated Bearing

The pump containing one salt-lubricated journal bearingzo continued to oper-
ate, circulating LiF-BeFp-UF), (62-37-1 mole %) at 1225°F, 1200 rpm, and 75 gpm.
The pump has operated for 11,772 hr and has been started 62 times thus far. There
has been no visible leakage of oil from the shaft seal.

2.9 GRAPHITE-MOLYBDENUM COMPATIBILITY TEST

The two natural convectilion loops were shut down and examined, as reported in
the metallurgy section of this report.

2.10 MSRE ENGINEERING-TEST LOOP

The engineering-test loop21 began operation April 20 with MSRE coolant salt
(LiF-BeF5, 66-34 mole %). The purpose of the first test was to evaluate the
effectiveness of the flushing operation by following the oxide content of the
salt immediately after filling the system for the first time.

2.10.1 Oxide-Flush Run

The loop remained in operation, isothermally at llOOOF, for approximately
1300 hr while 50 samples were removed from the pump bowl and drain tank. The
samples were removed in a hydrogen-fired copper dip tube by the use of a ball
valve and sliding seal arrangement. Both the dip tube and the sample were kept
in a helium atmosphere until loaded into the analyst's dry box. The results of
the oxygen analyses were so scattered that it was not possible to determine how
much oxygen had been dissolved or whether the solubility had been exceeded. As
seen in PFig. 2.21, the variation reported in oxygen content was from 300 to
1100 ppm.

55
UNCLASSIFIED
ORNL—LR—DWG 63260
1200 ,
[
1000
BeO ADDED
5 800 . . .
= L J
= 3 °
=
N :
= H
g 600 i
© L
2 .
[ ]
o ° H o °
-
S I O 3 L
400 o - .
o e °83 .
[ [ 2
e *° %
200
0
0 200 400 600 800 1000 1200

OPERATION TIME (hr)

Fig. 2.21. Results of Oxygen Analyses, Engineering Test Loop.

Between the 950th and 1100th hr (see Fig. 2.21), approximately 35 g of BeO
pellets were added to the system through the pump bowl. This was sufficient to
raise the entire oxygen content approximately 200 ppm; such a difference did not
show up in the analyses, however.

From this and other tests on the accuracy of the oxygen analysis, a re-
evaluation of eguipment and technique was begun by the analysts. The engineering-
test loop was dumped after 1300 hr of operation to begin other tests., The chro-
mium content of the salt increased from 170 ppm to 370 ppm during the above oper-
ation. This represents leaching of all the chromium from about a l/Z-mil thickness
of the metal surfaces of the loop. The chromium pickup 1s believed to be reason-
able for the initial operation of an Inconel system.

2.10.2 TFreeze-Valve Operation

The loop freeze valves and thelr operating characteristics are described in
Sec., 2.2.

The first attempt to dump the loop after the 1300-hr operation required over
an hour, which is excessive. The difficulty is thought not to have been in the
freeze valve but in the drain line. The alternate valve (fuel freeze valve),
which had seen the same conditions, opened in normal time. The partial plug may
have been caused by the settlement of undissolved BeO particles added to the loop
a week earlier. An attempt to establish a similar plug will be made by adding
additional BeO in the same manner.

56

2.10.3 Level Indicator for Molten Salt (Test Loop)

The differential pressure or bubble-tube type of level indicator system was
installed on the test loop along with the standard spark plug type. A helium
flow rate of approximately 1.8 scfh was used continuously through a 1/h—in.-OD
tube inserted into the liquid of the DANA pump bowl. The level was recorded from
the output of a O to 20 in. of water differential-pressure transmitter. Flow was
supplied by a constant-pressure-differential type of regulator.

The tube has plugged twice, once during a draining operation and again when
attempts were made to duplicate the original conditions. It was later realized
that the pressure in the pump bowl could increase faster than the pressure in the
bubble tube (considering the volume of the bubble tube and the helium flow rate),
causing salt to enter the tube.22

The system volume of the bubble level tube was therefore decreased by a
factor of 3, and, with the same flow rate and pressure change rate, the tube did
not plug. Since then the bubble tube has been used for 11 fill and drain oper-
ations and has been valved off (equalized so that salt rises in the tube to pump
level) for 700 hr and valved on for 300 hr of level indication, all without diffi-
culty.

2.10.4 (Craphite-Handling Facility

An 8-in.-dia vertical graphite container with a longitudinal frozen-salt
seal access has been designed and is being fabricated of INOR-G plate. The
graphite, on order, will have the same dimensions and properties as that for the
MSRE core. The installation will allow the examination of the graphite after
various operations, without contamination by air. As shown in Fig. 2.22 a dry
box will be lowered over the uccess flange to the freeze seal after the test
loop has been drained and cooled. Conventional gasket materials make connection
between the dry box handling facility and the graphite container while the box
is evacuated and purged of Op. The access flange is opened through the use of
the glove ports and the hoist. Two sealed graphite carriers are provided so
that a full length (65-1/L in.) sample may be added or removed in an inert atmos-
phere.

The access flange contains a cone-shaped plug, with cooling-air coils to
form the frozen-salt seal in an annulus. A standard metal oval ring gasket is
used for the gas seal.

The main purpose of the test will be for verification of laboratory-scale
tests of the removal of Oz from the graphite, the ease of graphite removal from
the container, wetting characteristics, completeness of draining, etc.

2.11 MSRE MAINTENANCE DEVELOPMENT
2.11.] Maintenance Plan

The dual maintenance approach, using both semidirect and remote techniques
as described in the design section (Chap. 1 of this report), is being implemented.
Figure 2.23 shows the building model with the primary-cell upper-shield beams in
place. Access to a small component, using the portable maintenance shield and
remote mamual techniques, will require the removal of three or four of these
beams. Replacement of a large component by use of remote maintenance techniquesz3
will require the removal of all the beams. They will be handled by the building
30-ton crane and a directly-operated 1lift fixture.

57

UNCLASSIFIED
ORNL-LR-DWG 63261

—=——GRAPHITE-HANDLING DRY BOX
{INSTALLED AFTER LOOP IS
DRAINED AND COOLED)

ACCESS FLANGE TO NEOPRENE GLOVE
GRAPHITE CONTAINER-] TMETAL OVAL RING GASKET
kY !
\ A
Q :
COOLING AIR I= fi 9
(BOTH EXTERNAL AND INTERNAL) : o !
i i l FROZEN-SALT SEAL IN ANNULUS
g STORAGE CYLINDER FOR ACCESS COVER
|
|

~ | e TO GRAPHITE CONTAINER
|

VACUUM CONNECTION

OUTLET

A —

WL ‘:‘N}p«

L]
¥ a2
& —_—
D |
|

GRAPHITE —I7 BB = | HOLDDOWN (DURING VACUUM)
i
- NN ] ‘ ‘_l " . —_—

= &  f.. ° T“FLOOR LEVEL

E‘L.““— — ) .

—= Y

GRAPHITE CONTAINER —=|
(8-in. DIA)

CONTAINER
SUPPORT

GRAPHITE

NN

INOR-8

’-rfl/z—in. SCHED-40

INLET

Fig, 2.22. Graphite-Handling Facility for the Engineering Test Loop.
|
|
- UNCLASSIFIED
VanTenance covrro. roow I
CooLm ceL B

REACTOR CELL ¢ / SPENT FUEL STORAGE CELL
TOP SHIELD BEAMS 4 /
;

WASTE TREATMENT CELL

Fig. 2.23. One-Twelfth-Scale Model of Building 7503, Column 3 to Column 9, Sections A, B, and C.

Figure 2.2h shows the primary cell with the lower shield beams in place. To
obtain access to a small component, the seal membrane (not shown) will be cut
away from the two beams to be removed and the portable maintenance shield set
over them and opened. The shield beams will then be removed, using the 30-ton
crane and an automatic 1ifting tong which can attach to and release from the =
beams without direct attendance, and the portable maintenance shield will be
closed. The operations of removing the lower shield beams and closing the port-
able meintenance shield will be conducted from the shielded maintenance control
room visible to the right and above the primary cell in Fig. 2.23.

Maintenance to or replacement of the component will be accomplished by work-
ing directly through the portable shield. To replace a large component the
entire seal membrane will be removed. The manipulator and telev:
be set up over the primary cell. From this point until the lowe
are replaced, all work must be conducted from the maintenance control room. All
the lower shield beams and the two support beams will be removed. Maintenance
or replacement will be accomplished with the manipulator, with vision provided
by the stereo-television system and through shielded windows in the maintenance
control room.

2.11.2 Remote-Maintenance Mockup

The remote-maintenance mockup of the MSRE is being constructed to develop,
test, and improve maintenance tools and techniques.2% It occupies the space -

59

UNCLASSIFIED
== PHOTO 37263
i o

PUMP MOTOR
PIPE HEATER
PUMP P

PUMP BOWL HEAT EXCHANGER

FREEZE
FLANGE

Fig. 2.24. One-Twelfth-Scale Model of Primary Cell.
60

left vacant following dismantling of the remote-maintenance demonstration facility
in the southwest corner of Building 9201-3. Most of the new structural steel was
erected, and the pump, reactor thermal shield, and some piping were installed.
Shop-fabricated models of the heat exchanger and primary loop piping are being
installed.

Heater, component, and instrument replacement appear to be straightforward
maintenance operations, except in areas of crowding, such as around the pump.
Extensive practice in the closely-duplicated pump mockup will resolve any problem
in this area. At present, it appears that the inventory of special long-handled
tools will be small because many operations have been reduced to the rotation of
a standard hex nut or to the vertical translation of a standard eyebolt.

Making and breaking the freeze flanges, which is the primary concern of the
mockup, includes the following: (1) removal of the freeze-flange clamp, (2) stowing
the clamp, (3) jacking the piping to separate the faces of the flange for gusket
replacement or to replace a component, and (4) providing force and visibility to
re-position misaligned flanges. Test equipment for this process was designed
(Fig. 2.25) and is being fabricated.

UNCLASSIFIED
ORNL-LR—DWG 63262

LIFT HOOK

(@) REMOVABLE

um PIPE HEATER
Al

(@
I SOCKET
WRENCH
LIFT HOOK AU

FLANGE
CLAMP

m’ N
FREEZE g
| I3
FLANGE - \. N * JACK

INSTALLATION

CABLE TO DISENGAGE
JACK HOOK

PERMANENT INSULATION

\

PIPE SUPPORT

STRUCTURAL STEEL

Fig. 2.25. Equipment for Jacking Flanges Apart in the MSRE.

61

2.11.3 MBRE Model

The one-twelfth scale model of Building 7503 follows closely the reactor
design and has aided in working out several design changes. The building and the
primary cell are shown separately in Figs. 2.23 and 2.24, respectively.

2.11.4 Portable Maintenance Shield

Conceptual layout drawings for a portable MSRE maintenance facility were com-
pleted. Orders were placed for approximately 80% of all materials required for
fabrication; most of these items have been received. Component detailing was
started.

2.1l2 BRAZED-JOLINT DEVELOPMENT

To standardize tools and techniques, all points in the salt-system piping
that will require a braze joint in the event of component replacement will be
1—1/2—in. sched-40 pipe run on a slope 3° from horizontal. A clear or readily
clearable space 3 ft along the axis of the pipe and 1 ft on either side; a perma-
nent base-plate 3 ft long and 2 ft wide, mounted 1 ft below and parallel to the
axis of the pipe; and access from above are required at each point.

After testing several designs, it was tentatively decided to use a sleeve
joint with a 1-in. engagement and a 3° taper, shown disassembled in Fig. 2.26 and
assembled and sectioned in Fig. 2.27. The male stub remaining in the system will
be machined in place to the required taper. The replacement component will carry
the female sleeve with the 82% Au - 18% Ni braze metal preplaced. The present
test pieces use two braze-wire rings preplaced in grooves in the sleeve
(Fig. 2.26). The Metallurgy Division is developing a technique to "tin" the
sleeve with about 0.005 in. of braze metal so that only cone surface will have to
wet during brazing.

The brazing will be accomplished by heating the joint to about 1800°F in an
induction furnace similar to that of Fig. 2.26. The furnace requires electrical
leads, cooling water, and purge gas. The furnace will be preplaced with the new
component and will be destroyed upon completion of the braze.

Figure 2.27 shows a completed, sectioned joint. Although braze metal did not
flow the entire length of the anmulus, an adequate seal was formed. Development
efforts are continuing in order to ensure wetting of the entire annulus. A photo-
micrograph and discussion of the Joint is in the Metallurgy section of this report
(Part II, Materials Studies).

Tools for use in the reactor have been designed to hold the pipe, cut it,
machine the taper, and assemble the joint. These will be tested in simulated
reactor geometries.
UNCLASSIFIED
PHOTO 36994

TNGHES
e 7

Fig. 2.26. Disassembled Test Furnace and Joint.

UNCLASSIFIED
T-20322

ONE INCH

Fig. 2.27. Assembled and Sectioned Brazed Joint.

64

2.13 MECHANICAL-JOINT DEVELOPMENT

The molten-salt reactor program requires simple, leak-detectable, radiation-
resistant joints to permit component replacement and maintenance, and several
types of disconnects have been developed (Fig. 2.28).

The integral dual-seal ring joint25 closed by steel spring clamps offers a
satisfactory disconnect for many reactor applications in the process temperature
range 100 to 1500°F. Remote operation is facilitated because only low torque is
required to operate the two bolts on the clamp. Use of the metal-to-metal seal
eliminates material radiation damage problems. Integral ring joints were selected
in preference to separate rings to prevent ring buckling when the connecting
flange halves are slightly cocked on makeup, to reduce the number of seal contacts
from four to two, to reduce the probability of foreign particles into seal surfaces,
and to eliminate the need to handle gaskets during remote operation of the
disconnect.

Two integral ring-joint spring-clamp flange sets were designed, built, and
tested for coupling E?z- and l—l/z-in. pipe, respectively. Final helium leak
rates with the 1/2-in. unit following eight assemblies and & total of 1020 thermal
cycles between 100 and 1200 to 1500°F were 9 x 10-6 atmospheric cc/sec at room
temperature, and 2 x 10-7 atmospheric cc/sec at 1500°F. Following fourteen assem-
blies and 69 steam cycles up to 400°F with the 1-1/2-in. unit, no detectable
readings above background were observed in a helium mass-spectrometer test on both

UNCLASSIFIED
PHOTO 369984

INTEGRAL DUAL-SEAL BLOCKS,
YOKE MOUNTED

RING-JOINT SPRING-CLAMP FOR

E HIGH-TEMPERATURE SERVICE
COMBINATION RING-
JOINT FLEXITALLIC
FLANGE |
SINGLE-CONE SEAL BLOCKS,
FOR YOKE MOUNTING

Fig. 2.28. MSRE Disconnects. -

65

inner and ocuter seals, assuring a seal tightness in excess of 1.3 x 10-8 atmos-
pheric cc/sec. The joint was also thermal cycled (250°F AT) with bending moments
of 4200 in.-1b applied both parallel and perpendicular to the clamp separation
line. ©No detectable leakage was noted.

A combination ring-joint - Flexitallic disconnect20 was developed to serve as
a multiple disconnect in crowded spaces. The joint contains a tight R-26 ring at
the periphery, and four l/Z—in. lines, each with a Flexitallic gasket. It 1s most
useful where small amounts of interline leakage can be tolerated but leakage to
the outside cannot be accepted.

Farly difficulties in achieving tight ring seals were eliminated by modifying
the usual flange-assembly procedure. In the successful procedure the center of
the joint is loaded temporarily with a C-ciamp before the flange-bolts are tight-
ened. Seals can then be consistently made with bolt-torque loads well within the
maximum limits recommended for the bolt threads. Proper seals can be obtained by
using this procedure, even with rings manufactured only to commercial specifi-
cations. The addition of the Flexitaliic inserts appears to add no further diffi-
culties.

No leakages in excess of 1 x 10-6 ¢td cc/sec were apparent for ring seals
(and 1 x 10-3 sta cc/sec for Flexitallic seals) in tests run with oval and octag-
onal rings and four l/Z—in. 150-1b Flexitallic inserts reused for three successive
assemblies. The flange was successfully steam-cycled to 250°F with 500-psi helium
pressurization in the buffer zone of the ring.

Development testing with cone-seat disconnects®! demonstrated that such
Joints can be made to adhere to the low leak rates required in reactor service.
Both single and multiple units were tested. Cone-seat units represent a great
simplification over other commonly used metal-to-metal seals for guick disconnects.
Low-torque single bolt loading, compactness, minimum space required for assembly
(l/H—in. travel), interchangeability of parts, and inexpensive construction are
some of the major advantages of this seal. The cone-seat disconnects are well
suited for reactor leak-detector-iine disconnects.

Development work with ycke-mounted l/fi— to l/B—in. pipe dual-seal block dis-
connects revealed that seals fashioned to variations of the ring-joint gecmetry
work reliably repeatedly. Test biocks with both solid and slotted integral oval
rings operating within 60°-included-angie grooves were operated in excess of
12 assemblies and 30 thermal cycles from 80 to 3259F, with leak rates of less than
8 x 10~7 std cc/sec as measured with a helium mass spectrometer. Additional test-
ing 1s planned.

Two commercial quick-disconnect ccocuplings were received from the On-Mark
Company. Preliminary testing indicated good sealing action at room temperatures.
On thermal cycling up to 3250F, however, one unit failed to seal on the tenth
downcyecle. Visual inspection revealed considerable peeling of the seal's silver
plating. Additional testing is scheduled.

2.14 STEAM GENERATOR

Preliminary design of a high-pressure, electrically heated, single-tube,
steam-generator - superheater was initiated. The generator would be a bayonet-
tube unit of the type proposed for the experimental molten-salt fueled 30-Mw power
reactor.20 Completion of the design was postponed until the feasibility of a
separator that will fit the small available space is determined.

A test chamber was designed and is under construction to test separators for
use in the steam generator. TIigure 2.29 is a schematic drawing of the test
UNCLASSIFIED
ORNL-LR-DWG 63263

§§“\———TEST SEPARATOR

AIR AND
"CARRYOVER

NN

i

ol
.

#

 WATER AND
"CARRYUNDER

Fig. 2.29. Separator Test Chamber,

67

chamber. Alr and water will be used in the tests. Water carryover in the air and
alr carryunder in the return water will be determined for several air and water
flow rates.

2.15 MSRE INSTRUMENT DEVELOPMENT
2.15.1 Pump Bowl Level Indicator

Development of a continuous level-element®? for use in measurement of molten
salt level in the MSRKE fuel and coolant salt pump bowl 1is continuing. As reported
previouslycY a graphite float assembly will be used in the level element of the
reactor pump bowl, and an INOR-8 hollow-ball float will be used in the level ele-
ment of the coolant salt pump bowl. Design of the graphite float and core assembly
has been completed, and two prototype units have been fabricated. This assembly
(see p 66 in ORNL-3122) has been designed to provide sufficient buoyancy, after
absorption of fuel salt by the graphite, to support the core assembly and to over-
come frictional forces.

A molten-salt-level test stand for use in testing prototype level transmitter
was designed. Construction of this stand, which is shown in Fig. 2.30, 1s nearing
completion.

The major effort in this program has been devoted to the development of a
differential transformer and core assembly which will be capable of contlnuous
operation in the range of 1200 to 1300CF and which will not exhibit excessive
shifts in zero, span, and phase of the transmitted signal as the temperature
varies over the operating range of 250 to 13009F. Several transformer designs
have been tested during the course of this work. A transformer design using a
ceramic-insulated nickel wire, manufactured by the Secon Metals Corporation, with
Fiberfrax paper insulation between the windings, chowed some initial promise of
success but falled when operated at elevated temperatures. Two transformers of
this type were tested. One developed intermittent shorts after 12 hr operation
at 1100°F. This failure produced rapid and erratic changes in the output signal.
Subsequent examination revealed that the insulation had cracked off the wire in
places, and several turns were shorted. Another transformer of this type failed
in the same manner at QOOOF.

Figure 2.31 shows a recent transformer design which promises success. Insula-
tion ig provided by three concentric, unfired, grade A lava sleevec. Bare nickel wire
is wound into threaded grooves machined into the inner two sleeves, forming the
primary and secondary of the transformer. The threcads act as insulation for the
wire., The external sleeve servegs to keep the wire in place. DSufficient clearances
are allowed to permit differential expansion between the wire and the sleeve. A
transformer of this type has accumulated more than two weeks of operation at tem-
peratures in excess of 1200°F., This period included 24 hr of operation at 1500°F
and several periods when the temperature was reduced to room temperature. 5So far,
there has been no evidence of insulation failure.

Calibration data obtained from this transformer are also very encouraging.
Figures 2.32 and 2.33 show calibration curves obtained at various temperatures
with constant current excitation and with constant vollage excitation. Note that
the sensitivity of the transformer, as indicated by the slope of the curves, in-
creagses with temperature when constant current excitation is used and remains
essentially constant when constant voltage excitation is used. This result was
entirely unexpected. It had been expected that the increase in transformer losses
would result in & decrease in sensitivity as the temperature was increased when
constant voltage excitation was used. A probable explanation for the increase in

8 UNCLASSIFIED
PHOTO 37084 |

Fig. 2.30. Molten-Salt-Level Test Stand.

)

LR R R S R R Dl B ll P R E 5 i AR
lLo 1 2 3 a
INCHES

Fig. 2.31. Lava-Insulated Differential Transformer.

UNCLASSIFIED
PHOTO 37261

69
70

UNCLASSIFIED
ORNL-LR-DWG 63264

50
1200°F ‘
15.5 mv/in. /
N | \»4 900°F ]
p i0.2 in,
100°F S mvn
12.6 mv/in. /. - ®
™
~ °
— ® -~
> 30 A
£ ° o
5 - I
z -
2
© 20
.
cb/.’
—
10 —
ROOM TEMPERATURE
5.6t mv/in,
. ‘
[
.
H
0 0.5 1.0 1.5 2.0 2.5 3.0 3.5

DISPLACEMENT {in.)

Fig. 2.32. Calibration Curves for Lava-Insulated Transformer, Constant Current
Excitation.

UNCLASSIFIED
ORNL-LR-DWG 63265

QUTPUT (mv)

500-mv AC EXCITATION

/‘
0/
/‘. CONSTANT VOLTAGE ,

__ IRON CORE

/
/ I
o 7' F = 1000 cps

8] 0.5 1.0 1.5 2.0 2.5 3.0 3.5
DISPLACEMENT [in.)

Fig. 2.33. Calibration Curves for Lava-insulated Transformer, Constant Volt-
age Excitation.
71

sensitivity isc found in the data presented in Fig. 2.34. These data, extracted
from the literature,3C indicate that the intrinsic induction (permeability) of
Armco iron increases with temperature at Low excitation levels. Thic increase in
permeability would result in an increase in coupling, with & resultant increase
in transformer sensitivity. The fact that the increase in sensitivity due to

the permeability effect almost exactly offsets the decrease due to transformer
losses 1is apparently a fortunate coincidence.

The offset between the curves in Figs. 2.32 and 2.33 is due to residual
voltage capacitively coupled through the transformer. This offset can be elimi-
nated by appropriate circultry in the receiving instrument. Figure 2.35 shows
a calibration curve obtained by using a Foxboro dynalog recorder, with special
input circuitry, as the receiving instrument.

Figure 2.34 indicates that the curie point of the Armco iron core occurs at
1L0CPF. Thic effect has been verified experimentally. Tests have shown that the
transformer becomes insensitive, and signal output repeatedly falls to zero when
the temperature is raised above 1400°F. This characteristic is not believed to
be objecticnable since the point where the failure occurs is above the expected
operating temperature range and since the failure is in a safe direction. How-
ever, an attempt is being made to extend the useable range of the transformer.

UNCLASSIFIED
ORNL-LR-DWG 63266

3.0 e

20— 4 | T H=0.3

INTRINSIC INDUCTION (kilcgauss)

200 400 600 800 1000 1200 1400 1600
TEMPERATURE (°F)

Fig. 2.34. Armco Iron Magnetization vs Temperature.
72

UNCLASSIFIED
ORNL-LR-DWG 63267

5
4
£
o 3 ]
=
o
=
W
x
0}
S
a 2
b=
>—
O
1000°F
1 R . _ —
1250°F
0
O 1 2 3 4 5 13)

ACTUAL POSITION (in.)

Fig. 2.35. Calibration Curves for Lava Transformer with Dynalog Readout.

A cobalt core, which has a higher curie point, has been fabricated. Tests will
be run to determine the feasibility of using this core with the present trans-
former design.

The lava-insulated transformer assembly is being mounted on an Inconel coil
form which will give mechanical protection to the coil. The complete transformer

assembly will then be tested to determine whether the transformer characteristics
are adversely affected by the coil form. If the results are satisfactory, a

prototype level transmitter will be assembled and tested on the molten-salt-level
test stand.

2.15.2 Single-Point Level Indicator

As previously reported,3l a conductivity type of level indicator is being
developed for use in the single-point measurement of the molten-salt level in the
MSRE storage tanks.

Extensive measurements of the resistivity of a molten salt have been made at
various voltages and frequencies, a new probe-design concept has been developed
and tested, and a conceptual design has been made for a two-point level probe
suitable for use in the MSRE application.

The salt resistivity was investigated to determine the reason for and nature
of the observed changes,32;33 in the apparent reslstivity of the salt with
changes in excitation voltage and to determine whether any barrier or interface
effects existed which would require the excitation voltage to be above some mini-
mum value. A spark-plug probe was used for these measurements. With the salt
in contact with the probe, a 60-cycle ac excitation voltage was applied between

the probe and the salt pot.

73

The excitation voltage was varied from 2 mv ac to

2 v ac, and the probe current was plotted as a function of excitation voltage.

The plotted curve was linear over the entire range,

thus indicating that there

are no interface effects in the region of interest when ac excitation is used.
These tests were repeated, using dec excitation and with the sparik plug positive

in one case and negatlve in another.

The plotted curves were not linear and thelr

shape changed considerably when the polarity of the excitation voltage was reversed.
From these tests it was concluded that a definite polarization effect exists when

de excitation is used and that the use of de excitation ic undesirable. Additional
tests were run in which the temperature of the salt was varied; the results have yet
to be interpreted.

Figure 2.36 shows a new design concept for a conductivity probe. In previous
probe designs both the potential and excitation leads were connected at the bottom
of the probe, and the signal level, when the salt was below the plug tip, was a

funection of the excitation current.

This produced a floating-zero effect and

required the detection of a small change in a relatively large voltage. In the
new design, the excitation lead is connected at a point above the probe tip, and
the signal is obtained by measuring the potential between the point where the

excltation lead is connected and the probe tip.

Since no current flows to the

probe tip when the salt is below the probe, a true zero characterictic is obtalned,
and the signal is a small voltage deviation from zero. This

detected.

A model of this probe was constructed and tested.

can be amplified and
It has operated

UNCLASSIFIED
ORNL—LR—DWG 63268

( )
N | B

S

EXCITATION

] READOUT

TEST UNIT NO.2
AS FABRICATED

SCHEMATIC

|

Fig. 2.36. Single-Point Level Indicator.

| MOLTEN SALT

SALT POT

74

several hundred hours at temperature in both fuel 130 (LiF-BeF2-UF),, 62-37-1 mole %)
and MSRE reactor salt (LiF-BeF;-ZrF),-ThFj-UF), 70-23-5-1-1 mole %) without apparent
damage. Figure 2.37 shows a calibration curve cbtained using MSRE salt. The

UNCLASSIFIED
ORNL-LR-DWG 63269

MSRE REACTOR SALT TEMPERATURE: 1369°F /

2.3 amp AT 1-kc EXITATION
TEST PROBE NO. 2. ’ T - ]

35

30

25 - - : e - —

|
20 -4 // s —

SIGNAL {mv, AC)

o] 1 2 3
SURFACE AREA OF PROBE IMMERSED IN SALT (in.2)

Fig. 2.37. Calibration Curve, Single-Point Level Indicator,

curve was obtalned by immersing the probe, which is 3/8 in. in diameter, to a
known depth and recording the output signal level. This signal level was then
plotted as a function of the calculated surface area exposed to the salt. Although
this curve is linear, it 1s not expected that the probe would be useable as a con-
tinuous level indicator unless some means can be devised to compensate for changes
in the resistivity of the molten salt with time. The curves shown in Fig. 2.38
were obtained using various excitation frequencies. From these curves it can be
seen that the sensitivity of the probe is increased by increasing the excitation
frequency. The use of higher excitation frequencies has the additional advantage
that stray 60-cycle pickup voltages can be rejected with appropriate filter
circuitry.

Figure 2.39 shows a conceptual design for a two-level conductance probe,
based on the new design concept and compatible with the configuration of the MSRE
storage tanks. Potential lead 1 is connected to the high-level contact plate.

75

UNCLASSIFIED
ORNL-LR-DWG 63270

80 I T
; |
| ;
|
70 T /
MSRE REACTOR SALT TEMPERATURE: 1369°F
EXCITATION CURRENT: 2.3 amp 1ke
) //
50 /
O
< v
e
£
@
S 500cps}
w
-J
-J
T
=
©
") /
30 /
20 -
10 100cps -
/
.—-—_'—___
50 ¢cps
0
0 1 2 3 4 5 6

SALT LEVEL (in.}

Fig. 2.38. Signal vs Frequency Characteristics for Single-Point Level Device.

Potential lead 3 and the excitation lead are connected to the excitation ring.
Potential lead 2 is connected to the high-level contact plate. Excitation voltage
is applied between the excitation terminal and the mounting flange. The low-level
signal is developed between potential leads 1 and 3, and the high-level signal isg
developed between potential leads 2 and 3. All leads are brought out through
l/M-in. INOR-8 tubes and are insulated from contact with the probe assembly except
at the point of connection. The assembly is all welded, and INOR-8 is the only
material which will be in contact with the salt. The assembly will be instailed
through a 3-in. ring-joint flange located at top center of the storage tank.
76

UNCLASSIFIED
ORNL-LR-DWG 63271

POTENTIAL LEAD |

/ POTENTIAL LEAD 2

POTENTIAL LEAD 3

EXCITATION TERMINAL

MOUNTING FLANGE

HIGH-LEVEL CONTACT PLATE

LOW-LEVEL CONTACT PLATE

Fig. 2.39. Conceptual Design, MSRE Conductance Level Probe, Two Level.

77

2.15.3 Temperature Scanner

A temperature-scanning system is being developed to display the output of
approximately 250 thermocouples attached to the reactor system pipes and components.
These thermocouples will be used to indicate the temperature of the pipes and com-
ponents during system startup and shutdown operations. The display of the tempera-
tures will be used to prevent excessive thermal stresses of the pipes and components
during these operations. An operator will observe the temperature profile of the
system and adjust heaters to keep the entire system at a common temperature.

The scanner will consist of three separate systems, each capable of handling
100 thermocouples. The thermocouple signals from the pipes and components in one
part of the reactor system will be switched at a 2000-point-per-second rate and
compared to a reference thermocouple attached to the pipe or component having the
slowest thermal response. If a temperature difference, elther high or low, exists
between the commutated thermocouples and the reference couple, an alarm signal is
produced. The difference signal is simultaneously transmitted to a 17 in. oscil-
loscope for display. If all signals are of the same value a straight line will be
. seen on the scope. A block diagram of a 100-point system is shown in Fig. 2.LO.

UNCLASSIFIED
ORNL—LR-DWG 58910R{

1200
RPM
MOTOR

|
|
\ (e}
o
D’j o)
Pam—
Vo) |
\n : W
» X_nj 1
fl:
|
|
i
]
i < )
o ! o
— : > o 00000
| ]
< i .
S ! AMPLIFIER 17" 0SCILLOSCOPE
W 1
o o \
> | 2§ :
Sk ]
clo '
x| ONZ '
= S
ol - !
L o~
o -
ALARM
MERCURY JET SWITCH
{100 CONTACTS) \ JDISCRIMINATOR]
< /) —
REFERENCE
THERMOCOUPLE AMPLIFIER ANNUNCIATOR
\_<J ON REACTOR

PIPE

- Fig. 2.40. Block Diagram of 100-Point Temperature Scanning System.

78

The alarm setpoints, high and low, will be adjustable through the alarm dis-
criminator. The signal is amplified to a suitable level to drive the oscilloscope
and alarm discriminator.

The absolute-temperature value of the commutated signals will be determined
by a spare thermocouple located adjacent to the reference couple. This couple
signal will be measured by a precise temperature indicator. Then, knowing the
calibration of the difference signal, all the temperatures can be determined from
the scope display.

A separate part of the reactor syctem will be monitored by each 100-point
scanner system. For example, the coolant system, fuel system, and fill-and-drain
system will each be monitored by & separate scanner.

The operator will monitor the scopes and adjust appropriate heaters to hold
the temperatures constant. If a section of the pipe or a component deviates
from the reference couple, an alarm i1s sounded. The operator then identifies the
high- or low-temperature position from the scope display and makes the necessary
heater adjustment.

2.15.4 Thermocouple Attachments and End Seals

The development of techniques and procedures for attaching sheathed thermo-
couples to INOR-C pipe walls and vessels i continuing. Sample gquantities of
1/10-in.-0OD, two-wire, single-conductor, and 1/8-in.-0D, two-conductor Inconel
and INOR-8 sheathed thermocouples, have been procured for use in this program.
Figure 2.41 shows some of the more promising methods of attaching these thermo-
couples to INOR-8 pipes and vessels. In all the methods shown a grounded hot
Junction is formed by fusing the conductors into the end seal of the sheaths,
using Heliarc welding techniques.

It was originally expected that attachments could be made by field brazing
the couples to the INOR-8 pipe or vessel with 82 Au - 18 Ni brazing alloy.
Although adaptor lugs have been successfully brazed to thermocouple sheaths, the
results of field brazing tests have so far been unsatisfactory.

Though the results of experiments to date are inconclusive, the use of
Heliarc-welded INOR-8 adaptor lug attachments appears promising. Since brazing
is preferable to Hellarc welding when thermocouples are to be attached to thin
wall components such as radiator tubes, efforts to develop satisfactory field
brazing techniques will be continued. TFigure 2.42 shows a method of installing
l/l6—in., single-conductor, two-wire, sheathed couples on pipe, presently being
evaluated. The sheaths are insulated from each other and from ground except at
the hot junction. This method of installation has the advantage that detachment
of a thermocouple from the pipe or vessel will be detected with conventional
burn-out circultry in the receiving instrument. Insulation of the sheath from
ground also increcases the effectiveness of the sheath in shielding the thermocouple
wire from stray magnetic fields produced by currents in adjacent pipe heaters. In
the installation shown, ceramic beads are used for insulation. Other types of in-
sulations are being investigated for use in areas where space limitations will not
permit the use of beads.

In addition to permitting the detection of detached couples, the use of
l/l6—in., single-conductor, two-wire sheathed couples offers the additional advan-
tages of smaller size and simplification of installation of cold end seals.3

22 AWG Chromel wire

22 AWG Alumel wire

jiom Incenel sheath, MO
insulation. Wire, single conductor.

Yinedion incons sheath, H30
insulation, Wire, duplex, 22 AWG,
Chromel/Alumel

=

INOR-8 odopter lugs Assemblies

Fig. 2.41. Typical Thermocouple Attachments.

UNCLASSIFIED
T-20401

Sheath Heliarc welded to lug. Lug
Heliarc welded fo INOR-8.

Sheath brazed 1o lug with 82 Au—
18 Ni olloy. Lug brazed to INOR-3
with 82 Au~

Ni olloy.

Sheath brazed fo lug with 82 Au—
18 Ni alloy. Lug brazed to INOR-8
with 82 Au-18 Ni alloy.

Sheath brozed 1o INOR-8 with 82
Au=18 Ni alloy.

Sheath Heliore welded 10 lug. Lug
Heliarc welded 1o INOR-8.

Sheath Heliarc welded 1o lug. Lug
Haliare welded 1o INOR-8.

Sheath brazed 1o lug with 82 Au—
18 Ni alloy. Lug brozed 1o INOR-8
with 82 Au=18 Ni alloy.

Sheath brazed 1o INOR-8 with 82
Au-18 Ni alloy.

UNCLASSIFIED
T-20405

Fig. 2.42. One-Sixteenth-Inch Single Conductor, Sheathed Thermocouple Attachment to INOR-8 Pipe.

08

81

REFERENCES

1. MSRKP Prog. Rep. Feb. 28, 1961, ORNL-3122, p 206-27.

2. P. P. Holz, Ring-Joint Spring-Clamp Disconnects, ORNL CF-61-7-92 (July 1k, 1961).

MSRP Prog. Rep. Feb. 28, 1961, ORNL-3122, p 27-28.

3
4. Ibid., p 17.
5. Ibid., p 31.

6. R. B. Gallaher, MSRE Sampler-Enricher System Proposal, ORNL CF-01-5-120
(May 24, 1961).

7. MSRP Prog. Rep. Feb. 28, 1961, ORNL-3211, b 30.

8. 1Ibid., p 37-ho.

9. E. S. Bettis, MSRE Component Design Report, MSR-61-G7 (June 20, 1961).

10. MSRP Prog. Rep. Feb. 28, 1961, ORNL-3122, p 47.

11. P. G. Smith, Water Test Development cf the Fuel Pump for the MSRE,
ORNL CF-61-8-43 {August 1961).

12. MSRP Prog. Rep. Feb. 28, 1961, ORNL-3122, p 13k4.

13. MSEP Prog. Rep. Feb. 26, 1961, ORNL-3122, p 51.

14. Ibig., p 51.
15. Ibid., p 49.
16. Ibid., p 51-52.
17. Ibid., p 51.

18. T. B. Fowler and E. R. Volk, Generalized Heat Transfer Code for the IBM 704
Computer, ORNL-2734 (Oct. 16, 1959).

19. MSEP Prog. Rep. Feb. 28, 1961, ORNL-3122, p 53.

20. Ibid., p 53.
21. Ibid., p 53.

22. J. L. Crowley, Bubble Tube Level Indicator Operation on the Enginecering Test
Loop, MSR-61-98 (Aug. 14, 1961).

23. MSRP Prog. Rep. Feb. 28, 1961, ORNL-3122, p 41.

24, Ibid., p 40-41.

25. P. P. Holz, Ring-Joint Spring-Clamp Disconnects, ORNL CF-61-7-92 (July 14, 1961).

26. P. P. Holz, Combination Ring-Joint Flexitallic Disconnect (in preparation).

27.

28.

29.
30.

31.
32.

33.

3k,

82

P. P. Holz, Development of a Six-Station Manifold Cone-Seat Disconnect,
ORNL CF-61-5-117 (May 18, 1961).

L. G. Alexander et al., Experimental Molten-Salt-Fueled 30-Mw Power Reactor,
ORNL-2796, p 30-31 (Mar. 38, 1960).

MSRP Prog. Rep. Feb. 28, 1961, ORNL-3122, p 6h4.

D. E. Meehan, "Magnetic Properties of Core Materials at Elevated Temperatures, "
Electronic Design, May 11, 1960, p 26, Fig. 13.

MSRP Prog. Rep. Feb. 28, 1961, ORNL-3122, p 67.

N. D. Greene, Measurements of the Electrical Conductivity of Molten Fluorides,
ORNL CF-54-8-6L (August 195k).

R. F. Hyland, Test of a Resistance-Type High-Temperature Sensor for the
Continuous Measurement of Salt Level in Molten Salt Fueled Reactor,
ORNL CF-61-3-25 (March 1961).

MSRP Prog. Rep. Feb. 28, 1961, ORNL-3122, p 6l.

3. REACTOR ENGINEERING ANALYSIS

3.1 REACTOR PHYLICS

3.l.l Analysis of MSRE Temperature Coefficient of Reactivity

Tne temperature coefficient of reactivity in the MSRE arises from two prin-
cipal effects, namely, the expansion of the fuel salt due to a rise in salt tem=-
perature and the shift in the neutron temperature of the reactor due to a rise in
the grapnite temperature. Both effects increase the leakage from the reactor and
produce a negative temperature ccefficient.

The calculation of the temperature coefficient was done according tc the
following assumptionss:

1. The neutron temperature and graphite temperature are identical.

2+ The effects of changes in reactor size and fuel channel size are

negligible.
3. The Fermi age 1s a function only of the grapnite density.
L. A two-group, bare reactor model is adequate.
5. Resonance escape probablility, fast effect, and 7 of U235

are 1lndependent of temperature,

6. Microscopic transport cross sections are independent of temperature.

With these assumptions and the current design data,l it was found that the tem-
perature coefficients of reactivity are:

salt: = 2.8 x 10'5/°F
grapnite: - 6,0 x 10-5/°F
total: - 8.8 x 10'5/°F
3.1.2 Rod Worth, Flux, and Power Density Distributions in the MSRE

Two-group, two-dimensional calculations were done using the IBM 7090 program
Equipoise=3 (ref 2) to obtain estimates of the worth of the currently proposed con-
trol rods in the MSRE. The reactor model used 1n the calculations is shown in
Fig. 2.1. The reactor was assumed to be a rectangular parallelepiped with a square
base, with radial geometric buckling equal to the geometric buckling of the actual
cylindrical core, and a height equal to the height of the actual cylindrical core.
This assumption was convenient since x-y geometry i1s used in the code calculation;
a pseudocylinder could be constructed, but more mesh points would have to be used,
with consequent increase in computing time. Control rod regions are labelled 1,

2, and 3 in Fig. 3.l1; the region marked 5 is the graphite-sample holder, which was
assumed to be identical to the core in these calculations. With the rods inserted,

83
84

UNCLASSIFIED

ORNL-LR-DWG 63272

Ax=504cm Ax=2.54 cm Ax =5.04 cm
4 2 3 4 5 6 7 8 9 {0 {1 12 14 16 18 20 22 23 24 25 26 27 28 29 30 31 32 33

21,

O ©® N b N

8 in,

Fig. 3.1. Square Model of MSRE Core.

the rod regions were treated in the thermal group by the following logarithmic de-

rivative boundary condition:

- Dgaggan o

(1)

where D is the diffusion coefficient in the medium adjacent to the rod, afi/an de-

notes the derivative of the flux in a direction normal to the surface, ¢ is t
flux at the surface of the rod, and C is a constant. The constant C is relat
to B, the albedo of the rod region, by the relation

he
ed

(2)

85

The albedo of the rod region was assumed to be given by
2R
1 -p = ) (3)

where R 1s the radius of the rod and a is the side of the square block enclosing
the rod. For the currently proposed rods, 2R = 0.990 in. and a2 = 2 1in., giving

1 -6 = 0495 (&)
C = 0.16k (5)

With the rods out, diffusion theory was applied in the rod regions. Results of
the calculations are given in Table 3.1. It will be noted that rod 1 (and, by
symmetry, rod 3) was worth slightly more than rod 2; this is due to the flux de-
pression caused by the INOR-8 rod thimbles. If there is an appreciable amount of
INOR-8 in the sample holder, this effect will be reduced, since the presence of
the sample holder was neglected in these calculations.

Table 3.,1. Rod Worths in the MSRE

Rod Configuration Worth (% 6k/k)
Rod 1 in, 2 and 3 out 2.9
Rod 2 in, 1 and 3 out 2.8
Rods 1 and 3 in, 2 out 5.3
Rods 1 and 2 in, 3 out 4.9
All rods in 6.7

Flux and power density distributions for the core were calculated by using
the somewhat more complicated model shown in Fig. 3.2, assuming all rods to be
withdrawn. The results are plotted in Figs. 3.3 and 3.4; flux and power-density
shapes for a homogeneous bare reactor were included for comparison. The addition
of the rod thimbles lowered the peak-to-average power density ratio by about 8%
of the homogeneous reactor value and moved the position of peak power density to
a point about 6 in. away from the centerline. The effect of the sample holder is
again evident in the higher flux along traverse A-B than along A=C; but since the
presence of INOR-8 in the sample holder was neglected, this effect should be smaller
than that shown here.

3.1.3 Gamma-Ray and Fast~Neutron Heating in Thimbles

Estimates were made of the gamma-ray heating in the control-rod thimbles
using the IBM 7090 program Nightmare.3 The peak gamma=-ray heat deposition rate,
at the reactor midplane, was 2.5 w/cms; the value at the upper end of the core
was asbout 0.5 w/cm®, but this value is uncertain since the method used in the
Nightmare program is not applicable to two-dimensional reactors.

The fast-neutron heating in the rod thimbles was estimated assuming the
heating to be proportional to the fast flux and assuming that the production of
a thermal neutron by slowing down represents an energy deposition of 2 Mev at
the point of production. The peak heating rate in the rod thimbles (with the
control rods out) was estimated to be 0.12 w/cm®,
86

UNCLASSIFIED
ORNL-LR-DWG 63273

i
|
|
| VESSEL
I
[ ANNULUS
t
| CAN
[
| GAP
Tc
|
\
I
!
l
|
|
\
| .
‘ N
\
|
\ -
|
|
\
|
|
E
|

2 | 3
|

o -—g.fi-\-__.__ ———————————————————— —“—mfl»fi—~——~—f —

I

4 | 4
|
!

BLOCKS {, 2, AND 3 ARE CCNTROL ROD THIMBLES
BLOCK 4 IS THE GRAPHIC SAMPLE BLOCK

Fig. 3.2. Reactor Model for Radial-Flux Calculations.

3,1.4 Reactors with INOR=-8 Fuel Tubes

Some nuclear calculations were done for molten=-salt reactors with INOR-8
tubes separating the fuel salt from the graphite. The core size was assumed to
be that of the present reactor (5S¢ in. in diameter and 66 in. high). Fuel-volume
fractions of 0.10, 0.1k, and 0.18, and tube-wall thicknesses of 0.040, 0.060, and
0.080 in. were used. Calculated total system inventories of U235 are plotted in
Fig. 3.5 as a function of tube-wall thickness, with fuel volume fraction as a
parameter; results previously obtained for reactors without tubes are included
for comparison.

Tn addition to raising the U®3% inventory, the use of the fuel tubes de-
creased the magnitudes of the temperature coefficient of reactivity, the average
value of 1, and the percentage of fissions caused by thermal neutrons.

L)l
¢(x) /¢, NORMALIZED RADIAL FLUX

2.2

2.0

1.4

0.6

87

UNCLASSIFIED
ORNL-LR-DWG 63274

|
\/ l

BARE HOMOGENEOUS REACTOR

\ FAST, 4-8 TRAVERSE

'd\\
FAST, 4—-C TRAVERSE
S0

THERMAL,
A-8 TRAVERSE

THERMAL ,
A—C TRAVERSE

0 10

20

30

x, DISTANCE FROM CENTER LINE (in.)

Fig. 3.3. Calculated Radial Flux in MSRE.
POWER DENSITY / AVERAGE POWER DENSITY

2.2

2.0

88

UNCLASSIFIED

ORNL-LR-DWG 63275

~~. —-BARE HOMOGENEOUS REACTOR
/ A -8 TRAVERSE
|
N !
N\, __-A4-C TRAVERSE
\
0 10 20

x, DISTANCE FROM CENTER LINE (in.)

Fig. 3.4. Calculated Radial Power-Density Distribution.
89

UNCLASSIFIED
ORNL-LR-DWG 63276

200 ‘
: FUEL VOLUME
FRACTION

_ 150 : +—
fe]
M
o

]

w

(@]

o

>

)_

[aag

o

Zz 100 |— -

(W)

>

<

b3

w

—

[fa)

o

=

'C_) .\

P50 \,O,‘, __‘_ _ _ _ _ - -

i
‘ |
O
0 0.020 0.040 0.060 0.080

INOR-8 WALL THICKNESS {in.)

Fig. 3.5. System Inventory for Reactors with INOR-8 Fuel Tubes.

3.2 NITROGEN=16 ACTIVITY IN FUEL SALT

A preliminarg estimate® of the radioactive sources within the reactor cell
showed that the N'€ activity from the Flg(n,a)N16 reaction in the fuel salt was
one of the major contributors. This estimate was based on a fission-spectrum-~
averaged cross section and the conservative assumption that the total reactor flux
had a fission spectrum. The K© activity has been re-estimated by integrating the
reacticon cross section obtained from BNL=-325, with the fast neutron fluxes obtained
from rmltigroup reactor calculations.”® A lower value was obtained, but N® activ-
ity still gives an appreciable dose. At a reactor power of 10 Mw(th), the average
production rates obtained were 1.8 x 10° atoms of N8 per sec per cc in the headers
and coolant annulus, and 1.1 X 10%° in the core proper. Assuming these production
rates constant over the respective regions gave the activities listed in Table 3.2.
90

Table 3.2. Nitrogen-16 Activity in Fuel Salt System

Location 0 Activity, (dis sec™ cec™ of fuel)
x 101°
Pressure vessel entrance 0.27
Core proper entrance 0.3U
Core proper exit 0.65
Pressure vessel exit 0.5

3.3 RESIDUAL ACTIVITY IN PUMP BOWL AND HEAT EXCHANGER

Two of the major sources of residual activity in the reactor cell will be
the pump bowl and the heat exchanger if appreciable quantities of fission prod-
ucts plate out in either of these components. The magnitude of these sources was
estimated, making certain assumptions about the behavior of the fission products.
In addition, the average energy release associated with the decay of these prod-
ucts was estimated.

The residual activity in the pump bowl was based on the assumption that all
the daughter products of xenon and krypton plate out at the point of formation.
The method of Stevenson® was used to determine the formation rate of daughter
products. Considering the gamma activity to be equivalent to l-Mev photons emit~
ted from a point source, the calculated dose rate 10 £t from the pump bowl was
1000 and 400 r/hr after 10 and 30 days cooling time, respectively, following
1 year's operation at 10 Mw.

The residual activity in the heat exchanger was based on the assumption
that 100% of the isotopes which might plate out on INOR-8 (ref 7) plated out in
the heat exchanger. Assuming that both parents and daughters remained in the -
heat exchanger and that the gammas could be represented as ~l-=Mev photons origi-
nating from a point source, the dose rate at a distance of 10 ft was 2 x 10% and
1.4 x 10* r/hr after 10 and 30 days cooling time, respectively, following 1 year's
operation at 10 Mw.

The estimated beta- and gamma~energy emission rates in the heat exchanger
following a year's operation at 10 Mw were, respectively,

18.4 e"7'8lt + 9.56 o 100t 5.19 e'0'0u38t, 10°

Q Btu/hr ,

B

10.0t -0.896t .

20.0 e~ + 9.01 e e'o'00692t,

Q lO3 Btu/hr ,

y 13.9

n

where t = cooling time in hours: O St < 24 hr.

It was assumed that the heat loss from the primary heat exchanger through the in-

sulation is linear with the average temperature of the metal in the heat exchanger,

that there is 5-kw loss at 1200°F (no coolant circulation), and that all the energy

emitted by the fission products is absorbed by the metal., Then, for 2000 1lb of

metal in the heat exchanger, the peak temperature was ~1350°F and occurred about

L hr following reactor shutdown and stoppage of coolant flow. .
?1

3.4 REACTOR-CELL SHIFLDING

3.4.1 Top Shield

The dose rates to be expected above the reactor cell have been re-estimated,
using the new values of the N'® activity in the fuel salt (see sec 3.2) and the
revised layout of the fuel-salt system. Basic changes in tne layocut pertinent to
shielding considerations consisted of the removal of the pump from directly over
the reactor vessel and the removal of the INOR-8 shield located in the upper header.
The resulting 2-ft-diam hole in the top of the reactor shield is to be used for
removing the graphite sample and control equipment (the effect of this equipment on
shielding requirements was neglected in the shielding calculations).

Both a solid shield and the existence of a 1/2-in. crack between the upper
snield beams has been considered. Dose-rate estimates above the crack were based
on tihe uncollided photon flux, whereas the sclid shield estimates included bulld-
up. Self-shielding by the source and source container were included in all cases.
In addition to the H'® activity, other radiation sources considered were fission
product activity, tnermal-neutron capture gammas, prompt-fission gammas, and prompt-
fission neutrons.

Table 3.3 gives the calculated dose rates directly over thie reactor vessel
for the present shield thickness (upper shield beams, 3.5 £t of ordinary concrete;
lower shield beams, 3.5 £t of barytes concrete). Results for the balance of the
shield have been reported,® and, except for the region above the reactor vessel,
the general radiation level is less than 2.5 rw/hr for the solid sihield.

It is presently planned to insert a filler material in the cracks between the
upper shield beams. Should the measured dose rate during reactor operation still
be excessive, provision has been made for placing additional shield material as
needed.

Table 3.3. Dose Rates Above the Reactor Top Saield (10-Mw
Reactor Power; Shield Consists of 3.5 It of Barytes and
3.5 £t of Ordinary Concrete )

Dose Ratcs (mr/hr)

With Solid Shield With 1/2-in. Crack Between
Lource Upper Snield Beams
Gamma Rays from:
Core 7.8 51
Iron capture 5 14
o 0.7 k.9
Concrete capture 1.3 I ,6

Neutrons from:
Core 1.8 L4600

92

3.4.2 $ide Shield

Estimates of the dose rates outside the side shield of the reactor were made
considering reactor operation at 10 Mw for one year and various times after shut-
down, assuming that the fuel system was not drained. In obtaining these estimates,
the fuel-salt-system components were considered individually. An IBM 7090 shielding-
calculation code prepared by E. D. Arnold® was used for the calculations. Activity
due to the decay of fission products was obtained from the IBM Internuc code ,+°
Other sources®® consisted of the saturated fission-product activity and capture
gammas in the reactor shield. The calculations included the effect of 2 in. of
steel (containment-vessel wall) and the self-shielding associated with the source
medium and source container. Typical results obtained are given in Table 3.4.

Table 3.4. Dose Rates Outside the Side Shield Due to Individual
Fuel-Salt-System Components (20-ft Attenuation
Distance Between the Source and Dose Point)

Dose Rate (mr/hr)

Component 7 ££-10 Mw® L ££-105 sec cooling® -
Thermal shield 40 0
Heat exchanger, side view 3 1.5
Heat exchanger, end view 0.k 0.3
Pipe (6 ft£ 5 in.), side view 0.6 0.5
Pump bowl, side view 0.5 Q.4

aT ft of ordinary concrete, 10-Mw reactor power, 1 year's operation.

bu 't of ordinary concrete, 10° sec cooling time following 1 year of
operation at 10 liw.

REFERENCES

1. W. L. Breazeale, MSRE Design Data Sheets, MSR-61-100 (Aug. 15, 1961). »

2, T. B. Fowler and M. L. Tobias, personal commnication.

3. M. L, Tobias, D. R. Vondy, and M. P. Lietzke, Nightmare - An IBM 7090 Code .
for the Calculation of Gamma Heating in Cylindrical Geometry, ORNL report
(in preparation).

4. D. W, Vroom, Preliminary MSRE Gamma Ray Source and Biological Shielding
Survey, ORNL CF-61-4-97 (Apr. 20, 1961).

5« Ca. W. Nestor, personal communication, May 1961.

6. R. B. Stevenson, Radiation Source Strengths in the Expansion Chamber and
Off-Gas System of the ART, ORNL CF-57-7=-17, (Nov. 18, 1957).

7. I. Spiewak, Proposed MSR Small Pump Loop Program for FY 1962, MSR-61-12,
(Feb. 10, 1961},

8. E. S. Bettis, MSRE Component Design Report, Section VIII, MSR-61-67,
(June 20, 19617).

9. E. D. Arnold, Shielding Design Calculation Code, ORNL-~304l, to be published.

10, D. M. Shapiro, EAPPR, Internuclear Co., TM=-DMS-58-1, (Dec. 31, 1958).

PART Il. MATERIALS STUDIES

4. METALLURGY

. 4.1 DYNAMIC-CORROSION STUDIES

4.1.1 Examination of INOR-8 Forced-Convection Loops

Examinations of eilght additional INOR-8 forced-convection loops were
completed during this period, concluding the investigation of long-term
corrosion effects in fluoride systems. 1In all, 24 loops (9 of Inconel and
15 of INOR-8) were operated under this program. Table L.l gives a summary of
the operating times of these tests.

Table L.1. Summary of Forced-Convection Loop Operation

Length of Test TNOR-8 Inconel
(hour ) No. of Total Hours No. of Total Hours
Tests Operated Tests Operated
3,000 to 5,000 2 6, 440
. 6,000 to 10,000 6 52, 680 5 4L, 760
11,000 to 15,000 L 54,370 2 28,190
16,000 to 20,000 5 99,950
Total hours accumulated 207, 000 79,390

A summary of the operating conditions and test findings for the 24 loops is
presented in Table 4.2. The last eight loops in this table represent those
examined during this period.

With the exception of loops 9354-3 and MSRP-16, corrosive attack was
1imited to the hottest segments of the loops and was in the form of a pitted
surface layer. Metallographic examinations of four 20,000-hr loops (MSRP 6, 7,
10, 11) showed corrosion depths generally in the range of 1/2 to 2 mils for
systems based on LiF-BeF, and NaF-BeF,. Figure 4.1 shows the general appearance
of attack in these loops. In loop MSRP-6, one of those which circulated a salt
. similar to the proposed MSRE fuel mixture but without ZrF), the depth of attack

was greater, reaching a maximum of 2 mils, as shown in Fig. 4.2.

. 93
Table 4.2, Operating Conditions of Forced-Convection Loops and Results
of Metallographic Examinations of Loop Materials

Maximim
Duration Fluid-Metal Flow
Loop of Test Interface AT  Reynolds Rate
Number Material (hr) Salt Mixture> Temp. {°F) (°F) Number (gzpm) Results of Metallographic Exemination
9075-1 Inconel 8,801 122 1250 100 5000 2.0 Intergranular penetrations to 1-1/2 milsP
93k4-1  Inconel 8,892 123 1300 200 3250 2.0 Heavy intergranular voids to 38 mils®
93442  TInconel 8,735 12 1200 100 8200 2.5 Intergranular voids to 8 mils®
9354-1  INOR-8 14,563 126 1300 200 2000 2.5 Heavy surface roughening and pitting to
1-1/2 milsd
9354-3  INOR-8 19,942 Bl 1200 100 3000 2.0 No attack, slight trace of metallic deposit
in cooler coil
935L-4  INOR-8 15,140 130 1300 200 3000 2.5 No attack®s®
9354-5  INOR-8 14,503 130 1300 200 3000 2.5 No attackS
3T77=-1 Inconel 3,390 126 1300 200 1600 2.0 Intergranular voids to T m:i.ls-b
9377-2 Inconel 3,046 130 1300 200 3000 2,0 Intergranular voids to 8 mils
9377-3 Inconel 8, 76k 131 1300 200 3Loo 2.0 Intergranular and general voids to 1k mils®
9377-4+  Inconel 9,57L 130 1300 200 2600 1.75 Intergranular and general voids to 1lb mils
9377-5 Inconel 15,038 13k 1300 200 2300 1.8 Intergranular voids to 24 mils
9377-6  Inconel 13,155 133 1300 200 3100 1.8 Intergranular voids to 13 mils?
MSRP-6  INOR-8 20,000 134 1300 200 2300 1.5 Pitted surface layer to 2 mils
MSRP-7 INOR-8 20,000 133 1300 200 3100 1.8 Pitted surface layer to 1 mil
MSRP-8  INOR-8 9,633 124 1300 200 000 2.0 No attack
MSRP-9  INOR-8 9,687 13L 1300 200 2300 1.8 No attackd
MSRP-10 INOR-8 20,000 135 1300 200 3400 2.0 Pitted surface layer to 1/2 mil
MSRP-11 INOR-8 20,000 123 1300 200 3200 2.0 Pitted surface layer to 1 mil
MSRP-12 INOR-8 14,498 13k 1300 200 2300 1.8 No attack
MSRP-13 INOR-8 8,085 136 1300 200 3900 2.0 THeavy surface roughening and pitting
MSRP-1L4 INOR-8 9,800 Bult 1+0.5 U 1300 200 Pitted surface layer to 1/2 mil
MSRP-15 INOR-8 10,200 Bult 1%4+0.5 U 1400 200 Pitted surface layer to 2/3 mil
MSRP-16 INOR-8 6,500 Bult 14+0.5 U 1500 200 Moderate subsurface void formation to
I mils
85alt Compositions: bMSRP-Quar. Prog. Rep. Oct. 31, 1958, ORNL-2626, p 5L4-55.
éfi fiii-iig-gzégfléi-gglgshioigli * °MSRP Quar. Prog, Rep. Oct, 31, 1959, ORNL-2890, p 35-i2.
122 NoF-Zrf),-UP), 57-h2-1 mole % “MSRP Quar, Prog. Rep. Jan. 31 and April 30, 1960,
3 NaF-BeFo-UF),, 53-46-1 mole % e e
124 NoF-BeFo-ThF),, 58-35-T moled% -2913, p 36-3%.
126 LiF-BeF,-UF), 53-46-1 mole % e
130 LiF_Bng_UFJ, EQ_BE'i oL ? MSRP Quar. Prog. Rep. July 31, 1960, ORNL-3014, p 55-58.
131 LiF-BeF,-UF,, 60-36-4 mole % £
133 LiF_Bng_Thfiu, T1216-13 mole % MSRP Proz. Rep. Feb. 28, 1961, ORNL-3122, p T7-81.
13k LiF-~BeF,-UF),-ThF),, 62-36.5-0.5-1 mole %
135 NaF-BeFo-UF),-ThF), 53-45.5-0.5-1 mole %
136 LiF-BeF2-UF), T0-10-20 mole %
Bult 1%0.5 U LiF-BeFp-ThF),~UF),, 67-18.5-14-0.5 mole %

1Z0)
95

UNCLASSIFIED
T-20553

Fig. 4.1. Appearance of Specimen Removed from Point of Maximum Salt-Metal Interface Temperature
Region (1300°F) of INOR-8 Forced-Convection Loop MSRP-10. The loop was operated for 20,000 hr.

UNCLASSIFIED
T-20038

Fig. 4.2. Appearance of Specimen Removed from Point of Maximum Salt-Metal Interface Temperature
Region (1300°F) of INOR-8 Forced-Convection Loop MSRP-6. The loop was operated for 20,000 hr.

96

Examination of loop 9354-3, which circulated a coolant salt of the system
NaF-LiF-BeF, for 19,942 hr, revealed no evidence of attack in either the hot
or cold areas of the loop. However, macroscopic examination of specimens
removed from the end of the cooler coil did reveal the presence of a small
amount of metallic deposit. Analyses of the deposit have not been completed.

The last three loops in Table 4.2 afford an evaluation of temperature
effects on corrosion by the system based on LiF-BeF,. Two of the loops, which
operated at maximum temperatures of 1300CF and ll&OOgF, respectively, sustained
approximately the same depths of attack (1/2 to 2/3 mil) after 10,000 hr. Loop
MSRP-16, which operated at 1500°F, exhibited a considerably higher rate of
attack (~l4 mils) in only 6500 hr. Corrosion at 1500°F was manifested by sub-
surface void formation, and the surface film that is characteristic of the
corrosion process at lower temperatures was conspicuously absent, as shown in

Fig. b4.3.

A final report covering the results of all forced-convection loops
operated under the MSRE program is being written.

UNCLASSIFIED | @
T-20428 Gz

o

£

Fig. 4.3. Appearance of Specimen Removed from Point of Maximum Salt-Metal Interface Temperature
Region (1500°F) of INOR-8 Forced-Convection Loop MSRP-16. The loop was operated for 20,000 hr.

4.1.2 Molybdenum-Graphite Compatibility Tests

As previously reported,]‘ tests have been in progress to investigate the
compatibility of molybdenum in contact with graphite, INOR-8, and the proposed
MSRE fuel mixture (LiF-BeFy-ZrF)-ThFy-UF), 70-23-5-1-1 mole %). The tests were
carried out in INOR-8 thermal convection loops which incorporated test sections
containing type RO025 graphite, molybdenum strips, and INOR-8.2 The graphite,
in the form of cylinders 6 in. long, was encased in a 1-1/4-in. sched-80 INOR-8
pipe section and mounted in the hot leg of the thermal convection loops. The
first test (loop 1250) contained two graphite cylinders, and the second (loop
1251) contained four cylinders. The salt-to-graphite volume ratios of the two

97

loops were approximately 2:1 and 1:1, respectively. Table 4.3 summarizes the
operating conditions of these tests.

Table 4.3. Operating Conditions for Graphite-Molybdenum Compatibility Tests

Molybdenum: Universal-Cyclops arc-cast Mo + 0.5 T alloy
used in as-rolled condition.

Graphite: ©National Carbon Grade R-0025.

Fluoride mixture used: Proposed MSRE fuel — LiF-BeFy-ZrF) -ThF) -UF),
(70~23-5-1-1 mole %).

Condition Loop 1250 Loop 1251
Max. salt — metal interface temp. (°F) 1350 1300
Max. salt — graphite interface
temp. (°F) 1360 £ 5

Max. salt temp. (OF) 1300 1300
Min. salt temp. (OF) 1140 1140
Loop AT (°F) 210 160
Operating time (hr) 5585 5545
Salt-to-graphite volume ratio 2:1 1:1

The tests were terminated after 5585 and 5545 hr, respectively, and results
of examination have been obtained for the longer-duration loop. A "scratch"
test of the molybdenum strips contained in loop 1250 showed that the surface of
the strip, where in direct contact with graphite, was significantly harder than
the surfaces in contact with either the salt or the INOR-8. Metallographic
examination of the strips showed a layer ~0.1l-mil thick on the surface exposed
to the graphite; otherwise, no apparent changes in microstructure were found
as a result of the exposure conditions, as shown in Figs. 4.4 and 4.5,

Tensile tests of the molybdenum strips in both the as-received and as-
tested condition showed gqualitatively that the ultimate strength and per cent
reduction in area of the molybdenum were reduced as a result of exposure.
However, there was no apparent difference in the hardness of the as-tested
and as-received molybdenum. Table 4.4 gives the results of the tensile tests
and hardness measurements. (Tensile results must be regarded as qualitative
because of the necessarily small specimen size).

Table 4.4, Mechanical Properties of Molybdenum
(Qualitative Only)

Ultimate Reduction Hardness —
Strength in Area Four Determinations
(psi) (%) (DPH)
As Received 102, 600 50 267
105,200 49,7

As Tested 79, 650 3.5 273

JUNCLASSIFIED
T-203014

Fig. 4.4. As-Received Appearance of Molybdenum Strips Contained in INOR-8 Thermal-Convection Loop
1250, Etchant: 1 part NH,OH, 1 part H,0,.

UNCLASSIFIED:
T-20302

Fig. 4.5. As-Tested Appearance of Molybdenum Strips Contained in INOR-8 Thermal-Convection Loop
1250. Etchant: 1 part NH,OH, 1 part H,0,.

Results of chemical analyses made on both the as-received and as-tested
molybdenum are shown in Table 4.5. An increase in the chromium concentration
of the molybdenum during test is clearly indicated.

99

Table 4.5. Results of Chemical Analyses of Molybdenum Strips

Constituent (ppm)

Sample Ni Cr Fe
As-received 50 210 L20
As-tested 250 1600 860

Metallographic examination of INOR-8 specimens removed from the loop reveal-
ed relatively heavy surface roughening and pitting; however, examination of the
as-received tubing revealed a similar surface condition. Examination of one of
the INOR-& spacers, which held the molybdenum in contact with the graphite,
revealed attack in the form of surface roughening and pitting to a maximum
depth of 1 mil along the inner surface, which was exposed to the salt. The
outer surfaces of the spacers, which were in intimate contact with both the
graphite and molybdenum, were found to have been carburized to a depth of >10
mils. Considering the difference in thermal expansion between INOR-8 and
graphite, it is likely that there was a relatively large contact pressure
between the graphite and INOR-8 along the carburized areas.

Chemical analysis made on the as-received and as-tested salt showed the
chromium content to have increased from an initial value of 550 ppm to a final
value of 960; no other changes were apparent in the salt analyses.

4.1.3 Compatibility of INOR-8 — 2% Nb and Molten Fluorides

As a consequence of recent developments to improve the weldability of
INOR-8,3 corrosion tests have been scheduled to evaluate the resistance to
fluoride attack of INOR-8 that has been modified by the addition of 2% Nb,
In addition to corrosion data on INOR-8 — 2% Nb, these loop tests will also
supply data on the corrosion properties of various types of weld junctions
listed below.

1. TINOR-8 welded with INOR-8 weld rod.
2. INOR-8 welded with INOR-8 — 2% Nb weld rod.
3. INOR-8 —~ 2% Nb welded with INOR-8 — 2% Nb weld rod.

Corrosion tests will be conducted in two INOR-8 thermal-convection loops,
each loop to contain two tubular inserts in the hot leg. One of the inserts
will be fabricated from INOR-8 modified with 2% Nb, while the other will be
of standard INOR-8. The inserts will be separated from each other, and the
loop by spacers. They will be protected from the atmosphere by an outer
jacket welded to the loop. The spacers will also function as the weld-test
specimens. Tentative operating condltions for both loops are as follows:

Salt composition LiF-BeF,-ZrF) -UF), - ThF), (70-23-5-1-1 mole %)
Maximum salt temp. 1300°CF

AT 170°F

Operating time One year

h,1.4 Fluoride-Salt Contamination Studies

Corrosion studies of fused fluoride mixtures have conventionally been
conducted under closely controlled test conditions to maintain a high degree
100

of salt purity. However, maintaining such stringent control becomes problemati-
cal in the case of a large-scale reactor experiment, and there is a probability
that small amounts of certain impurities, which act as strong oxidants, will be
encountered throughout normal reactor operation. Examples of these impurities
include moisture, dry air, oil, and combinations thereof. Accordingly, a test
program was started to systematically investigate the effects of such contami-
nants on the corrosion behavior of fused fluoride mixtures and to define the
limits of the various contaminants which can be safely allowed to enter the
salt.

The first series of tests was designed to establish the effect of
moisture on the corrosion of INOR-8 by fused-salt mixtures. It is known that
water immediately reacts with the salt to form HF and oxides of the melt
through hydrolysis; thus, this phase of the program is being divided into two
series of tests. In the first series of tests, salt mixtures containing
controlled amounts of HF will be circulated to study both the mechanism and
rates of attack as a function of HF concentration. In order to translate
these results into an understanding of the corrosive effect of H,0, a second
series of tests utilizing small static pots will be operated to determine the
rate of HF buildup in salts exposed to wet helium.

The initial thermal-convection test under this program was terminated
prematurely after 255 hr as a result of a sudden flow restriction. The test
circulated an NaF-ZrF, (50-50-mole %) salt mixture to which HF had been added
immediately before operation.

The salt mixture was a standard production batch obtained from the Reactor
Chemistry Division. Before beilng placed in the loop, the mixture was
repurified through the use of pure zirconium shavings and a hydrogen purge.
Radicactive FeF, was used to monitor the progress of the repurification.
Chemical analysis of the salt before and after repurification showed the
nickel and iron content to be reduced, whereas the chromium content increased.

After the repurification process, the loop was filled, and the HF was
introduced through a dip leg attached to a salt pot at the top of the loop.
The HF-bearing salt was clirculated through the loop under approximately a
550F AT. Determination of the amount of HF introduced to the salt was
rendered impossible because of several leaks which occurred in the gas lines.
However, chemical analysis of the salt circulated will provide some measure
of the amount of HF absorbed. Metallographic and chemical analyses are in
progress.

4.2 INOR-8 DEVELOPMENT

4.2.1 Structural Stability of Nickel-Based, 18% Molybdenum Alloys

Work is continuing in order to determine the solubility limit of chromium
plus iron in nickel-base alloys containing 18% Mo over the temperature range
900 to 2000°F. The purpose is to confirm that INOR-8 will not be subject to
the precipitation of intermetallic compounds when the maximum amount of alloy-
ing additions allowed by the specification are present, that is, 18% Mo —

8% Cr — 5% Fe. It is of further interest to determine the location of the
solid-solution phase boundary known to exist near the upper limit of the
INOR-8 compositional specification in order to establish what deviations from
specifications might be permitted.
101

The previous reportl'L summarized the results obtained from a study of the
900-t0~-2000°F microstructures of a series of 18% Mo alloys encompassing the
maximum compositional limit of INOR-8. The interpretation of these results
wvas complicated by the presence of a fine grain-boundary precipitate in the
alloys over the range 1100 to 1650°F. This precipitate, thought to be a
carbide, redissolved at 1800°F. Two alloys, both containing iron and
chromium contents in the range 8 to 10%, retained a precipitate at lBOOOF,
indicating that this phase was an intermetallic compound. Based on these
results, an 1800°F phase boundary was proposed, as shown in Fig. L.6.

UNCLASSIFIED
ORNL-~LR-DWG 63277

14 r
VT—154|| Ll
12 I
1
\
PREVIOUSLY \ NEW BOUNDARY
10 PROPOSED //“
BOUNDARY —1~ \
\
\
. VT—-156 \
< 8 [ Y@ VT-157
— \
= \
~ \
=
o '\\
x g i \-\
~
SINGLE - PHASE FIELD VT-158 . TWO-PHASE FIELD
. |
~
VT-159 Tw VT-160
4 - > ¢
~
~
~
~
Mo VT-162
2 .
0
0 2 4 6 8 10 12 14 16

CHROMIUM (wt %)

Fig. 4.6. 1800°F Phase Boundary for 18% Mo—Balance Ni Alloys Containing Varying Chromium and Iron
Contents.

Recent effort has been concerned with confirming the proposed phase
boundary (alloys VI-154 through VI-160, Fig. L4.6). Prior to heat treating
samples of these alloys within the range 900 to 2000°F, a decarburization
heat treatment was carried out at 22000F in dry hydrogen to eliminate as much
carbon impurity as possible from the materials. Metallographic examination
of samples receiving 100-hr heat treatments at 1800°F showed in the construction

102

of a new phase boundary, as shown in Fig. L4.6. Again it is thought that carbon
contamination presents a problem in so far as all the alloys generally showed

a small amount of fine grain-boundary precipitate in the range 1100 to 1650°F.
This precipitate dissolved at 1800CF in those alloys located in the single-
phase region of the diagram shown in Fig. L4.6. Particles cobserved at 18000F
in those alloys located in the two-phase region are thus believed to be an
intermetallic compound. According to this interpretation, the decarburization
treatment given the alloys was insufficient in lowering the carbon content
below the solubility limit at temperatures up to 165CCF even though subsequent
carbon analyses for the alloys were in the range of 0.005 to 0.01%.

4.2.2 Temperature Range of Melting for INOR-8

Apparatus has been set up to determine the temperature range of melting
for INOR-8. The purpose is twofold: (1) +to establish the melting range of
the alloy more accurately, and (2) to compare the melting range of the various
heats of the alloy with their respective nil ductility temperature as reported
by Rensselaer Polytechnic Institute. The latter purpose is related to the
incidence of base-metal cracking found in certain heats of INOR-8 when welded
under restraint. The particular heats involved were found (by means of the
TPI Hot-Ductility Apparatus) to exhibit inferior ductility behavior at high
temperatures; therefore, a comparison of the nil-ductility temperature with
the solidus temperature might establish whether the presence of liquid was
the cause of the poor elevated-temperature ductility. Differences in the
solidus temperatures between various heats of the alloy could be related to
varying amounts of trace impurities in the material.

The data obtained to date are presented in Table 4.6. Although no
definite conclusions are yet possible, it does appear that solidus tempera-
tures are somewhat higher than nil-ductility temperatures, presenting some
doubt that liquid phases are the cause of the elevated-temperature brittleness
reported.

Teble 4.6. Comparison of Nil-Ductility Temperature with
Melting-Temperature Range for Various Heats of INOR-8

Melting-Temperature Range

Nil-Ductility Liquidus Solidus
Heat Temperature Temperature Temperature
Nunber Vendor (°F) (o) (°F)
Y-8488 INCO 2300 2L62-2498
NI-5055 Haynes 22hs5 2hob-2Lhl
SP-19 Haynes 2L00 2L62-2498
SP-16 Haynes 2200
SP-20 Haynes 2350
SP-26 Haynes 2557—2561,
SP-37 Haynes 2453—2471.
DI-0L0OO Haymes 2h53-24 71
DI-OLO1l KJ Haynes 2453-2L71.
M-1 Westinghouse 2250
M-1566 Westinghouse 2400

M-1666 Westinghouse 2552

103

4.2.3 Specific Heat of INOR-&

The specific heat of annealed INOR-8 has been determined by direct
calorimetric measurements at the University of Tennessee. The value was
observed to increase uniformly from 0.409 joule/g, ©C at 60°C to 0.483
joule/g, ©C at 5L0OPC. At 550°C, an anomalous rise of about 20% in the
specific heat was observed, reaching a maximum value of 0.586 joule/g,

OC at 610°C. Above this temperature, the specific heat decreased to 0.578
joule/g, OC at 700°C. The phenomenon has been attributed to an ordering
reaction. The values for the specific heat are presznted in Table 4.7,

Table 4.7. Specific Heat of TINOR-8

Temperature c_(3oule/g, ©C) Temperature c (joule/g, ©C)

(°c) o (°c) P

60 0.4090 LLo 0.4730

&0 0.Lk152 LG0 0.47h9
100 0.4207 480 0.4766
120 0.4257 500 0.4781
140 0.4303 520 0.4794
160 0.4340 530 0.4805
160 0.4370 540 0.4833
200 0.4403 550 0.hkoz2
220 0.4439 560 0.5060
2Lo 0.L475 570 0.5226
260 0.4510 580 0.54202
280 0.4539 590 0.5637
300 0.4566 600 0.5808
320 0.4592 610 0.5858
340 0.4616 620 0.5848
360 0.46k41 640 0.5825
380 0.4666 660 0.5808
LOO 0.4688 680 0.5792
L20 0.4710 700 0.5778

4,3 TOTAL HEMISPHERICAL EMITTANCE OF INOR-O

The temperature dependence of the total hemispherical emittance of
platinum, matte and bright-finished INOR-8, and several oxidized INOR-8
specimens was measured between 100 and 700°C. The method employed an
instrumented thin-strip specimen heated electrically in a black-body
vacuum chamber which was held at a constant known temperatures A steady state,
a heat balance on the specimen yields the following:

VI

E =
Ac(TSh - T,

h) ’

104

where
E = total hemispherical emittance (t.h.e.),
A = surface area of the specimen between the voltage taps,
g = Stefan-Boltzmann constant,
T_= average specimen temperature, CK,

S
Te= average black-body vacuum-chamber temperature, CK,

VI = electrical power per unit length of the specimen.

Heat gains or losses considered in obtalning this equation are:
(1) the electrical power dissipated by the specimen, (2) the radiant energy
incident upon the specimen, and (3) the radiant energy emitted by the specimen.
Heat gains or losses which are neglected as being cumulatively less than 4% are:
(1) radiant energy emitted by the specimen that is reflected by the chamber walls
back to the specimen surfaces, (2) the heat loss by atmospheric convection and
conduction, (3) the heat loss by conduction through the specimen to the input
electrodes, and (4) the heat loss by conduction through the thermocouple leads.

The t.h.e. of platinum measured by this method serves as a guide to the
validity of the measurements on the INOR-8 specimens. The platinum data as seen
in Table 4.8 and Fig. 4.7 compare quite favorably with other literature values.
The authors' data for platinum are approximately 16% higher than data obtained
by a similar method (run II, Table 4.8) at 500°C. A large part of this percent-
age difference is because the platinum tested did not have a bright polish.

Table 4.8. Total Hemispherical Emittance at Selected Temperatures

Temperature
Specimen  Run No. Condition Prior to Run 100°c  200°c  300°c  Loo°c  500°C  600°Cc  700°C  B0O°C
INOR-8 No. 1 1 Matte finish 0.220 ©.210 0.214% 0.226 0.242 0.260 0.275 0.2852
2 1 hr at 8309 0.412 0.451 0.502 0.561 0.625
3 0.5 hr at T30°C plus 0.%417 0.468 0.528 0.579 0.607
five l-min cycles to 730°C
Iy 5 hr at 730vC G.436 0.496 0.570 0.622 0.6L5
5 5 hr at 730°C 0.465 0.478 0.513 0.567 0.605
INOR-8 No. 2 6 Bright Finish 0.2012 0.202 0.212 0.226 0.239 0.239 0.226
7 25 l-min cycles to 730°C 0.311 0.34% 0.398 o0.L46L 0,503
8 50 cycles to T30°C 0.31% 0.3k 0.4L18 0.497 0.525
9 75 cycles to T30°C 0.293 0.315 0.34% 0.389 0.433 0.509
10 100 cycles to T730°C 0.286% 0.318 0.353 0.394 0.his5 0.L88
11 1 hr at 830°¢ 0.349 0.39C 0.434 0.481 0.528
INOR-8 No. 3 12 Bright finish 0.194% 0.208 0.221 0.232 0.2L2 0.24s5
13 1 hr at 1000°C 0.502 0.523 0,630 0.704 0.757%
Platinum 1k Bright finish 0.127 0.109 0.106 0.110 0.119 0.129 ©0.1k2 (§.156
ks 0.071 0.087 0,102 0.116 0.130 0.1k4
II¢ 0.062 0.070 0.083 0.103 ©0.122 (©.lk2
rrrd 0.061 0.070 o0.080 0.088 0.103 0©.l22

aExtrapolated.

Phavisson and Weeks, J. Opt. Soc. Am. 8, 585 (1924).

°A. H. Sully et al., Brit. J. Appl. Phys. 3, 97 (1952)

9. D. Foote, N.E.S. Bull. 11, 607 (191k)

The data on three of the INOR-8 specimens reflect the importance of surface
finish. Runs 6 and 12 were on different specimens having a bright polish, and
the results are remarkably close. Run 1 was on a matte-finish INOR-8 and is
7.1% higher than runs 6 and 12 at 500°C.

105

UNCLASSIFIED
ORNL-LR-DWG 63278

0.8

TOTAL HEMISPHERICAL EMITTANCE

p—""9

1 O A e
0.2 12
6

b

-

14.F§-‘““*flb—— p

0.1 = & ——
O
C 1C0 200 300 400 500 600 700 800

TEMPERATURE (°C)

Fig. 4.7. Total Hemispherical Emittance vs Temperature for Platinum, Matte and Bright-Finished
INOR-8, and Several Oxidized INOR-8 Specimens. (See Table 4.8 for meaning of numbers.)

Following the initial runs, the matte-finished and bright-polished
specimens were given the treatments listed in Table 4.8. These treatments
resulted in an increase in the t.h.e. which was insensitive to thermal cycling
effects. The increase in t.h.e. was greatest for the 1000°C treatment and
least for the T730°C treatments.

Based on these measurements, i1f INOR-8 is given an oxidizing heat
treatment at 1000°C, its t.h.e. at 600°C will be 0.76 as compared to an
unoxidized surface value of 0.24; and at 5000C values of 0.705 and 0.225,
respectively, are to be expected. Oxidation at temperatures below 1000°C will
not yield as high a t.h.e.
106
4.4 WELDING AND BRAZING STUDIES

4.4,1 Welding of INOR-8

A program is under way to investigate and circumvent weld-cracking which
has been observed in some heats of INOR-8 material. Cracking and microfissuring
have been reported in test welds made with certain base-metal and filler-metal
heats.” Since the melting practice used by the material supplier in pouring the
initlal ingots appeared to influence the cracking behavior markedly, the major
effort has been directed at investigation of this wvariable.

Since different vendors use their own melting practice or practices,
quantities of l-in.-thick material were obtained from three vendors. 1In some
cases, the vendor furnished individual heats of material melted according to
different proprietary practices. Test welds were then made with these
materials under conditions of high restraint and were then carefully examined
for cracking by visual, dye-penetrant, and metallographic inspection.
Mechanical tests on these welds were conducted at room and elevated tempera-
tures, and, in addition, hot-ductility testing6 is being performed and
evaluated on a subcontract at Rensselaer Polytechnic Institute. Complete
chemical analyses on all heats used 1n the study are being obtained in an
attempt to correlate composition and melting practice with performance.

A discussion of the different phases of the study, together with the
results to date, are presented in the following paragraphs.

44,2 Metallographic Examination and Bend Tests on Welds

An extensive metallographic examination was performed on 14 weld-test
plates fabricated from 9 different heats of INOR-8 base metal from three
different vendors. When available, the filler wire used was of the same heat
of INOR-8 as the base plate. However, in several cases, filler wire of the
desired heat was not available, and, another heat was used in the interest of
expediency.

A compilation of the results of the metallographic examination of typical
test plates is presented in Table 4.9. The results of side-bend tests made on
these plates are also shown for comparative purposes. It will be noted that
those welds containing cracks and porosity were fabricated from INOR-8
produced with certain air-melting procedures. Heat-affected-zone cracking
and porosity typical of these welds are shown in Fig. 4.8 (a photomicrograph
of the fusion-line area of weld No. SP-37-Wl). A sound microstructure,
typical of welds of the improved heats of material, is that shown in Fig. 4.9,
the fusion-line area of a weld in heat Y-8488.

4.4.3 Transverse Tensile and Creep Tests

Transverse tensile and creep tests were performed on welds of seven of the
heats included in the program. The results of the room- and elevated-temperature
tensile tests are presented in Table 4.10, while creep-test results performed
at 1300°F at a stress of 27,500 psi are presented in Table 4.1l. From the
testing conducted thus far, it appears that the vacuum-melted heat, SP-OLO1,
and the air-melted heat, Y-8488, exhibit the best over-all weldability. The
air-melted heat, SP-5055, did not exhibit any cracking or porosity, but the
creep and tensile properties were generally somewhat lower.
Table 4.9.

Results of Metallographic Examination and Side-Bend Tests on INOR-8 Weldability Samples

Metallozraphic Results

S5ide Bend Results

Heat-Affected

Heat-Affected

Weld Base-Plate Base-Plate Filler-Metal Zone of Weld Zone of Weld
Plate No. Heat No. Melting Practice Heat No. Base Metal Metal Base Plate Metal
C = Cracked; P = Porous; 0.K. = No cracking or
Porosity
SP-19-W1 SP=-19 INOR-8 air-melting Haynes SP-19 C C C C
procedure (vendor A)
SP-20-W1 SP-20 INOR-8 air-melting Haynes SP-19 C C C C
procedure {vendor A)
SP-23-W1 SP-23 TNOR-8 air-melting Haynes SP-19 P,C C C C
procedure (vendor A)
SP-26-W1 SP-26 INOR-8 air-melting INCO Y-8LE8 P,C 0.K. 0.K. 0.K.
procedure (vendor A)
SP-37-Wl SP-37 INOR-8 air-melting Haynes SP-37 P,C C C C
procedure (vendor A)
SP-5055~W1 SP-5055 Hastelloy-W air- Haynes SP-0401 0.K 0.K. 0.K 0.K
melting procedure
(vendor A)
SP-0401-W2 SP-0LO1 Vacuum-melting Haynes SP-0401 0.K. 0.K. 0.K 0.K.
procedure (vendor A)
W-16T71-W1 W-1671 INOR-8 ajir-melting Haynes SP-19 0.K. 0.X. C C
procedure (vendor B)
Y-8488-W1 Y-8488 INOR-8 air-melting INCO Y-8488 0.K. 0.X. 0.K. 0.K.

procedure (vendor C)

LOL
108

Fig. 4.8. Heat-Affected-Zone Porosity and Cracking in an INOR-8 Butt Weld (No. SP-37-W1) Made Under
High Restraint. Etchant: CrO;-H,0-HCI.

Note that the cracking and porosity seen in the metallographic examination
are generally reflected in the transverse tensile tests by markedly reduced
ductilities (particularly at 1300CF) and by slightly lower tensile strengths.
The good tensile properties of the weld in SP-26 material are attributed to
the fact that the filler wire (heat Y-8488, see Table 4.9) possessed excellent
weldability.

In the creep tests, the weld porosity and microfissuring were reflected
in short rupture times. In most cases, lowered ductilities were also observed.
Note however that the weld in SP-26 material responded rather poorly in the
creep test, indicating that this test is really a more selective criterion
for true joint integrity.

L.4.4 Welding of INOR-8 to Stainless Steel and Inconel

The qualification of tests for tungsten-arc welding of INOR-8 to stainless
steel and Inconel has been completed. Procedure specifications for INOR-8-to-
stainless steel (PS-35) and INOR-8-to-Inconel (PS-33) have been prepared. As
a result of the findings of the weldability study being performed on INOR-8,
the existing specification for the welding of INOR-8 alloy pipe, plate, and
fittings is being revised. The revisions will incorporate into the welding
procedure the requirement that the interpass temperature does not exceed 200°F.
Experience with this type of material indicated that interpass-temperature

“UNCLASSIFIED
Y-42433
7 S

Fig. 4.9. Fusion-Line Area of Weld No. Y-8488-#1. No cracks or porosity is evident. Etchant: CrO,-
H,0-HCI.
2

control was desirable to ensure high joint integrity when welding this material
in thick sections under high restraint.

k.4.5 Remote Brazing

If removal of any MSRE primary components is required, provisions must be
made for remotely disconnecting and subsequently rejoining the necessary pipe-
work. For this reason, a study is being conducted in cooperation with the
Reactor Division Remote Maintenance Group to develop procedures for constructing
these remotely brazed joints.

The design of such a joint is influenced by both brazing and remote-
maintenance requirements. Maintenance procedures are simplified by use of
relatively loose fits and a minimum of remote machining, but optimum brazing
procedures require relatively tight fits and very clean surfaces. Accordingly,
a tentative joint design was adopted (see Figs. 2.26 and 2.27, Chap. 2). The
major features of this joint are the provisions for brazing alloy preplacement
and a radial clearance of 0.0025 to 0.0035 in. (0.005 to 0.007 in. in diameter).
This clearance is considered optimum in terms of ease of fitup and reliable
brazing-alloy flowability.

Preliminary work with the brazing of I]\IOR-B7 led to the selection of a
. brazing alloy composed of 82 Au — 18 Ni (wt %). The alloy is ductile and very
110

Table 4.10. TINOR-8 Transverse Tensile Tests
on Butt Welds in l-in. Plate

%
Test Tensile Yield Elongation
Weld- Temp. Strength Strength in Location
Plate No. (°F) (psi) (psi) 1-1/2 in. of Failure
SP-23-W1 Room 105, 500 67,800 28 Weld metal
1300 59, 700 43,000 8 Heat-affected
zone
SP-26-W1 Room 116,300 69, 800 L7 Weld metal
1300 70, 400 47,500 o2 Heat-affected
zone
SP-37-WL Room 107, 600 63, 200 b1 Weld metal
1300 54,900 41,800 7 Weld metal
SP-5055-W1 Room 118, 200 68, 900 39 Weld metal
1300 73,200 49, 600 17 Base metal
SP-O4O1-w2 Roon 115,900 67, 600 43 Base metal
1300 71, 400 48, 600 18 Base metal
W-1671-W1 Room 11k, 400 75,600 31 Weld metal
1300 67,000 53, 600 10 Weld metal
Y-8L488-wl Room 117,000 67,200 5h Weld metal
1300 69, 400 L4, 300 22 Weld metal

Table 4.11. INOR-8 Transverse Creep Tests
on Butt Welds in l-in. Plate

Tests conducted at l300°F at 27,500-psi stress

Rupture Total

Time Strain
Weld Plate No. (hr) (%)
SP-23-Wl 6 1.0
SP-26-W1 41 9.0
SP-27-Wl 13 ~0,2
SP-5055-W1 127 L.5
SP-OU01-W2 493 25.0
W~1671-W1 19 k.0

Y-8488-wW1 370 15.0

111

resistant to both molten fluoride-sglt corrosion and high-temperature oxidation.
Miller-Peaslee shear-test specimens® of INOR-8 brazed with this alloy exhibit
good shear strengths at MSRE-operating temperatures, as shown in Table 4.12.
Joints were brazed in the flat position in both inert- and dry-hydrogen
atmospheres with gaps ranging from contact to 0.010 in. Testing at 13C0°F in
air showed the joints to have a minimum shear strength of 12,500 psi and an
average of 18,100 psi. The joints brazed in helium exhibited slightly higher
strengths than did equivalent hydrogen-brazed Jjoints, and attempts are being
made to determine the reason. Additional shear-test specimens have been
brazed and will be tested to provide room-temperature shear-strength data as
well as data on specimens aged at 1300°F for 1000 and 10,000 hr.

Table 4.12. Miller-Peaslee Shear-Strength Data for Brazed Joints

Base metal: INOR-8
Brazing alloy: 82 Au — 18 Wi (wt %)
Brazing temperature: 1830°F
Brazing time: 10 min
Testing temperature: 1300CF
Testing atmosphere: Air
Average Shear
Brazing Gap Brazing Number of Strength
(mils) Atmosphere Specimens (psi)
0 Hy 3 16, 800
0 He 2 23,100
1.5 Hy 3 19, 800
1.5 He 2 19, 800
2 Hy 3 15,300
2 He 2 17, 400
L Ho 3 18, 200
Ly He 2 20, 400
6 Ho 3 16,100
6 He 2 19,000
10 Ho, 2 15,700
10 He 2 17,900

Long-time diffusion studies of INOR-8 lap joints brazed with the gold-
nickel alloy have also been conducted. Specimens were aged from 1000 tec 10,000
hr at both 1200°F and 1500°F in air. The maximum amount of diffusion of
brazing alloy into the base metal, as detected by microhardness measurements
was only about 2 mils at 1200°F and 7 mils at 1500°F. This minor amount of
diffusion is expected to have little effect on joint strength and over-all
base-metal properties.

To date, four Jjoints of the type shown in Figs. 2,26 and 2.27 have been
brazed in the laboratory to determine the quality of joint. Ultrasonic
inspection of the Jjoints indicated a few unbonded areas ranging from very
small scattered voids to occasional regions up to one-fourth the length of

112

the joint. This Jjoint was greatly overdesigned to accommodate occasional
defects and, no doubt, would be adequate in a molten-salt system. Never-
theless, a slightly tapered joint design is being investigated; it is
expected to eliminate nearly all nonbonded areas and to make remote assembly
somewhat easier.

4.5 MECHANICAL PROPERTIES OF INOR-8

A program has been started to determine the strain-fatigue behavior of
INOR-8, and, to date, four tests, two at 1300 and two at 1500°F, were performed
on rod specimens of heat SP-25. These test results are shown in Fig. 4.10
with a plot of Coffin's equation,

1
F] 1l -
Nf ep = g_ef
vhere Ne is the number of cycles to failure, €, 1s the plastic strain range and

€r is the true fracture ductility. The tensile ductility at Ny = 1/4 was used
as the intercept for these curves.

UNCLASSIFIED
ORNL-LR-DWG 63279

100 —
11
® HEAT SP-25: 1300°F
O HEAT SP-25; t500°F

B | - . ! 40 cycles per hour
- L1 R
I I

S : |
NN |
i “h !,

4.7
77

I
I
1
Ll
I
+
:
|
L
|
4
|
! I/IH

€5, TRUE PLASTIC-STRAIN RANGE (%)

.

o

|

| ‘ A
0.1 1 10 100
N, ,CYCLES TO FAILURE

Fig. 4.10. Strain Fatigue Curves for INOR-8.

Further strain-fatigue tests, both mechanical and thermal, on tubular
specimens from another heat of material are in progress.

A number of tensile tests have been performed on specimens subjected to
creep loading for various times. No deleterious effect of prior creep was
observed in the tensile properties.

113

Creep tests on a weak, low-ductility heat material, M-1566, indicate that
soaking at 1600 and 1800°F for 2 hr increased the rupture life and ductility.

4,6 TIMPREGNATION OF GRAPHITE BY MOLTEN SALTS

The use of unclad graphite in the molten-salt reactor requires that a
minirmum of the graphite pore space be filled with salt. A test program is in
progress to study the impregnation of different grades and sizes of graphite by
LiF-BeF,-ThF), -UF), (67-18.5-14-0.5 mole %) at 1300°F for various times and
pressures.

Three additional grades of graphites failed to meet the low salt-impregna-
tion requirements in the standard screening test in which the exposure was for
100 hr under a salt pressure of 150 psig. However, the tests indicated that
increasing the "as-fabricated" diameter of a rod of graphite or fabricating it
in the shape of a pipe made it more "open" to the molten salt. The results of
the tests are sumarized in Table 4.13.

Table 4.13. Impregnation of Various Grades of Graphite
by LiF-BeF,-ThF)-UF), (67-18.5-14-0.5 mole %)

Test conditions: Temperature: 1300°F
Pressure: 150 psig
Time: 100 hr
Nominal dimensions of test specimens: 0.5 in. in diam and

1.5 in. long
Grade of Nominal Dimensions Bulk Density Bulk Volume
Graphite of Graphite, As- of Graphite of Graphite
Fabricated (g/cc) Impregnated
(in.) by Salt
(%)2
Rods (diameter only)
GB-3.6 3-5/8 1.91 3.8
GP-2 2 1.87 h.1
GP-3.4 3-3/8 1.88 k.9
B-1-B-1 1-1/4 1.8k 6.2
B-1-B-2 2 1.76 8.5
B-1-B-3 3-1/2 1.70 13.3
Tube
B-1-B-4 3-1/16 0D x 2 ID x 3 long 1.67 16.4

%gach value is an average of three.

114

Grades GB and GP were extruded and impregnated types of graphite. Grade
B-1-B was the base stock for a low-permeability graphite, grade B-1l. (Base
stock as used here means a graphitized graphite shape that has not had its
open pore spaces filled and/or sealed by a series of impregnations with carbon-
base materials and graphitizing treatments to make it into a low-permeability
graphite.) Most of the rod-shaped graphite was fabricated with relatively
large diameters.

L,7 AMMONTIUM BIFLUORIDE AS AN OXYGEN-PURGING AGENT FOR GRAPHITE

Tests were run to determine the effects of temperature on removing oxygen
and oxide contamination from graphite by the thermal decomposition products of
NH,F«HF. Purging temperatures varied from 1300 to 930°F, but times were held
constant at 20 hr. INOR-8 specimens were also included in these test runs to
observe the effect of the decomposition products on this alloy., Test conditions
are shown in Table 4,1k,

Table L.1Lk. Effect of Temperature on Purging Oxygen and Oxide Contamination from
Graphite by Means of the Thermal Decomposition Products of NHhF-HF

INOR-8 Reaction Time Exposed to

Test System Purging Conditions During Purging b

Grade of ) .a Weight Payer Salt W1?hout )
romite vewl (S T Peswe (ple) G Memess 00 Preeiptiacing
AGOT 1300 20 86 22 2000

AGOT 1110 20 83 58 2000

AGOT 930 20 €0 ho 2000
R-0025 INCR-8 1300 20 90 35 +0.137 <0.5 2000

R-0025 INOR-8 1110 20 80 46 +0.130 1000

R-0025 INCR-8 930 20 70 Ly +0.002 o¢ 2000

aPressures at the purging temperature.

bLiF-Bng-UFh (62-37-1 mole %).

CReaction layer too thin to measure.

4.7.1 Removal of Oxygen from Graphite

The LiF-BeF,-UF) (62-37-1 mole %) salt is a sensitive oxygen detector
and in the presence of oxygen at 1300°F readily precipitates part of its
uranium as UO,. This reaction is easily detected using radiography. Therefore,
to determine %he effectiveness of the purging of graphite crucibles with the
thermal decomposition product of NHhF-HF, the purged crucibles were charged
with the LiF-BeF -UFh salt and heated to 1300°F. At intervals of time the
crucibles containing the salt were radiographically monitored for a UO
precipitate. Results shown in Table L.1l4 indicated that purging at tempera-
tures of 930, 1110, and 1300°F was equally effective in removing oxygen from
AGOT and R-0025 graphites. All the purged graphite crucibles contained the
LiF-BeFe-UFh salt at 1300°F for more than 2000 hr without U'O2 being detected.

115

There has been no indication, at this time, that the different porosities of
AGOT and R-0025 graphites affected the purging action. It is expected that
porosity might be a factor in a more impermeable grade of graphite.

4,7.2 Effects of the Thermal Decomposition
Products of NH)F-HF on INOR-&

To secure additional data on the reaction of the thermal decomposition
products of the NH),F-HF with INOR-8,9 0.040-1in.-thick sheet-type tensile
specimens were included in the purging test using grade R-0025 graphite. A
layer was formed on the INOR-8 at all purging temperatures. This layer is
being identified. INOR-8 reaction data observed as weight changes are
summarized in Table 4.1hk. The thickest of the reaction layers apparently did
not seriously harm the tensile properties of the INOR-8. These data are
shown in Table L4.15.

Table 4.15. Tension Test Results® on INOR-8 Treated
with the Thermal Decomposition Products
of NHMF-HF for 20 hr at 1300°F

Yield Strength

Extension (psi,
(%) 2% Offset)
At Room Temperature
Controls® 43.5 45,000
Treated specimens 38.5 45,600
At 1250°F
Controlsb 19.2 33, 400
Treated specimens £2.3 33, 600

%Fach value is an average of three. All specimens were
nominally 0.040O-in, thick.

bControl specimens had the same thermal history as the
treated specimens but were exposed to an atmosphere of
pure argon.

One-hundred-hour accelerated corrosion tests at 1832°F in the LiF-Belo-
ZrF) - ThF),~UF (70-23-5-1-1 mole %) salt were made on control and exposed
(with reaction layers) specimens from the 1300°F purge to determine the
effects of this layer. The reaction layer was removed by the salt but
there was no difference in corrosion attack (primarily intergranular to a
depth of 6 mils) between the control specimens and those with reaction
layers. More detailed studies will be made to determine the effect of the
reaction layer on INOR-8 corrosion resistance to molten fluorides.
116

REFERENCES

MSRP Prog. Rep. Feb. 28, 1961, ORNI-3122, p Ll-L6.

Ibid., p 81.

Ibid., p 81-84.

Ibid., p 89.

J. C. Richmond and W. N. Harrison, "Equipment and Procedures for

Evaluation of Total Hemispherical Emittance,"” J. Am. Ceram. Soc., 39
(11), 668-T73 (1960). -

E., F. Nippes et al., "An Investigation of the Hot Ductility of High-
Temperature Alloys," Welding J. 34 (L), 183-85 (1955).

MSRP Quar. Prog. Rep. July 31, 1960, ORNI~301L4, p 63-6k.

R. L. Peaslee and F. M. Miller, Proposed Industry Wide Standard Procedure
for Testing the Shear Strength of Brazed Joints, Wall Colmonoy Corp.,
Detroit 3, Michigan.

MSRP Prog. Rep. Feb. 28, 1961, ORNI~3122, p 98.

5. IN-PILE TESTS

5.1 GRAPHITE-FUEL CAPSULE EXPERIMENTS

A molten-salt-fuel capsule irradiation experiment, ORNL-MIR-4T7-3, has been
operated at the Materials Test Reactor and is now undergoing postirradiation
examination. The experiment was designed to determine the effect of irradiation
on: (1) graphite at or in excess of MSRE temperatures and heating conditions,
(2) the wetting characteristics of the fuel with graphite at high temperature,
and (3) the compatibility of the principal MSRE structural materials with the
molten salt.

The experiment contained four capsules with fuel [UF4(93%UZ°%)-ThF4-ZrF,-
BeFo-Li7F (1.4-1.2-5.2-23,2-69.0 mole %)] in graphite boats, as described
previously.l The graphite and fuel were encapsulated in cans of INOR-8 which
were irmersed in molten sodium for convenience in heat removal and temperature
measurement, Two capsules contained AGOT graphite impregnated prior to irradi-
| ation with fuel salt, and two contained R-0025 graphite, unimpregnated. Also,
included in each capsule were specimens of molybdenum, pyrolytic carbon, and
INOR-8 in contact with the fuel.

The temperature history of the capsules during the irradiation period, May 5
to July 24, is summarized in Table 5.1. The over-all time-averaged temperatures

Table 5.1. Temperature History of Graphite-Salt
Interfaces in ORNL-MIR-4T-3 Capsules

Type of Sample

. Temperature Unimpregnated Impregnated Impregnated
Interval Graphite Graphite Graphite
(°c) (15, 16)% (8)2 (3)%
i} Time at Steady Temperature (hr)
0-50 318 318 318
50-750 33 33 27.5
750-800 803 803 L
800-850 702 702 0.5
850-900 601
900-950 89l
950-1000 11
Total: 1856 1856 1856
Time at Non-Steady Temperature (hr)
56 56 56
Total Irradiation Time
1594 159k 1594

aCapsule identification number.

117

118

of the graphite-to-salt interfaces were about T90°C for the capsules with unimpreg-
nated R-0025 graphite and 810 and 900°C for the capsules with impregnated AGOT
graphite. The difference between the temperatures of the two impregnated capsules
corresponds to the differences between the local ambient temperatures in the
sodium. These temperatures conservatively are above the design temperature of
1300°F proposed for the MSRE reactor. They are, however, below the proposed tem-
perature for the capsules.! This is attributable largely to the fact that the
impregnated graphite contained only 9 g of fuel instead of the 17 g assumed in

the design. It also is probable that the changes made in the MIR core, namely,
the removal of ANP experiments and the relocation of control rods, reduced the
thermal neutron flux at the HB-3 beam hole and also contributed to the lower
temperatures.

The steady-state temperatures of the capsules were constant within *10°C.
The accumilated time during non-steady state operation, listed in Table 5.1,
includes only the transient times for temperature changes of 30°C or more.

The experiment was removed from the MIR, partially disassembled, and is now
at Battelle Memorial Institute, where further dismantling and postirradiation
examination will be conducted. As an aid to the interpretation of the irradi-
ation experiments, out-of-plile experiments have been initiated by the Reactor
Chemistry Division to duplicate the temperature history, including transients,
on control sample capsules.

REFERENCE

1. MSRP Prog. Rep. Feb. 28, 1961, ORNL-3122 (June 20, 1961).

6. CHEMISTRY
6.1 PHASE-EQUILIBRIUM STUDIES

6.1.1 Systems Involving NaF with ZrF4, ThF4, and UFg
and. Other Constituents

During the past four years the phase relationship for one after another of
the systems limiting the quaternary systems LiF-BeF,-ThF4-UF4 and NalF'-BeFs-ThPg-
UF4 have been completely explored and described. Systems involving LiF have had
priority because of their direct bearing on fuel, blanket, and coolant materials
for the MSRE, and the results have been summarized previously.l For comparison,
and because of the possibility that NaF-based materials will be useful in the
future, a parallel investigation of the phase relationships in the binary and
ternary systems limiting the quaternary system NaF-BeF,-ThF4-UF4 has been pur-
sued; diagrams of these limiting systems have been reported® % as they were
completed, and the last one, for NaF-ThF4-UF4, is described below. DBefore the
inclusion of 7ZrF4 in the MSRE fuel was found necessary, only preliminary infor-
mation was available regarding the systems LiF-BeFs-ZrFs and NaF-BeFo-7rFg4;
brief studies have been made for evaluating the usefulness of NaF-BeF, and LiF-
BeF- mixtures as solvents for zirconium alloy fuel elements in the Fluoride
Volatility Process.

Current investigations involve the systems LiF-BeFs-ZrF4, NaF-BeFo-iZrFg,
and composition~temperature studies of several complicated mixtures whose compo-
sitions are similar to the MSRE fuel, LiF-BeFo-ZrFy-ThFs-UFy (70-23-5-1-1 mole %).

6.1.2 The System NaF-ThF,~UFg4

With but few exceptions, melts containing ThF,a, UF4, and NaF crystallize on
cooling to form complex compounds in which the Th4* and ust positions are to some
extent interchangeable. In all but two composition areas of the ternary, as
shown in Fig. 6.1, the final products of crystallization are a mixture of solid
solutions containing both ThF4 and UF4. Mixtures of NaF-ThF,-UF4 containing more
than 66.7 mole % NaF produce pure NaF among the final products of equilibrium
crystallization; solid solutions of iNaF:ThF4 and 3NaF.UF4 are not stable equi-
librium phases at low temperatures and decompose to yield NaF. The only Nal-ThF,
or NaF-UF, complex compound in which at least partial interchangeability of UFs4
and ThF4 does not occur is in NaF-ThF4. Continuous solid solutions occur across
the composition section ThF4-UFs and in the composition-temperature range in which
3NaF.2ThF4--5NaF+3UF4 and f-4NaF«ThFs--f-3NaF-UF4 solid solutions are stable.
Extensive limited solid solutions are formed along the constant NaF compositions
33.3, 46.2, 66.7, 75, and 80 mole %. Of these sections, limited solubility is
observed between the compound pairs NaF.2ThF4--NaF:2UF4, TNaF-.6ThF4--TNaF.6UF,,
oNaF «ThF4--2NaF.UFy, in @-3NaF.UFs, and in Q-4NaF.ThF,.

The principal difference between the systems NaF-ThF4-UF4 and LiF-ThF4-UF4
in respect to segregation of uranium- and thorium-bearing phases is that 1n the
NaF system limited solid solutions are the rule, while in the LiF-based system
they are the exception. Solubility of a small amount of either ThF4 or UF4 in

119
120

UNCLASSIFIED
ORNL-LR-DWG 45512R

ThE,
1

@ NaF
a—4NaF - Thfy ss
(C) B-4NaF -Thiy — §-3NaF-Ufg ss
(D) a-3NaF-UFgss
(E) Ba-2NoF - Thigss
® 8 -2NaF - UFsss
(G) 3NoF - 2ThF, -5NoF - 3UF, s
(H) a-NoF-Thr,
(1) 7NoF -6ThF4-7Naf - 6UF4 55
@ NaF - 2ThFsss
® NaF - 2UF4ss COMPOSITION IN mole Y
L) ThE-Ufgss TEMPERATURE IN °C

100

NoF-2Thf,

834

NaF-ThFy

3NaF-2Thfy” 742

£630~
m.p. 705

ZNGF-ThE" ‘ &O
o, /
B
£ 618
m. p. 651
4NaF-ThF,-
4 g @
£632° \\
710
)
) O\ o 706} (e
N 3 Wss N ©)] 692
) \
o) %b@oo & @ ® ® A)Oo
= [@)
% @ \\ N7 e 2 \
) N \
- X KON ALV .

995 £618 £l Fe4B P 673 m.p. 78 £ 680 NaF-2UF,

m.p. 629 2NoF-UF" SNoF-3UF"  7NaF-6UR,
BNGF'UE‘

Fig. 6.1. The System NaF-ThF,-UF,.

NaF-~-Thl'4-UF4 solld-sclution phases containing a much larger amount of the other
tetrafluoride is practically complete. Solid-phase segregation in mixtures con-
taining nearly equal amounts of ThF, and UF4 yields solid solutions moderately
richer in one of the tetrafluorides, though in such processes the residual
liquids do not become appreciably more concentrated in either ThFs4 or UF4 than
the original composition.

The composition sections UF4-ThF$, 2NaF.ThF4--2NaF-UF,;, and 4NaF.ThF,--
3NaF.ThF, were reported previously.s" The composition sections TNaF.6UF4- -
TNaF+6ThF, and NaF.2ThF4--NaF«2UF4 are shown in Figs. 6.2 and 6.3, respectively.
The temperatures and compositions of invariant equilibria and singular points
in the system NaF-ThF4-UF, are given in Table 6.1.

TEMPERATURE (°C)

121

UNCLASSIFIED
ORNL-LR-DWG 64964

750
[ |

11+41:2 ss+ LIQUID

1:2 ss + LIQUID
_J 1:2 ss+(7:6—3:2) ss + LIQUID
Mmoo\ - /- 1:2 ss+7:6 55+ LIQUID i
e — 7:6 ss + LIQUID
[ I —\\ A S [
o / ! 114+4{(7:6-3:2)ss + LIQUID
>
(e}
] i
?20 — + - i T P P —
:/ 7:6 ss+ LIQUID ‘

710 -
M'G*}Z)SS'F !

f LIQuID
700

1
: 11+(3:2-7:6)ss + LIQUID
“:\‘h—(B:Z—?:G)ss+(7:6—3:2)ss+1:1

|

690 i ————— ]
|
| 1:2 = NaF - 2ThF, s
: 7:6 = 7NaF - 6ThFy— 7NaF - 6UF4 55

680 I 3:2=23NaF - 2ThF;ss Lo -
l 1:1 = a—NgF - Thf
[
|
|
I 1

670 L '

0 2 4 o 8 10 12 14

Uk, {mole %)

Fig. 6.2. Composition Section 7NaF-6UF ,—7NaF.6ThF , (Partial).
122

UNCLASSIFIED
ORNL-LR-DWG 61554

950 T

\ LIQUID

900 - ‘ ——

850 | R R .
83 1 ThF4-UF4 ss + LIQUID
: ~ THF4 ~UF, ss + NaF -+ 2ThFg4 ss+
/7 LiQuip |
S ]

NaF -2UF4 ss +
Thfg-UFg s5 +

LIQuo —_

800 |-

750 | P

ThFg—UFy ss+
7NaF- 6 (Th,U)F, +

uoum\ L—

TEMPERATURE (°C)

|
700 - ---1»——— g | x!
| |
NaoF - 2ThF, ss + I :
NoF-2UFg ss— | | ThFy~UFg 55+
| T t 7NaF - 6(Th, U) Fq
650 [ | : ! =
| ! NoF - 2UF4 55+ \|
! | 7NoF - 6(Th, U) g+ )
| : I| ThF, - UF, ss
|
600 e et -
' [
: L NeF-2UFg ss
| |
| |
550 : .
0 10 20 30 40 50 60
UF, (mole 7o)

Fig. 6.3. Composition Section NaF.2ThF ,~NaF.2UF,.

6.1.3 Stannous Fluoride as a Component of Reactor Fuels

Low-melting nonvolatile fluorides such as SnFp and PbF, have sometimes been
proposed as possible additives to lower the melting point of fuels. Unfortunately
they are too corrosive to contain in anything except graphite and possibly molyb-
denum. For concentrations of UFs4 or ThFs over 10 mole %, the use of LiF-BeF; as
a solvent gives a lower melting mixture than found for the SnF, - quadrivalent
fluoride binary even though pure SnFp melts at only 213°C,°B
Table 6.1. Invariant and Singular Points in the System NaF-ThF4-UF4
Composition (mole %) Temp. Type of X
Nab Thi s UFs (°c) Equilibrium S50l1lid Phases Present
5.5 10.5 1k 588 Eutectic B-UNaF«ThF4--B-3NaF«UFy ss, Bo-2NaF-ThFy ss,
6-2NaF.UFy4 ss
66.7 23.3 10 688 Maximum on boundary Bo-2NaF.ThFy ss, 3-2NaF.UFy4 ss
curve
64 27 9 655 Futectic Bo-2NaF+ThF, ss, ©0-2NaF-UF4 ss,
5NaF «3UF4-3NaF.2ThF4 ss
6k 3k 12 665 Maximim on boundary 3NaF.2ThF4-5NaF«3UFs ss, 0-2NaF:UFy4 ss
curve
59 37 y T12 Maximun on boundary 3NaF.2ThF4-5NaF.30Fs ss, TNaF.6ThFg-
curve TNaF.6ThF, ss
58.5 Lo 1. TO3 Eutectic 3NaF«2ThFs-5NaF-3UFy ss, O-NaF-ThFg,
'TNaF-6UF4-TNaF-6ThF4 55
5k.5 k3.5 2 732 Peritectic 0-NaF.ThF4, NaF-2ThFs ss, TNaF:6UFg-
TNaF.6ThF, ss
54 41.5 4 T34 Maximum on boundary TNaF:6UF4-TNaF-6ThFy4 ss, NaF.2ThFy ss
curve
48 12 L0 T06 Peritectic TNaF«6UF4-TNaF.6ThFy ss, NaF.2ThF4 ss,
NeF.2UF4 ss
b7 11 Lo 692 Peritectic TNaF.6UF 4 -TNaF«6ThFy ss, NaF«2UF4 ss,
ThiF'g-UFs4 sSs
46 1k Lo 710 Peritectic NaF.2ThFs ss, NaF«2UF4 ss, ThF4-UFy4 ss

£zl
124 >

6.2 OXIDE BEHAVIOR IN FUELS

6.2.1 Behavior of MSRE Fuel on Freezing and Effect
of Segregation on Oxide Precipitation

Much effort has been devoted over a period of years to locating the lowest-
melting regions in phase diagrams of salts suitable for use in nuclear reactors.
But in practice, the lowest-melting compositions, which characteristically
solidify without change in composition of the liquid during the freezing process,
are rarely suitable for reactor use; somewhat altered compositions which retain
the desirable feature of a relatively low melting point are used. As a conse-
quence, fuels of practical interest are prone to segregate on freezing. 1In
phase-study terminology an important aspect of the segregation behavior is re-
ferred to as the crystallization path of the fuel. Complex mixtures such as the
MSRE fuel can follow numerous crystallization paths depending on the conditions
of freezing. Slow cooling, which favors the equilibrium crystallization path,
can greatly enhance segregation into strata of different crystalline compounds.
A fast rate of cooling, besides giving rise to some different crystalline com- .
pounds and to altered proportions of those that form in any case, can often
virtually eliminate stratification.

Freeze valves, which are frozen plugs in salt lines to stop flow or to allow
flow on thawing, have given generally satisfactory results in engineering tests.
The presence of a frozen zone does, however, provide an opportunity for segre-
gation and for the "cold trapping" of high melting or insoluble precipitates.
Consequently attention has been paid to gaining a better understanding of the
crystallization path of the fuel during freeze valve operation,

To provide an example of the worst case (from the standpoint of segregation),
a vertically aligned freeze valve® was slowly cooled to shut off a very slow down-
ward flow of the MSRE fuel (LiF-BeFp-ZrF4-ThF,-UFy; T0-23-5-1-1 mole %). When
sectioned and examined, a sharply segregated layer of 6LiF-BeFs-ZrF; (the equi-
librium primary phase) and 2LiF:BeF, (the equilibrium secondary phase) was
clearly marked and visible, due to the absence of green color associated with UFg.

The following sequence of events was surmised to have occurred during -
freezing.
The primary and secondary phases began to precipitate on the coolest metal -

and formed a layer which grew inward and upward, finally forming a complete arch
or dome which stopped the flow. TFurther growth of this layer depleted the ZrFa4
content of the overlying stagnant melt, and as the temperature fell to the point
that all the liquid in the frozen zone solidified, the last liquid to freeze,
found as a mixture of crystals too fine to be readily identifiable, was enriched
in UF4. In this portion, depleted in ZrF4 and enriched in UF4, some UO, was
found. Evidently the fuel originally contained a small amount of oxide which
was held in solution by complexing with Zr4+ ion, but which formed a precipitate
of U0z when the ZrF4-rich phases were frozen out.

Several days after the frozen plug had formed, finely divided BeO was added
to the still-molten fuel in the upper tank for the purpose of forming ZrO- in
order to check on the extent of cold trapping of Zr0s in the cold zone in the
vicinity of the frozen plug. A precipitate of ZrQOp formed, as expected, but it
appeared to have accumulated predominantly on the walls in the upper parts of
the system in crystals up to 20 p in size, rather than growing preferentially in
the cold zone of the freeze valve. Had the system been allowed to age for a long
time, presumably more ZrOz would have been cold trapped, but under the conditions
of this experiment no indications of a plug of ZrQ0s in the cold zone were found.

125

The main conclusion reached was that segregation in a stagnant fuel can
cause localized depletion of ZrF4 with resultant precipitation of small amounts

A contrasting and more typical case of ZrOp precipitation, without the for-
mation of UO,, was afforded by a melt which became contaminated with oxide from
insufficiently pure helium used as a protective atmosphere. 1In this case Zr0Os
accurulated in a layer at the bottom of the melt, and no UOp was found. Crystals
of ZrOs that grow in the MSRE fuel at 600 to T0O0°C are usually 20 to 30 p in
size, while those which are formed during freezing, by precipitation of soluble
oxide, are only 1 or 2 p in maximum diameter.

An illustration of an interesting deviation from the equilibrium crystal-
lization path of the MSRE has repeatedly been encountered. Slowly-cooled melts
correspond to the equilibrium crystallization path in which 6LiF.BeFs-ZrF4 forms
at 442°C, followed by LizBeF4 at 435°C. However, with cooling rates of 2 or 3°C
per minute or faster, the nucleation of 6LiF:BeFz+ZrF4 frequently does not occur,
and the supercooled melt yields both B~3LiF.ZrF4 and well-formed crystals of
2LiF.ZrF4, the result of metastable crystallization to phases which are not
present in the equilibrium sequence. Temperature-time curves for one example of
rapid freezing showed supercooling to 435°C followed by crystallization at 438°C,
perhaps of B-3LiF:ZrF4, and then another phase, presumably 2LiF:7ZrF4, at 431°C
after supercooling to 421°C. 1In all cases the major phase present is LigBeFq4,
which in the rapidly cooled samples appears as crystallites too small for recog-
nition by petrographic microscopy but which can, however, be identified by
x-ray diffraction. In the LiF-ZrF4 binary system, B-3LiF*ZrF4 has a lower
limit of stability at 4TO°C; thus its appearance in the MSRE fuel is even more
surprising.

The formation of a frozen zone, if carried out slowly enough,would produce
a plug of the primary phase throughout the region maintained below the liquidus
temperature. In actual practice, the zone is formed quickly enough that a
mixture of phases is deposited, and no extensive region of segregated high-melting
crystals occurs. In the temperature gradient leading away from the coldest
portion, however, there will be a contour corresponding to the liquidus tempera-
ture of the MSRE fuel., Here there should be a crust of the primary phase,
6LiF+BeFs+ZrFs. Fortunately, this crust has apparently always been thin enough
to dissolve rapidly on raising the temperature, and sluggish responses attribut-
able to high-melting phases, even in other compositions, have not been noted.

6.2,2 0Oxide Content of Fluoride Melts

When an attempt was made to use oxide analyses in the 1l00- to 1lOOC-ppm
range on a routine basis in several extensive experiments on fuels, a greater-
than-expected variance in the oxide content among replicate samples was en-
countered. The analysis of trace amounts of oxides has long been recognized as
exceedingly dif‘ficulti but a method based on a fluorination procedure had shown
considerable promise.® In view, however, of the frequent occurrence of appar-
ently irreconcilable results, a statistical survey was carried out in a search
for clues to the source of the difficulty.

Analyses of T5 replicate samples from a single composite source, or standard
mixture, were accumulated for statistical treatment to determine the differences
attributable to two different but supposedly identical assemblies, each involving
three fluorination cells as employed by two different operators. Temporal
factors were also included as a parameter. A mean value of 500 ppm oxide was
obtained, and 56 of the T5 analyses were distributed between 300 and 750 ppm.

No significant source of variance was attributable to any of the nonprocedural

126

factors which were tested. With this point estabiished, attention is currently
being turned to procedural problems.

When an atterpt was made to calibrate the analytical method on UzOg, which
provides a relatively large oxygen content, gocd results were obtained. The
results for eighteen samples ranged from 1h4.6 to 15.6% oxide with an average of
15.10%, whereas the theoretical content was 15.18%h. The effect of using larger
samples of salt will be studied next even though there are disadvantages associ-
ated with the use of large samples. Also, the influence of particle size, which
may affect both sampling and the completeness of the reaction, is to be investi-
gated in experiments making use of a factorial design and sequential analysis.

Consideration 1s being given to the feasibility of an alternative method
based on the isotopic dilution of 0'® introduced for equilibration as Hgol8
vapor in a cover gas containing HF.

6.3 PHYSICAL AND CHEMICAI PROPERTIES

6.3.1 Surface Tension Apparatus for Molten Fluorides

Apparatus has been designed, the major components constructed, and assembly
nearly completed, for the measurement of the surface tension of fused fluoride
mixtures by the bubble pressure method. Surface tensions have been measured for
very few fluoride mixtures; the measurements are expected to throw light on
changes in the structure and bonding with changes in the composition of molten
salt mixtures. The surface tension of breeder fuels is also one of the important
factors determining the wetting or permeation of graphite by fuel, a matter of
considerable interest for the MSRE. With the recently constructed apparatus a
search for strongly surface active additives in pure fluorides and in LiF-BeFo-
based melts will be made, although no examples are known in the literature on
molten salts. In conjunction with the concurrent work on the permeation of
graphite, the surface tension results will permit estimation of the individual
contributions of surface tension and contact angle effects. (The combined effect
is measured in permeation tests,)

The bubble pressure method was chosen for the surface tension measurements
because it is generally accepted to be the most accurate and most generally
applicable method for molten salts. Since many fluorides are sensitive to
moisture and air, the bubble tube and crucible of salt were enclosed in a gas-
tight flanged nickel reactor which can be evacuated and filled with a thoroughly
dried inert atmosphere, Provision was made for checking the water content of
the inert gas blanket and the bubble gas with a Meeco electrolytic water analyzer,
Other principal innovations include the use of a special controllable leak valve
(used on the mass spectrometer) for regulating the bubble flow, and the use of
a cistern-~type dibutyl phthalate manometer constructed from precision~bore tubing
in both arms. The procedure and apparatus for meking each measurement needed
for the surface tension determination were designed to yleld results as accurate
as any in the literature of molten salts.

6.3.2 Volatilization of Iodine from the MSRE Fuel
Consideration of proposed experiments to ascertain the extent of liberation

of fission product iodine from the MSRE fuel has led to a review of the factors
controlling the release.

127

As predicted previously,ll elemental iodine is not expected in the MSRE off-
gas in appreciable amounts because of the reducing action of chromium in the
INOR-8 used to contain the fuel; the detailed mechanism involves the agency of
a small concentration of UF3 in equilibrium with the chromium and UF4. The vapor
pressure of iodide salts from the fuel is expected to be below 10" mm of Hg at
reactor temperature,l?

In case of a spill involving contact of the salt with air or water there
will be some release of iodine activity. Oxygen in air can displace iodide ions
in the fuel to produce elemental iodine, and water hydrolyzes iodide salts to
give volatile HI, which also reacts with oxygen. Both direct oxidation and
hydrolysis come to virtual cessation as soon as the fuel freezes, but it is
difficult to predict the extent of reaction prior to solidification. Assuming
that the spill is quenched by flooding with water, hydrolysis to produce HI might
be expected to proceed at roughly the same rate per ion of halide as hydrolysis
to produce HF, Hence if less than 10% of the fluoride leaves as HF, which should
cover most cases, then a release of 10% or less of the iodine content could be
predicted.

6.4t GRAPHITE COMPATIBILITY

6.4.1 Effect of Cesium Vapor on Graphite

In the MSRE some of the xenon produced by fission is expected to diffuse
into the graphite moderator and there be converted into cesium by decay or by
neutron capture. Cesium is known to intercalate with graphite.l Changes pro-
duced in the graphite surface by the cesium might conceivably change the wetting
behavior, To evaluate this, graphite treated with cesium, along with untreated
specimens as controls were immersed in molten salt and compared.

Assuming the use of graphite with a DXe of 1075 cme/sec, it was calculatedl®
that in one year of operation of the MSRE at 10 Mw, 80 g of cesium would be de-
posited in the graphite, nearly all of it near the 1.4 x 10®° cm® of surface,
corresponding to an average of 0.057 g of cesium per 1000 cm® of graphite.

Three graphite (AGOT) cylinders, 12.77 rm in diam and 25.3 mm long, weighing
5.4 g each, were outgassed at 200°C by pumping a vacuum for 3-1/4 hr, and then
treated with 26.5 liters (measured at room temperature) of helium saturated with
cesium at 200°C, According to the vapor pressure data of Kubaschewski and
Evans,l5 0.032 g of cesium was exposed to the graphite. This is about 1.25 g of
cesium per 1000 em® of graphite.

These treated specimens, and three similar graphite specimens that had
merely been outgassed, were kept immersed in molten MSRE fuel at 650°C for 41 hr.
No wetting appeared to have occurred and none of the specimens was changed in
appearance by the experiment. Those without the cesium treatment lost 0.002 g
each, probably due to outgassing. The specimen nearest the cesium inlet gained
0.012 g. The other treated specimens lost 0.001 g each.

It was concluded that the exposure to cesium vapor obtained in this experi-
ment was too mild to alter the graphite detectibly.

128

6.5 FUEL PRODUCTION

6.5.1 Purification Treatments

Continued improvements on processing methods have been made at the facility
for the production of purified fluoride mixtures. Current procedure for fluoride
mixtures which melt below 800°C uses flowing hydrogen and anhydrous hydrogen
fluoride to achieve purification with a minimum of corrosion of the equipment.
During the period of this report approximately 885 kg of mixtures comprised of
LiF, BeFs, ZrF4, ThF4, and UF4 were processed for the molten-salt reactor program.
An additional 2000 kg of the mixture NaF-LiF-ZrFs (37.5-37.5-25.0 mole %) were
processed for the Chemical Technology Division.

The current production rate for purifying molten fluorides is approximately
1.3 £° per week for each of two units. This value is derived from a single-

shift operation during a normal five-day week. However, by employing a multi- *
shift operation and a seven-day week this production rate can probably be
increased to approximately 1.9 £ per week for each unit. -

The ant1c1pated operation of the MSRE will require the preparation of
approXimately 155 £+° of fluoride mixtures for the flush, fuel, and coolant media
of the reactor. From considerations of methods for transportlng these purified
fluoride mixtures to the reactor, the use of some of about 50 existing cans, now
in storage and filled with NaF-ZrF4-UF4 melts, appears feasible. To be suitable,
these containers should be reconditioned to remove residual salts from previous
usage. However, the costs incurred by this operation can be partially offset by
incorporating a modification to these vessels which will increase their capacity.
By lengthening each shipping container by approximately 12 in., the full capacity
of the purification treatment vessels can be utilized. This modification would
effectively 1ncrease the maximum weekly production rate of fluoride mixtures to
approximately 3.5 £t2 for each of two operational units. With these proposed
alterations the fluoride mixtures required by the MSRE could be processed during
a minimum operating time of approximately six months. Eight months of lead time
should allow for contingencies and unforeseen difficulties.

6.6 ANALYTICAL CHEMISTRY

6.6.1 Analyses of MSRE Fuel Mixture
Thorium
The evaluation of the proposed amperometric method® for the determination
of thorium in the MSRE fuel mixture was completed. This method is particularly

applicable to remote operations with radiocactive fuel mixtures.

Zirconium

An indirect method has been developed for estimating the zirconium content
of fuels; it is based on a method involving the determination of total thorium
and zirconium. In this method, an excess of ethylenediaminetetraacetic acid
(EDTA) is added to an aliquot of the solution containing both thorium and zir-
conium. After the addition of sodium fluoride, the acidity of the solution is
adjusted to pH b with sodium acetate, and the excess EDTA 1s back-titrated with -

129

a solution of vanadyl sulfate. The equivalence point of the titration is de-
tected amperometrically wilth a platinum electrode at an applied potential of

0.6 v vs S.C.E. At this potential, the addition of excess vanadyl sulfate re-
sults in an increased rate of current flow. Calculations based on this titration
yield the sum of thorium and zirconium. Since the thorium content is determined
independently, as mentioned above, the zirconium content can be estimated by
difference.

nggen

The equipment which is proposed for use in the determination of oxygen in
concentrations of the order of parts per million in highly radioactive fuel was
assembled and cursory tests were initiated. The number of such tests which have
been completed are not sufficient to permit a proper evaluation of the equipment
at this time.

6.6.2 Analyses of MSRE Cover Gas

Tests, which have been described previously,'® were carried out to determine
the effectiveness of titanium sponge as a "getter" for the removal of oxygen and
nitrogen from helium, the MSRE cover gas., Preheated helium that was contaminated
with oxygen and nitrogen at levels from 0.0l to 2% was passed over beds (10- and
20-in. deep) of titanium metal sponge in a heated quartz tube (30-mm ID). In
these tests, the flow rates were varied from about 100 cc/min to 4 liters/min,
and the temperature of the titanium was varied from 600 to 1200°F.

In order to increase the sensitivity of the chromatographic method for the
determination of oxygen and nitrogen in helium, a method was devised in which
these contaminants are concentrated from a 1/2-liter sample of the gas onto a
bed of molecular sieves at the temperature of liquid nitrogen. In this method,
the exit gas, from the purification apparatus, flowing at a rate of 100 cc/min,
is passed for 5 min through a tube (1/4-in. OD and 1 ft long) that is packed with
thoroughly dried, Linde, type 5A, molecular sieves. During this operation, the
trap assembly is submerged in a liquid-nitrogen bath. At the end of the 5-min
period, the molecular-sieve trap is inserted into a stream of specially purified,
helium carrier gas by means of a six-way valve, after which the liquid-nitrogen
bath is removed and replaced with warm water. The oxygen and nitrogen which are
trapped quantitatively from the sample are released immediately thereby and then
routed through a molecular sieve colummn (1/4-in. OD and 4 ft long) in a Greenbrier
model-050 chromatograph. The concentrations of the resolved gases in the eluate
from the chromatographic column are determined in the conventional manner by com-
paring the thermal conductivity of the eluate wlth that of the pure carrier-gas
stream. Calibrations were performed by carrying out the same operations on a
sample of the inlet gas to the purification train. The fraction of a contaminant
passing through the purification train is thus given by the ratio of the area of
a contaminant peak in the exit-gas chromatogram to the area of the corresponding
peak in the inlet-gas chromatogram. :

Concentrations of nitrogen as low as 0.1 ppm can be measured easily by this
method. The sensitivity can obviously be increased by trapping the contaminants
for periods longer than 5 min. On the basis of repeated calibration experiments,
it has been found that the precision (relative standard deviation) of such
measurements is about 3% for concentrations of the order of 100 ppm. The sensi-
tivity of the method for the determination of oxygen actually exceeds that for
the determination of nitrogen; however, the measurement of oxygen 1s complicated
by the presence of argon, which accompanies the oxygen when separations are per-
formed on colurms of molecular sieves at room temperature. Since the oxygen,
from which contaminated helium samples have been prepared, contains almost 1%

130

of argon, the exact measurement of oxygen becomes uncertain when more than 9%

of the oxygen is removed by the purification process. At present, the measure-
ments of oxygen are corrected by subtracting the contribution of argon on the
measured oxygen-argon peak. This method is not, however, completely satisfactory,
since under some conditions only a few tenths of a percent of the original oxygen
escapes capture in the titanium purifier. Future experiments will be carried

out with contaminated helium samples prepared from argon-free oxygen; the oxygen
will be generated by the thermal decomposition of lead dioxide.

REFERENCES

1. C. F. Weaver, R. E, Thoma, H., Insley, and H. A. Friedman, Phase Equilibria
in Molten Salt Breeder Reactor Fuels, ORNL-2896 (Dec. 1960).

2, R. E. Thoma (ed.), Phase Diagrams of Nuclear Reactor Materials, ORNL-2548
(Nov. 1959), p 110.

3. MSRP Prog. Rep. Feb. 28, 1961, ORNL-3122, p 111.

4k, C. F. Weaver et al., J. Am. Ceram. Soc. Lk, 146 (1961).

. C. F. Weaver et al., J. Am, Ceram. Soc. 43, 213 (1960).

. MSRP Quar. Prog. Rep. July 31, 1959, ORNL-2799, p T5.

>
6
7. MSRP Quar. Prog. Rep. Apr. 30, 1960, ORNL-2973, p 67.
8

B. J. Sturm, Stannous Fluoride as a Component of Molten-5alt Reactor Fuels,
ORNL CF-60-10-133 (October 1960).

9. MSRP Prog. Rep. Feb. 28, 1961, ORNL-3122, p 27.

10. MSRP Prog. Rep. Feb, 28, 1961, ORNL-3122, p 132.

11. F. F. Blankenship, B. J. Sturm, and R. F. Newton, Predictions Concerning
Volatilization of Free Todine from the MSRE, MSR-60-4 (Sept. 29, 1960).

12, B. J. Sturm, Calculated Vapor Pressure of Iodide from MSRE Fuel, MSR-60-6
(Oct. 11, 1960).

13. W. Rudorf, "Graphite Intercalation Compounds," in Advances in Organic
Chemistry and Radiochemistry, (ed. by H. J. Emeleus and A. G. Sharpe)
vol I, pp 22L-66, Academic Press, New York (1959).

14, I. Spiewsk, Cesium Distribution in MSRE Graphite, MSR-60-40 (Nov. 23, 1960).

15. 0. Kubaschewskl and E. Evans, Metallurgical Thermochemistry, 2nd ed., Wiley,
New York (1956).

16. MSRP Prog. Rep. Feb. 28, 1961, ORNL-3122, pp 130-3%.

7. ENGINEERING RESEARCH

7.1 PHYSICAL-PROPERTY MEASUREMENTS

The determination of the enthalpy of the coolant mixture LiF-BeFz (68-32
mole %) was completed. The data, shown in Fig. 7.1, can be represented by the
following equations:

() solid (50 to 360°C),

H, - H30 = -6.22 + 0.225 t + (k.24 x 10~%) tZ ; (1)

(b) Liquid (480 to 8620°C),

H, - H,. = 123.90 + 0.185 t + (2.97 x 10-*%) t% . (2)

The enthalpy, H, is in cal/g for t in °C. As seen in Fig. 7.1, the solid-
liquid transition was not sharply defined and extended over the broad span of
360 to LB0°C; there are suggestions of two-phase transformations in this region.
The heat of fusion, evaluated between 360 and 480°C, was 151.h4 cal/g.

UNCLASSIFIED
ORNL-LR-DWG 63280

500 |
i "/‘./irwn/,
400 fr——f— b o -
5 k
S - el
- 3 OO Y IS — e == - L - . . . | —_— . p—
N rg \
.| : I
<l .
T
'_
= |
. 200 - — — e —
(@] : :
m |
X | |
| ‘ ‘
T | _
100 + —
0
0 100 200 300 400 500 800 700 800 900

TEMPERATURE (°C)

Fig. 7.1. Enthalpy of LiF-BeF, (68-32 mole %) Mixture.

131

132

The mean value of the heat capacity of the liquid between 550 and 800°C
[derived from Eq. (2)] is 0.586 cal/g-°C. The value predicted by using the
previously described empirical correlation,

e, = 4.2l /w0110 (3)

is 0.550 cal/g. Based on the experimental value of c , the agreement is to
within 6.5%; since Eq. (3) was developed from data on  LiF-BeF2 mixtures with
UF4 and ThF4 additives, the agreement between the experimental and predicted
values for this heavy-metal-~free composition is felt to be reasonable.

In order to further clarify the heat-balance discrepancy noted in the
heat-transfer studies with the LiF-BeFz-ThF4-UFs (67.0-18.5-14.0-0.5 mole %)
mixture,2 the enthalpy of the liquid was redetermined, using a sample of c¢ir-
culated salt. The results are given in Table 7.1, showing the comparison with

Table 7.1. Comparison of Enthalpy Measurements for the Mixture
LiF-BeF2~-ThF4-UFs (67.0-18.5-14,0-0.5 mole %)

Sample Correlating Eguation
A Hy - Hyy = ~61.5 + 0.493 t - (11.71 x 107°) t2
B Hy - Hyy = -81.1 + 0.548 t - (15.79 x 107°) t%
Circulated Hy - Hyy = -17.1 + 0.35h ¢ - (1.37 x 1075) ¢

the correlating equations developed from (1) an earlier measurement on an un-
exposed sample [A] and (2) a redetermination using this same salt {B]. while
the variation in the magnitude of the enthalpy curves for the three samples is
a maximum of 3.5% around the mean value, the differences in slope and curvature
are sufficient to cause a discrepancy of 10% in the heat capacities at the ex-
tremes of the temperature range. This deviation {shown graphically in Fig.
7.2) may result from composition differences in the samples, as discussed pre-

UNCLASSIFIED
ORNL-LR-DWG 63281

300 0.40
%)
275 °  0.38 <
o CIRCULATED . =z ~ UNCIRCULATED
> | :>/£fi’ 5 4N A, By
S 250 | # - o 2 036 %fi}( - —_ -
2 ///‘fiu - N
o 3 E N
2 s R B« S o034 b L CIRCULATED ____|
= prad ) s 7
= / \ o N
w ? M,
= 200 —- b e e T E 0.32 e S
x UNCIRCULATED T N
i - -._.
o 175 L o 0.30 I i
f ~
‘ |
150 . \ 0.28
500 600 700 800 500 600 700 800
TEMPERATURE (°C) TEMPERATURE (°C)

Fig. 7.2. Comparison of Enthalpies and Heat Capacities of LiF-BeF ,-UF ,-ThF, (67-18.5-0.5-14 mole %)
Mixtures.
133

viously;a sample analyses are reproduced in Table 7.2. The mean value for all
three samples over the range 550 to 800°C is 0.335 cal/g-°C; this exact agree-
ment in the mean values suggests that the observed variations in c,, may lie
within the experimental precision associated with these measurements. This
aspect of the enthalpy program will be investigated further. Average values
estimated by using Eq. (3) agree within 7% with the experimental mean (sample A,
0.347 cal/g-°C; circulated, 0.312 cal/g-°C).

Table 7.2. Composition of LiF-BeFz2-UF4-ThF4 Salt Mixture Samples

Constituent (mole %)

Sample
LiF BeFz UF 4 ThF 4
Nominal 67.0 18.5 0.5 14.0
A 70.3 15.4 0.6 13.8
Uncirculated 70.8 16.6 0.5 12.1
Circulated” 66.8 15.8 0.6 16.8

aAverage of three samples taken from loop sump.

7.2 HEAT-TRANSFER STUDILES

Interpretation of the data obtained in the study of heat transfer with
the LiF-BeFz-UF4-ThF4 (67-18.5-0.5-14 mole %) salt mixture has been continued,
with primary emphasis on the resolution of the abnormal heat balances observed.Z
Since the convective heat gain by the fluid is calculated from the equation

Qp = W Cp At (4)

the experimental measurements for the flow rate (w), heat capacity (cp), and
fluid temperature rise {at) must all be re-examined.

The fluid-flow rate was determined by use of a turbine-type flowmeter for
which the primary data were displayed as the rate of turbine rotation. This
was converted to 1b/hr of salt flow, using a water-flow calibratlion corrected
for the specific-gravity difference between water and the salt. Since experi-
mental data on the density of the LiF-BeFz-ThF4-UFs (67-18.5-14-0.5 mole %)
mixture were not available, recourse was had to an empirical correlation devel-
oped from measurements on other BeFz-containing mixtures. This yielded the
relation (based on the nominal composition of the salt mixture):

o =3.9% - 0.00087 t (5)

where p is in g/cm8 for t in °C. However, as noted from Table 7.2, there is a
great deal of uncertainty as to the actual composition of the salt circulated

in the loop; this has a significant effect on the predicted density. A compari-
son of estimated densities (at 1399°F) is given in Table 7.3; also shown are
values obtained by using a stochastic correlation developed from data on the

134

Table 7.3. Estimated Densities. for an
LiF-BeF2-ThF4-UF4 Salt Mixture

Com ositionb
Correlation D
Nominal Uncirculated Circulated
0, g/cm’
Empirical 3.30 3.15 3.61
Stochastic 3.49 3.38 3.73

®Run MF-2, t = 1399°F.
bAs given in Table 7.2.

molar volumes of LiF-BeFz mixtures.® For a typical run (MB-2), the heat bal-
ance ratio, /qe (where 9. i1s the measured electrical energy input), varies
from 1.05, using the empirically derived density value for the uncirculated
salt, to 1.24, with the stochastic estimate of the circulated-salt density.
An empirical determination of the density of this salt mixture is planned.

A second possible error relating to the rate of salt flow derives from a
lack of information as to the influence of fluid viscosity on the output of the
turbine meter. SGince preliminary inquiries suggest that this may be signifi-
cant, a study is in progress using liquids of wvarious viscosities [water, eth-
ylene glycol, and possibly a noncorrosive salt such as NaNOz2-NaNO3-KNOs (L40-7-
53 wt %)] to calibrate a turbine-type flowmeter.,

A further uncertainty in the convective heat gain relates to the discrep-
ancy in heat capacities as measured with circulated and uncirculated salt sam-
ples (see Table 7.1). Thus (considering run MB-2, with an average salt temper-
ature of 1408°C in the INOR-O6 section), the heat-capacity ratio (circulated to
uncirculated) is 1.06. As noted in the discussion above on the enthalpy of the
LiF-BeFa-UF4-ThFs (67-18.5-0.5-14 mole %) mixture, the influence of this uncer-
tainty in ¢, is & minimum near 1247°F (values are equal at this temperature).
Again, furtger measurements on the heat capacity are indicated.

A re-examination of the calibrations for the thermocouples used in obtain-
ing the fluid mean temperatures shows that, while a systemic error exists in
the emf output, the readings from the thermocouple pairs (inlet and outlet) are
sufficiently close to disallow any effect on the value of Qp-

REFERENCES

1. MSRP Quar. Prog. Rep. July 31, 1960, ORNL-301k, p 85.
MSRP Prog. Rep. Feb. 28, 1961, ORNL-3122, pp 1l40-42.
3‘ Ibid-j pp 123-250

8. FUEL PROCESSING

8.1 PREPARATION AND ANALYSIS OF COMPLEX FLUORIDES OF
SbFs WITH KF, AgF, AND SrFs

The complex fluoride salts of SbFs with KF, AgF, and SrF» were prepared by
dissolving the KF, AgF, or SrFs in anhydrous hydrogen fluoride and adding a small
excess of SbFs liquid. In each case a preclpitate formed, showing that the com-
plex is less soluble than the simple fluoride. The KF-S5bFg complex is soluble
in anhydrous hydrogen fluoride at 20°C to 150 mg per gram of solution, or 34 mg
of KF per gram of solution. Qualitatively, the AgF-SbFg complex is rather ine
soluble, probably less than 1 mg per gram of solution, and the SrFz-SbFg salt is
intermediate in solubility.

The compounds were purified by recrystallization from anhydrous hydrogen
fluoride after two anhydrous HF washes of the original precipitate. The KI-SbFg
compound was so soluble that a rather concentrated solution was formed during
evaporation of the excess HF, and solids tended to form a surface film; good
single crystals were not obtained., The SrFo-SbFg compound formed crystals on
the sides of the bottle, above the liquid phase. The AgF-SbFs compound formed
granular brown crystals of reasonably good quality for crystal structure
determination.

Analyses showed that the KF salt was KSbFg, with K/Sb mole ratios of 1.07
and 1.14 in two preparations and F/(K + 5Sb) ratios of 1.07 and 1.00. Analyses
for the other two salts were not entirely consistent. The AgF-S5bFs compound was
probably AgSbFg since the Ag/Sb ratio was 1.10, but the total of Ag, Sb, and F
added up to llO%, probably because of a high F determination. For the S5rFz-SbFg
compound the constituents added up to only 8T% (compared with 96 to 98% gener-
ally). The Sr/Sb mole ratio was 0.71, and the F/(2Sr + 5Sb) ratio was 1.5; the
compound apparently contained considerable HF.

8.2 PREPARATION AND ANALYSIS OF THE NaF-MoFg COMPLEX

The NaF-MoFg complex was prepared by mixing solutions of NaF and MoFg, both
in anhydrous HF. There was little if any heat of reaction, and no precipitate
formed. The solution was boiled to dryness, both HF and MoFg being evolved, and
the boiling point increased slowly to about 60°C before any solids appeared.

The salt was further dried by heating to about 80°C under a stream of helium,
The resulting solid had an Na/Mo ratio of 5.15 and an F/(Na + 6Mo) ratio of 1.05,
in agreement within analytical error with the formula MoFg-5NalF. Later prepa-
rations were de-solvated by pumping off the excess HF solvent with a vacuum pump
for several hours at a little above room temperature. These samples generally
had Na/Mo ratios of 30 or greater, indicating that the pumping removed most of
the MoFg. There was an indication that the solids contained 0.8 to 0.9 mole of
HF per mole NaF. Apparently MoFg and HF compete for the NaF, with HF probably
being complexed more strongly at these temperatures.

135

136

8.3 PREPARATION AND ANALYSIS OF COMPLEXES OF UFg
WITH LiF, NaF, AND KF

The preparation of Ufg complexes with LiF, NaF, and KF was attempted in a
manner similar to that for the MoFg complexes. All were quite soluble in HF,
and the solids obtained upon evaporation of the excess HF were yellow for LiF
and NaF and orange for KF. The LiF preparation had an Li/U mole ratio of 14 and
contained 0.6 mole of HF per mole of LiF. Three NaF-UFg preparations, with suc-
cessively more vigorous treatment to remove HF, had Na/U ratios of 10, 17, and
39, and Na/HF ratios of 1 to 2. Apparently UFg was removed rather than HF by the
de-solvation treatment, The KF complex had a K/U mole ratio of 8.5 and contained
2 moles of HF per mole of KF. Thus, all the compounds contained less UFg than
expected, and there was some indication that UFg was removed by the procedures
used to remove excess HF solvent., It appears that HF and UFg are complexed com-
petitively by the alkali metal fluorides and that all the complexes contain some
HF. They are therefore different from the complexes formed by reaction of UFg

gas with solid NaF. It appears, further, that the HF is held more tightly than 3
UFg, at least at room temperature. This suggests the possibility of separating
some of the volatile fluorides by displacement, for example, by displacing UFg -

from an NaF-UFg bed with HF, This would require a knowledge of the vapor pressure
of all these compounds as a function of temperature and composition.

8.4 ATTEMPT TO SEPARATE RARE EARTH FLUORIDES FROM
MSBR BLANKET SALT BY SbFs IN HF

Analytical results from one experiment in which SbFs in HF was used to
extract rare earths from an MSBR blanket salt indicate that the rare earths were
not separated from the ThF4. The salt (LiF-BeFo-ThF4-UF4, 67-18.5-14-0.5 mole %
containing 0.05 mole % rare earth fluorides and trace fission products) was
first contacted twice with 95% HF - 5% water to dissolve the LiF and BeF». The
remaining salt, containing the ThF4, UF4, and rare earths, was then contacted
twice with HF containing 5 vol % SbFs. This solvent dissolved very little of
anything and, in particular, did not dissolve a significant portion of the rare
earths. However, it should be pointed out that the analytical results were not
consistent with the known composition of the salt, so a large error could be
present. These results suggest that the rare earths may be present in the salt
in the form of a compound, probably with LiF, which is insoluble in HF-~SbFs
solutions, at least under the conditions of this test. -

AUGUST 34, 196!
R. B. BRIGGS, DIRECTOR o
P MOLTEN SALT ADYANCED DEYELOPMENT
REACTOR EXP ERIMENT
A. L. BOCH, PROJECT ENGINEER R
~
ENGINEERING RESEARCH COMPONENT DEVELOPMENT FUEL REPROCESSING METALLURGY REACTOR CHEMISTRY
H. W. HOF FMAN R I. SPIEWAK* R D. E. FERGUSON cT A. TABDADA® " W, R. GRIMES RC
R. G. DONNELLY " H. F. MCDUFFIE* RC
T. K. ROCHE " F. F. BLANKENSHIP® RC
C. M. BLOOD RC
J. E. EQRGAN RC
G. M. HEBERT RC
W. F. SCHAFFER RC
R. E. THOMA RC
B. F.HITCH RC
W. M. JOHNSON RC
DESIGN METALLURGY REACTOR CHEMISTRY RADIATION TESTING COMPONENT DEVELOPMENT CONSTRUCTION OPERATIONS
E. S BETTIS R A. TABOADA* M H. F. MCDUFFIE* RC 1 A.CONLIN R I SPIEWAK® R W. B. MCDONALD R 5. E. BEALL
%. H. Coox M F. F. BLANKENSHIP* RC C. A. BRANDON R D. SCOTT, JR. R C. K. MCGLOTHL AN R
R. G. GILLILAND M F. A. DOSS RC B. H. WEBSTER R
W. ). LEONARD M 5. 5 KIRSLIS RC
R.C. SCHULZE M B. 5. WEAVER RC
C. H. WODTKE M W. K. R. FINNELL RC
3. L. GRIFFITH M ¥ JENNINGS RC
M. 4. REDDEN M R. G. WILEY RC
‘DUAL CAPACITY
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137

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