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UNGLASSIFIED i
MARIETT A ENERGY 5 STEMS LIBRARIES

SRR w260 o 47

3 445 03klzap 2 UC81- Reactors—~

   
 

MOLTEN-SALT REACTOR PROGRAM

  
 

- QUARTERLY PROGRESS REPORT

 

FOR PERIOD ENDING OCTOBER 31, 1957

und; .. fl\é hbrary

OAK RIDGE NATIONAL I.ABORATORY
OPERATED BY
UNION CARBIDE NUCLEAR COMPANY

Division of Union Carbide Corporation

POST OFFICE BOX X * OAK RIDGE, TENNESSEE

UNCLASSIFIED
 

 

 

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Washington 25, D. C.

 

 

 

 

 

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UNGLASSIFIED

ORNL-2431
UC-81 =~ Reactors~Power

Contract No. W-7405-eng-26

MOLTEN-SALT REACTOR PROGRAM
QUARTERLY PROGRESS REPORT

For Period Ending October 31, 1957

H. G. MacPherson, Program Director

DATE ISSUED

FEB7 1958

 

OAK RIDGE NATIONAL LABORATORY
Operated by
UNION CARBIDE NUCLEAR COMPANY
Division of Union Carbide Corporation
Post Office Box X
Ock Ridge, Tennessee

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fipi>>.’£flmi&—)>§-ITI>CJS—EI"%'HE—EOOSOU?UmOflme'nCImF"

M.
F.
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M.

. G. Alexander

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UC-81 — Reactors—Power

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UNGLASSIFIED

CONTENTS

FOREWORD ... ettt et ate she s e st sttt sa e e e ssen et e e b e e ea b esbenbeereberaeatessenterentes
SUMMARY Lttt b et ettt e a et e e ke e e se et e s s ne e b e et et ea b e s e ete e e e e e ebe st etesrenes

PART 1. REACTOR DESIGN STUDIES
NUCL EAR CALCULATIONS. ..ottt sttt sttt sttt s e s et s e e st ese s
CAMMA HEATING OF CORE VESSEL .. et se e et ves e ene et b e
HEAT TRANSFER SYSTEMS o ettt s st e as et e v n s eeeba e ees

METALLURGY oottt et sttt st st e eass et e e s e te st etabesatasabesa e b besartan e naen et easesasen saenesrerens

Dl Gt O oo ee ettt s ettt e e ettt eeeeeeeeeeetaee e e e eaeeeeaeae ae s e taee——eae e t—te .t taertaae ar————ttttr ., atanaanrans
Mt eria] PO CUIEMONt ..o i ittt ettt e e e e e e e e s easeaee s e e e es e aeaaensrarssnsaaaeseseen

b Cation OF TN R -8 o et e eete e ee e e v e e e e e e eessanseesesresse s aseeemsas eesaeasessessessseeesassnnnseeses TR

AN S FOTMOTION KM TICS toviieereeretseeeeeereereeseerrereeesesseasesessasessessssseseeessansesasssseseassatassesenssasaseesas e,
Welding and Brazing. ..ottt et et e en et e eaeer s eeseeneas

DY NAMIC COFOSION Lottt vttt ettt et e e eas et e e e s e b st eae et b e st eb e baat s eeseaaterasaabbeseeneeosansaebstaarensan

T @St Program et st et et as e et a e st e e e s s etk e e et rae e ereaaeareeree
Thermal-Convection Loop Tests ittt cte et s te s e sbe s neste b e s sbeseresenna s
Forced-Circulation Loop Tests ..ottt e eaera et e e et st e s es b e ssreran sasensasenas
RADIATION DAMAGE .ottt ettt e et e et e st ase e e eve ke e ee s saeeeases e beassseantesesaseasaeseannese s besnn sneesnessans
[N=P il LO0D T @SS ceiereriiiieiiiieeie ettt cte e e e vt e et et b e sttt e s a et s s ses e ra bt est e bt eaaebesanabeenasre s st nrananeobensenens

S1atic IN-Pile Capsules ...ttt et bt sre e st aan e anesne e

CHEMISTRY oottt ettt e et e sttt et ebe s vt a s eatees e b et et oSk etec b e s en b ek e e b e bt e et enmecssabans

Phase Equilibrium StUdies ..o..oo i e
SOIVENT SYSTEMS woovitititeie et bbb bbb s b s e st
Systems Containing UF .o s
Systems Containing ThF ;..o s
Systems Containing Plutonium Fluorides ...

Chemistry of the Corrosion ProCess ...ttt

FisSion-Product BehaVior ...ttt st ere et drecen b e e sen s e e bt s b
Solubility of the Noble Gases ...t
Solubility of Rare-Earth Fluorides ..o

Production of PuUrified Molten Salts . iiie it iieserae e ste et e e e e e rar e e s ase et te s e rtessssneeeasaaes saransssnnes
Methods for Purification of Molten Sl et vctrcreessesresessresessesteeaeerarsssesss srsseesatntaneansessnenas

UNCLASSIFIED
UNCLASSIFIED

FOREWORD

This quarterly progress report of the Molten-Salt Reactor Program records the techni-
cal progress of the research at the Laboratory, under its Contract W-7405-eng-26, on
power-producing reactors fueled with circulating fused salts. The report is divided into
two major parts: 1. Reactor Design Study and 2. Materials Studies.

Until July 1, 1957, the Molten-Salt Reactor Program was largely a design study, with
only token expenditures for experimental work. As of July 1, the program was expanded
to include experimental work on materials. A further augmentation of the program occurred
on October 1, 1957, when personnel and facilities for additional research and experimen-
tation became available. As a result of this transition, the scope of this quarterly report
is considerably broader than that of the previous report, particularly with respect to
metallurgy and chemistry. Similarly, it is expected that future quarterly reports will

present the activities of the Experimental Engineering group.

UNCLASSIFIED
MOLTEN-SALT REACTOR PROGRAM QUARTERLY PROGRESS REPORT

SUMMARY

PART 1. REACTOR DESIGN STUDIES

Nuclear Calculations

Additional calculations were made of the nuclear
characteristics  of two-region, homogeneous,
molten-salt, converter reactors. Critical-inventory
calculations revealed that for a 9-ft-dia core, the
minimum inventory would be about 100 kg of U235,
The volume of fuel in the external system was
taken to be 340 ft3 for a power level of 600 Mw
(thermal).

Regeneration ratios were obtained as a function
of inventory for a 600-Mw system, with thorium
concentration as a parameter. From an analysis
of the ratios, an envelope was found which is the
focus of points of maximum regeneration ratio for
a given fuel inventory., From this envelope it may
be concluded that (1) with o fuel inventory of a
trifle over 100 kg, a regeneration ratio of 0.4 can
be obtained (in an 8-ft core), (2) by doubling the
inventory a regeneration ratio of 0.6 can be ob-
tained (also in an 8-ft core), (3) by doubling it
again a regeneration ratio of 0.65 can be obtained
(in a 10- to 11-ft core}, and (4) further investment
of fuel would have a negligible effect on the
regeneration ratio,

A design-point selection is to be attempted by
balancing fuel savings by regeneration against
inventory and processing charges; however, it
appears at present that 400 kg of U233 may be an
economical maximum for these systems. A few
calculations on systems fueled with U232 have
indicated much lower critical inventories and better

regeneration ratios than those obtained for U233
fuel,
Gamma Heating of Core Vessel
It was estimated that for operation of the

Reference Design Reactor at 600 Mw in a core
vessel 6 ft in diameter with T mole % ThF , in the
fuel, core gamma rays will liberate 13.4 w/em? in
the core vessel wall, Heating by gamma rays
emitted in the blanket was found to be 0.97 w/cm3,
and capture gamma rays originating in the wall
were found to contribute 1,63 w/cm? to the heating.

The calculations were made for a pure nickel core
vessel, and the results are somewhat lower than
those to be expected from the calculations now
being made for an INOR-8 alloy vessel.

Heat Transfer Systems

Two thermodynamic systems for producing power
from the heat from a molten-salt reactor are being
considered, and components and conditions repre-
senting preliminary optimization, with respect to
cycle efficiency and component sizes, have been

selected. Particular attention has been given to
limiting thermal stresses. One system being
studied ftransfers heat from the fuel salt to a

coolant salt to sodium to water, and the other
substitutes mercury for the sodium, The electrical
output from a 600-Mw (thermal)} reactor would be
258.6 Mw with the sodium system, and 295.8 Mw

with the mercury system.

PART 2. MATERIALS STUDIES
Metal lurgy

A coordinated program is under way for the in-
vestigation of container materials for molten salts
that will permit operation of a molten-salt reactor
for long periods of time at temperatures up to
1300°F, Of the materials investigated, the nickel-
base alloys are the most suitable to the specifica-
tions. Inconel is the best suited of the commer-
cially available materials; but, since its corrosion
resistance and high-temperature strength are
marginal, the alloy INOR-8 has been developed.
Inconel is being studied for comparison and as a
secondary choice, The metallurgical program for
investigating these materials will include material
property studies and fabrication development,
Techniques for remote welding and inspection are

also being studied.

The material property studies are, at present,
concerned primarily with the procurement of the
INOR-8 shapes (that

welding wire, etc.) needed for corrosion tests, and

is, pipe, sheet, tubing,

for physical and mechanical property tests, Pre-
liminary studies of cold-rolled sheet material have
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

indicated that the decrease in ductility of INOR-8
is greatest during the first 40% reduction in thick-
ness. The recovery and recrystallization charac-
teristics of the cold-worked INOR-8 alloy are being
investigated.,

Mechanical property studies of welded joints of
a nickel-molybdenum alloy similar to INOR-8 have
indicated that the alloy is weldable and that the
mechanical properties of the joint are satisfactory
both in the as-welded and the aged conditions.

Thermal-convection loop tests of Inconel and
INOR-8 are being made in order to provide data on
the corrosive properties of beryllium-bearing fuels
as compared with the properties of zirconium-base
fuels, the corrosive properties of fuel mixtures
containing large quantities of thorium for breeding,
and the corrosive properties of non-fuel-bearing
fluoride mixtures for use as secondary coolants.
These studies will be supplemented with forced-
circulation loop tests at flow rates and tempera-
ture conditions simulating those of an operating
reactor,

Radiation Damage

An in-pile thermal-convection loop is being pre-
pared for operation in the LITR in order to obtain
information on the effect of radiation on corrosion
of the container materials by the various fuels
being considered. Analyses of the fission products
in the fuels will also be made. An improved de-
sign hos been worked out for the installation of
thermocouple leads in the air annulus, and a
mockup test is being prepared. Calculations were
made for the equilibrium distribution of fission
gases in the system, and the charcoal trap to be
included in the system was sized accordingly,

Preparations are being made for irradiations of
lithium-beryllium-uranium flucride fuels in INOR-8
capsvules in the MTR.

Chemistry

A review has been made of the extensive body
of information available on fused salts in order to
orient the development of fuel solvents, fuels,

breeder blankets, and secondary coolants for use
in molten-salt power reactors. The systems of
most promise appear to be those which contain
BeF, with LiF and/or NaF.

Efforts were made to resolve discrepancies in
reported melting points of BeF,, and additional
confirmation of the reported value of 545°C re-
sulted.
vealed mixtures that may be of interest as fuel
solvents and carriers and as coclants in the sys-
tems LiF-BeF,, NaF-BeF,, and NaF-LiF-Bef,.
The NaF-LiF-UF, system does not appear to be
useful directly as a fuel, but low-melting mixtures
of interest are available in the LiF-BeF -UF,
system. A number of ThF ,-containing systems
are being studied, and tentative phase diagrams
have been prepared for the BeF ,-ThF,, LiF-ThF,
and NaF-ThF, systems. The LiF-BeF,-ThF, and
NaF-BeF,-ThF, systems have been shown to be
remarkably similar to their UF ,-containing analogs.
It thus appears that a relatively wide choice of
useful breeder materials is available. Studies of
containing plutonium fluorides

Phase-diagram investigations have re-

systems
initiated,

An analysis of the corrosion mechanism in
systems in which fluoride fuels are contained in
nickel-base alloys which contain molybdenum and
chromium was applied to the MSR (Molten-Salt
Reactor) system. |t appears virtually certain that
with small chromium activities such as those in
inconel and the INOR alloys, and with small
temperature drops such as those contemplated for
most reactors, chromium deposition will not result
if fuel mixtures based on the BeF,-containing
system are used,

The available information on fission-product
behavior in fluoride fuels has been analyzed in
terms of the MSR program. Preliminary experi-
ments with NaF-BeF, {57-43 mole %) have shown
the solubility of CeF, in this mixture to be con-
siderably less than in NaF-ZrF, mixtures and to
be more temperature sensitive,

were

The anticipated,
nearly ideal, solid solution behavior of mixtures of
rare earths in the BeF,-containing system, along
with reduced solubility, should make the fission-
product partition process quite attractive,
Part 1
REACTOR DESIGN STUDIES
 

NUCLEAR CALCULATIONS

L. G. Alexander

In further nuclear calculations critical inven-
tories and regeneration ratios were obtained as
functions of core diameter and thorium concentra-
tion in the core. For these calculations the new
intermediate-concentration
thorium were used.

cross sections for
Also, minor corrections to
some of the previously reported cases have been
obtained. Relevant specifications for the systems
studied are given in Table 1.1, The results of the
calculations are summarized in Table 1.2; analyses
of the results are being made.

A graph of the critical concentrations estimated
for these reactors is presented in Fig. 1.1, It may
be seen that the curve for 1 mole % ThF , does not
conform to the general pattern. The calculations
have been checked and the results are believed to
be correct. In an effort to clarify this behavior,

the flux-averaged cross sections for fission of y23s

Table 1.1. Specifications for Two-Region
Molten-Fluoride-Salt Reactors

 

Core

Diameter

5t0 10 ft

Carrier salt 69 mole % LiF,

31 mole % BeF2

Mean density 2.0 g/cm3
Li composition 0.01% Li6
U238 concentration 7.5% of U235 concentration

s See Table 1.2

concentration

Core Vessel

Thickness ]/3 in.

Composition INOR-8 alloy
Blanket

Thiekness 2 ft

Composition

25 mole % ThF4,
75 mole % LiF

Mean density 4.25 g/cm’®

Geometry Spherically symmetric

 

J. T. Roberts

and capture in thorium were computed from the
flux spectra:

g = .1-020 fvc No ¢(u) du dV ,

where V _ = volume of core, N = atoms per cubic
centimeter. The results for reactors having zero
and 1 mole % ThF, in the core are shown in
Table 1.3.

It is thought that increasing the thorium concen-
tration in the 10-ft core decreases the fission
cross section more or less uniformly so that the
critical concentration rises regularly. In the 6-ft
cores the spectrum is already hard. Adding thorium
hardens it further, but the effect on the fission
cross section is less., There is a regenerative
effect at work, The spectrum is hardened in the
following three ways: (1) by decreasing core

size, (2) by increasing thorium concentration,
and (3) by increasing uranium concentration,
If the spectrum is hardened relatively less

in the 6-ft core by the addition of thorium, the de-
crease in uranium cross section is relatively less;
hence, less uranium is needed, and the spectrum

UNCLASSIFIED
ORNL~LR —DWG 24920R

n
n

n
N

~N
Q

@

&

s

n

1 mole % ThF,

CRITICAL CONCENTRATION [atoms/(cm®x 15 '*)]
B o co 6

N

 

CORE DIAMETER (f1)

Fig. 1.1. Critical Concentrations of U235 for Two-

Region, Homogeneous, Moclten-5alt Reactors.
 

MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

Table 1.2, Nuclear Characteristics of Clean Two-Region Molten-Flucride-Salt Reactors at a Fuel Temperature of 1150°F

 

 

Case number

Fuel

Thorium concentration in fuel (mole %)
Diameter of core (ft)

Critical concentration (mole % UF,)
Critical mass (kg)

Critical inventory* (kg)

Regeneration ratio
Core
Blanket

Total

Neutron balance

Captures in

Fuel

U238 }
Th in core

Th in blanket
Li+ Be in core
F in core

Core vessel

Li + F in blanket

Leakage from blanket

Neutron yield,* 77

 

 

87 53 54 55 106 86 107 61 88 89 90 9 92 93 63 84 62 87A
U235 U235 U235 U235 U235 U235 u23.‘.5 U235 U235 u235 U235 U235 U235 U235 U235 U235 U235 U233
0 0 0 0 025 025 025 0.25 0.5 0.5 0.5 075 075 075 1 1 1 0.25
5 6 8 10 6 7 8 10 6 8 10 6 8 10 6 8 10 5
0.453 0,110  0.047  0.033 0.243 0.160 0.167 0.073 0.572 0.185 0.113 0.772 0.318 0.180 0.875 0.492 0.281  0.147
82 47 49 67 102 80 80 106 180 138 164 243 237 260 275 367 409 27
509 189 110 11 408 230 181 175 720 313 270 972 537 428 1100 831 674 165
0.039 0.025  0.014  0.012 0.170 0.201 0.224 0.256 0.229 0.332 0.390 0.261 0.378 0.459 0.288 0.394 0.485  0.160
0.541 0.531  0.408  0.303 0.432 0.410 0.350 0.262 0.368 0.305 0.233 0.338 0.267 0.206 0.322 0.236 0.183  0.718
0.580 0.556  0.422  0.315 0.602 0.611 0.574 0.518 0.597 0.637 0.623 0.599 0.645 0.665 0.610 0.630 0.668  0.878
0.5578 0.5210  0.4983  0.4918 0.5513 0.5227 0.5105 0.5007 0.5716 0.5279 0.5120 0.5781 0.5497 0.5262 0.5781 0.5663 0.5449  0.4574

0.0072 _
0.0220 0.0128 0,.0060 0.0935 0.1053 0.1145 0.1283 0.1308 0.1752 0.2000 0.1510 0.2071 0.2416 0.1657 0.2236 0.2643

0.0000 0.0729
0.3013 0.2768  0.2029  0.1490 0.2380 0.2124 0.1788 0.1310 0.2104 0.1612 0.1195 0.1955 0.1466 0.1086 0.1863 0.1136 0.0995  0.3281
0.0239 0.0686  0.1587  0.2254 0.0301 0.0636 0.0949 0.1414 0.0164 0.0552 0.0889 0.0116 0.0306 0.0573 0.0096 0.0191 0.0347  0.0329

0.0470 |

0.0741
0.0950 0.1208 0.0072 0.1278 0.0871 0.0942 0.1013 0.0986 0.0708 0.0805 0.0796 0.0638 0.0660 0.0663 0.0603 0.0574 0.0566  0.1087

0.0046
.79 1.93 2.01 2.03  1.81 .91 19 2,00 1.75 189 1.95 1.73 1.82 1.90 173 176  1.84 2.19

94

U233

0.25

0.092
29

116

0.175

0.592

 

F

0.767

0.4538

0.0896
0.2686
0.0696

0.1184

2.20

95

233

0.25

0.056
41

94

0.252

0.421

 

B

0.673

0.4 502

0.1133
0.1894
0.1298

0.1173

2,22

96
233
0.25
10
0.043
63

103

0.285

0.306

 

0.591

0.4486

 {ome)

0.1371
0.1763

0.1100

2,23

 

*For 600-Mw system with 340 #3 of fuel in system outside core.

 

 

 
Table 1.3. Flux-Averaged Cross Sections in

Molten-Salt Converter Reactors

 

 

Core Thorium Average Cross
Diameter Concentration Section
(1) in Core (barns)

(mole %) U235 Th

10 0 94 3.0

1 14 1.9

é 0 36 2.7

1 8 1.8

 

is hardened less, etc, |f the effect of the thorium
decreases sharply around 0.75 mole % ThF, be-
cause of resonance shielding, the deviation in the
results for the 6-ft core with 1 mole % ThF , in the
fuel would be satisfactorily accounted for, It is
planned to calculate the cross sections for cores
with 0,25 and 0.75 mole % ThF, in order to verify
this speculation.

The critical masses were computed from the core
diameters and critical concentrations, The results
are listed in Table 1.2 and plotted in Fig. 1.2,
together with four additional points resulting from
preliminary calculations for 5-ft cores. The
apparent deviation of the 6-ft core with 1 mole %
ThF, is more pronounced than in Fig. 1.1. It
should be remembered, however, that regardless of
the thorium concentration the critical masses of
all these reactors must decrease as the diameter
decreases to below some critical value. Two
points, one for a small, homogeneous, molten-salt,
beryllium-reflected critical experiment and the
other approximating a metallic uranium critical
assembly, have been plotted in the lower left-hand
corner of the figure. All the curves in Fig. 1.2
probably pass near these two points and through
the origin; a family of curves consistent with this
assumption is shown in Fig. 1.3. The calculated
points appear to correlate reasonably well on this
basis. Further calculations will be performed to
verify some parts of the curves, especially parts
of the curve for 1 mole % ThF . As will be shown
later, however, the reactors corresponding to the
portions of the curves to the left of the 6-ft ordinate
are not economically attractive, and the details of
the peaks will not be investigated.

The solid portions of the curves in Fig. 1.3,
which are believed to be correct, and the curve for

PERIOD ENDING OCTOBER 31, 1957

UNCLASSIFIED
ORNL—LR—DWG 26533
450 ] —! [

 

i i 1
O PRELIMINARY CALCULI{ATIO NS FOR 5-ft CORES

400

 

350

300

250

200

CRITICAL MASS (kg of UZ>%)

150

 

100

 

50

 

 

 

 

 

 

CORE DIAMETER (ft)
Fig. 1.2. Critical Masses for Molten-Salt Reactors,

1 mole % ThF,, which was accepted provisionally,
were used to compute the critical inventories
listed in Table 1.2 and shown in Fig. 1.4, (The
values at 7 and 9 ft were obtained by interpolation
on a graph of the regeneration ratio vs diameter
and ThF, concentration,) The volume of fuel in
the external system (pump, pipes, heat exchangers)
was taken to be 340 ft3, which corresponds to a
power level of 600 Mw (thermal). It may be seen
that a minimum inventory of about 100 kg of U23°
is obtained at a core diameter of 9 ft, The corre-
sponding power density in the core is 55 w/cm’
(average) and the specific power is 6 Mw per kilo-
gram of fuel in the system,

The regeneration ratio is shown in Fig. 1.5 as a
function of inventory (for a 600-Mw system), with
thorium concentration as a parameter, It is thought
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

UNCLASSIFIED

ORNL—LR—DWG 26534
450 -

 

400

350

300

250

 

200

CRITICAL MASS (kg of U23%)

100

50

 

CORE DIAMETER (ft)

Fig. 1.3. Tentative Correlation of Critical-Mass Data
for Molten-Salt Reactors Based on Assumption that All
Curves Pass Through Origin.

that for a given thorium concentration the critical
inventory falls as the core diameter increases
because the concentration falls more rapidly than
the volume increases., Simultaneously the ieakage
of neutrons to the blanket decreases, but this
decrease is, at first, more than offset by the in-
crease in captures by the thorium in the core be-
cause of the increase in the ratio of thorium con-
centration to uranium concentration, Eventually,
however, the uranium concentration tends to level
off ot the finite concentration corresponding
to the infinite reactor, and the volume increase
becomes the important factor. As a result, the
inventory passes through a minimum. At about

 

UNCLASSIFIED
ORNL-LR-DWG 24967R
1100

 

1000

 

 

900

8C0

700

800

 

500

 

400

 

CRITICAL INVENTORY (kg of UZ3%)

300

 

 

200

100 *1

4 5 6 7 8 9 10
CORE DIAMETER (ft)

 

 

 

 

 

 

 

Fig. 1.4,

Salt Reactors.

Critical Inventories for 600-Mw, Molten-

the minimum point, the thorium captures in the
core level off, But, since the leakage to the
blanket continues to fall with volume increase,
the regeneration ratio falls off. Also, parasitic
captures in fluorine, etc., increase as the spectrum
softens. Thus the curves in Fig. 1.5 turn down-
ward and eventually must turn back to the right.,

The downturn of the curves is surprisingly sharp,
It appears that the ‘‘elbows’’ define an envelope
which is the locus of points of maximum regenera-
tion ratio for a given fuel inventory, This envelope
is shown in Fig. 1.6, The principal conclusions
to be drawn are that {1} with a fuel inventory of a
trifle over 100 kg, a regeneration ratio of 0.4 can
be obtained (in an 8-ft core), (2) by doubling the
UNCLASSIFIED

ORNL-LR-DWG 24921R

 

 

 

 

 

 

 

 

 

 

 

0.8
0.7
o
g
xr 06
g
5 1o
& 0.5 |
=
Il
[
Ly /
0.4
0.3 * J

 

 

 

O

500 400 600 800 1000
CRITICAL INVENTORY (kg of U%3°)

1200

Fig. 1.5. Regeneration Ratios in Two-Region, 600-Mw,

Molten-Salt Reactors.

UNCLASSIFIED

ORNL—-LR-DWG 24922R

 

 

0.8

0.6 ’

 

 

REGENERATION RATIO

04

 

 

 

 

02 A

   
    

 

 

0 200 400 600 800
CRITICAL INVENTORY (kg of U239

Fig. 1.6. Optimum Regeneration Rotios for Two-

Region, 600-Mw, Molten-Salt Reactors.

1000

PERIOD ENDING OCTOBER 31, 1957

inventory a regeneration ratio of 0.6 can be ob-
tained (also in an 8-ft core), (3) by doubling it
again a regeneration ratio of 0.65 can be obtained
(in a 10- to 11-ft core), and (4) further investment
of fuel would have a negligible effect on the re-
generation ratio, The optimum diameter appears fo
increase steadily after the inventory increases

beyond 200 kg.

A plot of 3 — 1 is also included in Fig. 1.6 to
give the upper limit on the regeneration ratio. The
interval between the envelope and the 7 — 1 curve
represents parasitic absorptions in fluorine, etc,
These absorptions decrease with increasing in-
ventory, because of increased competition from
uranium and thorium, and also because of hardening
of the spectrum. Judging from the trends of the
curves in Fig., 1.6, however, it would seem that,
although n — 1 approaches a minimum at an inven-
tory of 600 kg, further significant reduction in
parasitic absorptions can be obtained only by
enormous increases in the inventory.

Selection of a design point for reactors of this
type will consist in balancing fuel savings by re-
generation against inventory and processing
charges. This will be attempted; however, it does
not seem likely that it would pay to invest more
than 400 kg of U?3% in such systems.

It must be remembered that the results discussed
above apply only to the initial, clean state; that
is, they apply only in the absence of fission
products and uranium isotopes (other than U238y,
The accumulation of these substances will of
course affect both the critical inventory and the
regeneration ratio. However, the performance of
the system may not be impaired significantly for
some time after startup. In the first place, the
U233 generated in the blanket will have a fission
cross section in these epithermal reactors about
twice as high as that of U235, Adding this
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

U232 15 the core will tend to offset the ac-

cumulation of poisons, and it may not be neces-
sary to add UZ33 for some time, Further, 7 for
U233 is much better than that for U23°, especially
in the intermediate-energy range, and replacing
U235 with U233 will reduce the parasitic absorp-
tions in the fuel. These effects will be studied
by means of the Sorghum program being prepared
for the Oracle,

A few calculations on systems fueled with y233

have been completed. Four cases, in which core
diameters range from 5 to 10 ft and the ThF , con-
centration is 0.25 mole %, are described in Table

1.2. The results are compared with those for cor-

responding U233 systems in Fig. 1.7. The critical
inventories are much lower with U233 fuel, and
the regeneration ratio is much better. It is hoped
that regeneration ratios well above 0.9 can be
achieved at inventories below 500 kg, The calcu-
lations will be extended.

UNCLASSIFIED
ORNL —LR—-DWG 24923R

 

 

 

 

 

 

 

 

1.0
U233
<z
o /
= 0.8 ®
i
o«
Z I
o .
2 U235
5 - .-g
.
LD .
w
a
0.9
0 200 400 600
CRITICAL INVENTORY (kg of U)
Fig. 1.7. Comparison of Regeneration Ratios in

Molten-Salt Reactors Containing 0.25 Mole % ThF
235 233 4
and U“"% or U“"~-Enriched Fuel.

 

GAMMA HEATING OF CORE VESSEL

L. G, Alexander

The study of gamma heating of the core vessel
of the Reference Design Reactor! was continued.
The spectra of the delayed, capture, and inelastic-
scattering gamma rays were estimated from data
in the literature, ond more reliable gamma-energy
absorption coefficients for the various materials
were computed from the basic data presented by
Grodstein.?

The calculations were carried out for a core
vessel of pure nickel, The heating by core gamma
rays will not be affected much by the presence of
alloying constituents. With the use of the new
data, it was estimated that during operation at
600 Mw in a core vessel 6 ft in diameter with 1

 

'H. G. MacPherson et al., Molten Salts for Civilian
Power, ORNL CF-57-10-41 (Oct. 10, 1957).

2G. W. Gredstein, Natl., Bur. Standards (U.S.) Circ.
583 (1957).

10

L. A. Mann

mole % ThF4 in the core (see Case 63 in Table 1.2
of preceding section of this report), core gamma
rays will liberate 13.4 w/em?® in the core vessel
wall,

Similarly, the heating effected by capture gamma
rays emitted in the blanket was estimated to be
0.97 w/em® at the inside surface of the core
vessel, The previously obtained value of 0.032
w/em? (for an INOR-8 core vessel) was obviously
in error by an order of magnitude and will be
recomputed,

The gamma heating effected by capture gamma
rays originating in the nickel wall was estimated
to be 1.63 w/ecm®. In an INOR-8 alloy vessel,
this contribution will be approximately twice as
great because of captures in the molybdenum of
the alloy. Calculations for an INOR-8 core vessel
are being made.
PERIOD ENDING OCTOBER 31, 1957

HEAT TRANSFER SYSTEMS

B. W. Kinyon

A study was made of two thermodynamic systems
for producing power from a molten-fluoride-salt
reactor capable of producing 600 Mw of heat. One
system involves transferring heat from the fuel
salt to a coolant salt to sodium to water and
achieves turbine steam conditions of 1800 psia
at 1000°F, with reheat to 980°F. The other system
involves transtferring heat from the fuel salt to a
coolant salt to mercury to water. Mercury vapor at
180 psia and 1000°F would expand through @
turbine and exhaust at 7 psia and 605°F, The
condensing mercury would generate steam at 1250
psia, which would be superheated in two steps to
950°F by mercury vapor bled from the turbine and
taken directly from the line.

Temperature, flow, and pressure-drop conditions
were selected, and pertinent data were obtained for
the heat exchangers required for the two systems,
The components and conditions represent pre-
liminary optimization with respect to cycle effi-

The conditions were
attention to

ciency and component sizes.
selected with particular
thermal stresses to low valves,

The steam generator in the sodium system is of
interest from the standpoint of low stresses. A
high-flow rate and a small temperature change of
the sodium are desired in order to achieve a low
temperature difference at the hot end. These con-
ditions can be achieved by recirculating sodium
through the boiler and by taking sufficient sedium

limiting

from the pump to supply the heat needed ond to
provide the head for recirculation, The steam
generator best suited for this service is the Lewis
boiler, which utilizes bayonet tubes, 1t is a
natural-circulation boiler in which water feeds
down a central tube. Boiling occurs during upward
flow in the annulus,

In the mercury system,
mercury would take place in a J-shaped heat ex-
changer in which mercury flows inside the tubes,
and the coolant salt flows in the shell, A pressure
of 180 psia will provide saturated mercury vapor
at 1000°F, while a 16-ft additional head (required
to give a 270-psia total pressure) will result in a
saturation temperature of 1075°F. Downward flow
of the mercury in the short leg of the J, counter to
the salt flow, will raise the temperature of the
nonboiling mercury to 1075°F. Vaporization will
occur in the long leg, where both the mercury and

vaporization of the

F. E. Romie

the salt will be flowing upward. Heat for the
vaporization will come both from the salt and from
the mercury, With the proper salt flow, the dif-
ference between the salt temperature and the
saturation temperature of the mercury con be made
uniform along the boiling section. This will result
in low thermal stresses and effective use of the
heat transfer area. To avoid difficulties caused by
vaporizing too large a fraction of the mercury, a
recirculation ratio of 10 to 1 will be employed.

In the binary mercury-steam cycle, the mercury
turbine will produce 35% of the power and the
steam turbine will produce 65%. The combined
turbine heat rate will be 6940 Btu/kwhr, which
gives 49.3% efficiency., The sodium system would
have a turbine heat rate of 7920 Btu/kwhr, or
43.1% turbine efficiency. The higher efficiency of
the mercury system tends to compensate for the
higher investment cost; the unit costs of electricity
from the two systems are probably comparable.

The heat transfer area from fuel to water for the
sodium system is 270 ft2/Mw (electrical), that is,
about one-third or one-fourth the area in a con-
ventional fossil-fuel furnace. A comparison of
the characteristics of the sodium and mercury
sysfems is presented in Table 1.4

Table 1,4. Comparison of Mercury and Sedium Heat
Tronsfer Systems for a 600-Mw Molten-Salt Reactor

 

 

P arameter Sedium Mercury
System System
Turbine heat rate (Btu/kwhr) 7,920 6,940
Cycle efficiency (%) 43.1 49.3
Electrical output {Mw)
Mercury 101.8
Steam 258.6 194
Total m 2_9;-;
Total heat transfer area (f12) 69,860 99,520
Heat transfer area per unit of 270* 336

electrical output (ffz/Mw)

 

*This value for the sodium system would be somewhat
greater than the value for the mercury system if the con-
denser and feed systems were included.

11
Part 2
MATERIALS STUDIES
METALLURGY

W. D. Manly

J. L. Scott

The metallurgical work of the Molten-Salt Reactor
Program has as an objective the provision of a
container material for the molten salts that will
have sufficient structural strength and compati-
bility with the molten salts to permit operation of
a reactor for long periods of time at temperatures
up to 1300°F, The engineering and fabrication
data required for successful design and con-
struction of the reactor must be acquired, and,
as a longer range objective, remote welding or
other joining methods must be developed which
will permit the maintenance of highly radioactive
equipment,

The requirements for a construction material
for a molten-salt reactor are:

1. corrosion resistance to the fused fluorides of
interest for long times (5 to 20 years) at
temperatures up to 1300°F,

2. resistance to mass transfer
similar times and at temperatures up to 1100°F,

3. sufficient strength and ductility, at operating
temperatures, to allow thermal cycling and the
stresses expected in the reactor — phase trans-
formations should not occur which might ad-
versely affect the mechanical properties,

4. fabricability such that reactor components
(pumps, heat exchangers, shells, etc.,) can
be made by using standard practices,

5. weldability such that sound welds can be made
in heavy members under restraint,

6. suitable nuclear properties,

7. good oxidation resistance at 1300°F.

Of the materials investigated, the nickel-base
alloys are the most suited to the specifications.
The commercially available alloy which best
approaches the specifications is Inconel, but its
corrosion resistance and high-temperature strength
are probably marginal.

In an effort to obtain an alloy with better cor-
rosion resistance and strength in fused-salt en-

in sodium for

vironments, the alloy INOR-8 has been developed.
Tests at higher-than-MSR service conditions have
indicated the corrosion resistance of INOR-8, in
the fuel mixtures NOF-ZrF4-UF4 (50-46-4 mole %)
and NaF-KF-LiF-UF, (11.2-41-45.3-2.5 mole %),

to be ten times better than that of inconel. High-

A. Taboada

temperature rupture strengths were also found to
be superior to those of Inconel.

The corrosion problems can be estimated quali-
tatively from the results of short-term tests at
higher-than-MSR temperatures because long times
at low temperatures and short times ot high
temperatures have the same effects according to
the accepted theory of corrosion in these systems
(see subsequent section of this chapter on
““Chemistry of the Corrosion Process’’), The use
of high-strength materials in the MSR is being
considered as a means of obtaining high efficiency
with a low fuel inventory, since with the high-
strength material the poisoning effect of the core
container material may be minimized by using
smaller sections. Therefore the properties of
INOR-8 are being investigated at MSR service
conditions.  Further studies of Inconel will be
made in order to provide a basis for comparison
with the new data obtained for INOR-8. Inconel
will also be considered as a secondary choice
for the material for construction of an MSR. The
metallurgical program for investigating INOR-8 and
Inconel may be divided into three major categories:
material property studies, fabrication
gations, and supporting work.

investi-

Material Properties Studies. — Corrosion studies
will be continued for salts of interest, with
emphasis on long-term effects, and the mass
transfer of liquid metals at MSR conditions will
be investigated.  Mechanical property studies
will include relaxation tests to establish the
conditions under which creep can be expected to
be a problem and those conditions where tensile
data may be reliably used as design values.
Tensile tests will be conducted at many temper-
atures in the range from room temperature to
1400°F, and creep tests will be run in air and
in fused salts to estimate the limiting strain for
long-term operation, The effect of dynamic
loading, either thermal or mechanical, will be
explored.

Physical property data needed for design, such
as thermal conductivity, expansion coefficients,
etc., will be obtained; and metallurgical properties,

15
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

such as aging and recrystallization, which might
affect mechanical properties, will be investigated.

Fabrication Development. — The weldability of
INOR-8 will be established and welding procedures
will be developed in order to enable reliable
reactor component fabrication. The melting,
casting, forging, and rolling characteristics of
INOR-8 will be investigated, and developmental
work on techniques for fabricating tubing will be
continued.  The acceptable range of chemical
composition and properties will be established
so that material specifications may be written,

Miscellaneous Supporting Work. — The effect of
carburization on the properties of INOR-8 and
Inconel and the susceptibility of these materials
to carburization in fused salts will be investi-
gated. Work will also be carried out on remote
welding and inspection techniques. Engineering
designs of reactor components will be reviewed
from the metallurgical standpoint,

Details of the metallurgical work carried out
during the quarter are presented in the following
sections of this report,

FABRICATION

H. T. Inouye T. K. Roche
J. Spruiell

A program for obtaining the information required
to moke INOR-8 a commercial item has been
developed. As an qid in procurement, melting
and casting techniques are being studied. Methods
for fabricating the material are also being investi-
gated. Information on the kinetics of recrystal-
lization and phase transformation is being obtained
for use in fabrication studies, and the physical
properties of the material are being measured.

Material Procurement

At the present time there are two sources of
INOR-8. It has been purchased from Haynes
Stellite Company, on a purchase order, at a cost
of about $5.00 per pound of delivered alloy, and
it has been produced on a subcontract basis by
the Westinghouse Electric Corporation.  Both
sources present some procurement difficulties,
Orders on the Haynes Stellite Company are ac-
ceptable provided that 10,000-1b heats are pur-
chased. To date, only small pieces have been
received from Westinghouse. Considerable de-

viations from the scope of the Westinghouse

16

contract will have to be made in order to supply
the material requirements for INOR-8 at ORNL,
In this respect, a standardization of the sizes and
shapes would facilitate the delivery of materials,
The shapes completed and available for the
fabrication of corrosion testing loops are listed
in Table 2,1. This material was obtained from
heat SP-16 produced by Haynes Stellite Company.

Table 2,1, Shapes of INOR~8 Available for the

Fabrication of Corrosion Testing Loops

 

ltem Size Amount

 

Bar 3 in. dia About 2 to 3 ft

%, ine dia 85 1b (50 f1)
1 in. dia 24 in.
1]/2 in. dia 72 in.
Tubing ]‘/2 in. OD, 0.065-in. wall 18
1in. OD, 0.065in. wall 182 ft
% in. 0D, 0.025in. wall 161 f
00250 in. OD, 0.025-in- qul 600 ft
0.500 in. OD, 0.045-in. wall 1700 ft
Pipe % in., sched 40 100 f
Sheet 0.065 x 36 X 96 in. 2 pieces
Plate 1 X 24 x 48 in. 1 piece
]/4 X 36 x 40 in. 1 piece
]/2 X 12 x 20 ins 9 pieces

 

Items currently scheduled for delivery are 200 ft
of l/B-in. sched-40 pipe, 100 ft of 3-in. sched-40
pipe, and a small quantity of welding wire.

The Westinghouse program involves three phases
of study. At present the developmental work is
concerned with providing material for the fabri-
cation of seamless tubing and stock for weldability
and mechanical property determinations, Numerous
heats have been made in order to supply the
material requests. The second phase of the
program will be an investigation and evaluation
of fabricated products produced from vacuum-
induction-melted and the orc-melted material. The
third and final phase of the study will involve
the reproducibility of results, the production of
standard products, and the reprocessing of de-
veloped scrap.

Fabrication of INOR-8

The fabrication properties of heat SP-16 pro-
duced by Haynes Stellite Company were studied.

PERIOD ENDING OCTOBER 31, 1957

The variations in the tensile properties of cold-
rolled sheet were determined for reductions up
to 90% in thickness. The data, presented in
Fig. 2.1, show that the decrease in ductility is
greatest during the first 40% reduction in thick-
ness. This phenomenon has been observed in
other alloys which exhibit a high work-hardening
rate,

 

 

 

3 UNCLASSIFIED
{x107) ORNL~LR-DWG 24088
250
| /‘
200 //

 

ULTIMATE TENSILE STREN;HH)/

 

0.2 % OFFSET YIELD STRENGTH

60

 

%
/

STRENGTH (psi)

40

 

100 \

- ELONGATION

 

50 k

N

 

 

 

 

 

20

PER CENT ELCNGATION (2-in. gage length}

 

 

 

 

 

 

 

 

0 10 20 30 40

50 &0 70 80 90

REDUCTION IN THICKNESS (%)

Fig. 2.1. Work-Hardening Curves for INOR-8, Heat SP-16, Annealed 1 hr at 2150°F Before Being Given the

Reductions Indicated.

17
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

Transformation Kinetics

The recovery characteristics of the cold-worked
INOR-8 alloy are shown in Figs. 2.2 through 2.5.
Of particular interest are the peaks in the curves
observed for the tensile strength, yield point, and
hardness at an annealing temperature of 1000°F
after cold working 80%. As yet it is uncertain
whether the increase in strength is due to recovery
or to carbide precipitation, Microstructural obser-
vations of the cold-worked specimens show that
recrystallization for the 80% cold-worked alloy
begins at 1350°F and is complete at 1400°F. For
the 40% cold-worked alloy, recrystallization begins
at about 1500°F and is complete at 1550°F.

The effect of the time at the annealing temper-
ature is shown by the isothermal recrystallization
curves in Fig. 2.6. These curves show that
although recovery and recrystallization are pro-
moted by increasing the time at temperature, these
variables are not so significant as temperature,

WELDING AND BRAZING
G. M. Slaughter

Mechanical property studies on welded joints
of a nickel-molybdenum alloy similar to INOR-8
have been made. The tests were conducted on a
Battelle Memorial Institute alloy, heat No. 3766,
and it was concluded that this alloy was weldable
and that the mechanical properties of the joint
were satisfactory both in the as-welded and the
aged conditions.

P. Patriarca

Preliminary tests have also been conducted on
two other alloys of INOR-8 composition: a
commercial-size heat from Haynes Stellite Com-
pany, designated SP-16, which had been annealed
]/2 hr at 2150°F and air cooled; and a small ORNL
heat, designated heat 30-38,

The initial studies on these alloys were primarily

screening tests for determining their relative sus-
ceptibilities to weld-metal cracking, One type

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

3 UNCLASSIFIED
{(x107) ORNL-LR-DWG 24040
300
L~ \
g0% REQY
@ 60% REDUGED
< 200 ——
I
O
= 40 % REDUCED A
a ___A'__, o .
B
L
W L
» 20% REDUCED
= —r lfi
L
! 10% REDUCED .
w —{— T —
-
<
= ™o
= 100
.|
0
O
0 200 400 600 800 1000 1200 1400 1600 1800 2000 2200

ANNEALING TEMPERATURE (°F)

Fig. 2.2, Room-Temperature Tensile Strength of INOR-8, Heat SP-16, Deformed by Rolling and Annealed 1 hr

at {ndicated Temperature,

18
(x10°)
260

240

220

200

PERIOD ENDING OCTOBER 31, 1957

UNCLASSIFIED
ORNL-LR—DWG 24042

 

 

 

 

 

 

 

L

 

60% REDUCGED
/"_-i ~

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

‘0
e 180 _;'—4 \__
=
(€]
&
= 1
}_
» 160 40 % Rep r
° RE
J
; \
E 140 \
o N
L
o
2
Jd
d 120 . I
20% REDUCED
100
10 % REDUCED 1
]
80 \
© 1
60
40 L
0 200 400 600 800 1000 1200 1400 1600 1800

ANNEALING TEMPERATURE (°F)

Fig. 2.3. Yield Strength (0.2% Offset) ot Room Temperature of INOR-8, Heat SP-16, Deformed by Rolling and
Annealed 1 hr at Indicated Temperature,

19
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

UNCLASSIFIED
ORNL-LR—-DWG 2408B7R

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

600
/0"“" »\;io REDUCED
500 =
/ 60, REDUCED
400 — 409, REDUCED
=4 /
> | \:
>
wn
w
=
0
x 20 %, REDUCED
T
I-—“-'_—l__._.—l--lI‘—"'_'. i,! - _—r—\
/A-/ A
300 \;
O G 10, REDUCED
)
200
100 :
o 200 400 600 800 1000 1200 1400 1600 1800 2000 2200

ANNEALING TEMPERATURE {°F)

Fig. 2.4, Hardness (VHN) of INOR-8, Heat SP-16, Deformed by Rollifig and Annealed 1 hr at Indicated Tem-

perature.

20
PERIOD ENDING OCTOBER 31, 1957

UNCLASSIFIED
ORNL-LR~-DWG 24041

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

70 [

60
e /
'_
2
W 50 A
Ll
o
& /
o
- g
£ 40 A
[aN]
Z ! 10% REDUCED |
5 1 T
> |
o
|
uwJ
F

P

z L 50 % REDUCED "]
& 20 -
a
)
a

10 /

40 %, REDUCED L
609, REDUCED
T | 80 % REDUCED
0 |
0 200 400 600 800 1000 1200 1400 1600 1800 2000 2200

ANNEALING TEMPERATURE (°F)

Fig. 2.5. Room-Temperature Elongation (2-in. Gage Length) of INOR-8, Heat SP-16, Deformed by Rolling and

Annealed 1 hr at Indicated Temperature.

of test, termed the circular-groove test, utilizes
a 3/32-in.-wide, 2-in,-diq, %-in.-deep, circular
groove machined in a 4 x 4 in. specimen of ]/2-in.
An inert-arc fusion pass (no filler metal
addition) is made around the groove and the
absence or presence of weld-metal cracks is
observed, Test samples of heats SP-16 and 30-38,
together with samples of four other alloys for
comparison, are shown in Fig. 2,7. It may be
seen that the restraint of the specimen caused
complete circumferential cracks in the Haynes
SP-16 specimen, while no cracking can be ob-
served in the ORNL 30-38 dlloy. As was ex-
pected, the Inconel specimen cracked extensively,
because no columbium-modified filler metal (known

as INCO 62) was added.

plate.

Another type of crack test, designated as the
open-end slot test, utilizes a 2% x 5-in, specimen
of ]/2-in. plate. A 3-in,-long, ]/]6-in.-wide slot
is cut from one end, and a fusion pass is made
from the open end of the slot toward the closed
end, As may be seen in Fig. 2.8, cracks were
found in the Haynes SP-16 alloy but not in the
ORNL 30-38 alloy.

As a means of further evaluating the weldability
of the Haynes SP-16 alloy, the preparation of
typical weld-test plates (Fig. 2.9) for radiographic,
mechanical-property, and metallographic studies
was initiated. The first test plate, NiMo plate
No. 36, was fabricated by using SP-16 alloy filler
metal which had been sheared from ]/16-in. sheet.
Preparation of this plate was terminated after

2]
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

 

 

 

 

 

 

 

 

 

 

 

 

{x 403) UNCLASSIFIED
ORNL~LR-DWG 241450
220
200
~ A
4 N
- N
— N, A
2 N
w480
o
1_.
» A
wJ
1
5 \
=
o
- 160 N
5 N
= \
= N —
< M 60
5 a sy

 

140

 

 

 

 

 

 

 

 

 

120

 

 

 

 

ELONGATION (%)

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

50 100 200 500 1000 2000
ANNEALING TIME (min)

Fig. 26. lsothermal Annealing Curves for INOR-8, Heat SP-16, for 60 and 80% Cold-Reduced Specimens.

deposition of the first few passes because gross
weld-metal cracking had occurred. The extent of
the cracking is evident in Fig. 2,10, Metallo-
graphic examination of the weldment revealed
weld-metal cracks of the type shown in Fig. 2.11
and base-metal cracks of the type illustrated in
Fig. 2.12.

In view of the unsatisfactory results with the
SP-16 filler metal, another weld-test plate, NiMo
plate No. 37, was fabricated by using Hastelloy W
filler wire. This plate is shown in Fig. 2.13.
Although no weld-metal cracks were found by
visual and dye-penetrant inspection, metallo-

22

graphic examination of the weld joint revealed
severe base-metal cracks of the type shown in
Fig. 2.14. Extensive fusion-line porosity of the
type illustrated in Fig. 2.15 was also found,
Metallographic examination of the SP-16 heat
revealed an abnormal number of stringers, which
may be responsible for the porosity and extensive
base-metal and weld-metal cracking. On the other
hand, ORNL alloy 30-38 appeared to be immune
to weld-metal cracking, and thus melting practice
may be responsible for the poor performance of
the Haynes heat. Additional tests are being
conducted to evaluate weld-metal and base-metal
cracking tendencies of the ORNL heat,
INOR-8
HAYNES SP—16 (HT 8284)

 

 

PERIOD ENDING OCTOBER 31, 1957

UNCLASSIFIED
Y-24145

HASTELLOY B

Fig. 2.7. Circular-Groove Test Specimens for Studying Weld-Cracking Susceptibility.

DYNAMIC CORROSION

J. H. DeVan J. R. DiStefano
R. S. Crouse

Test Program

A basic question underlying the development of
a molten-salt reactor for power production concerns
the long-time compatibility of structural materials
with the intended fluid heat-transfer mediaq, fluoride
salts and sodium.

As stated above, two structural alloys, Inconel

and [INOR-8, are presently being studied as
possible container materials for these appli-
cations. Corrosion experiments utilizing thermal-

convection loops and forced-circulation foops fab-
ricated of both alloys are presently under way.
The test program will, in general, involve three
phases. In the first, the relative corrosive
properties of the several fluoride mixtures listed
in Table 2.2, are to be determined in thermal-

convection loops operated for a period of 1000 hr.,
These tests will provide data on the corrosive
properties of beryllium-bearing fuels as compared
with the properties of zirconium-base fuels, the
corrosive properties of fuel mixtures containing
lorge quantities of thorium for breeding, and the
corrosive properties of non-fuel-bearing fluoride
mixtures for use as secondary coolants,

The second phase of testing, for which thermal-
convection loops will again be used, will subject
those systems which appear to be compatible in
phase-1 experiments to more extensive investi-
gations at longer time periods and ot two temper-
atures, 1250 and 1350°F. The third and final
phase of this out-of-pile testing will be conducted
in forced-circulation loops at flow rates and
temperature conditions simulating those of an
operating reactor system, The temperature con-
ditions to be employed in all three phases of
testing are shown in Table 2.3, The status of

23
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

 

 

 

UNCLASSIFIED
Y-24190

  
 
 
 
     
   

  

| INOR-8 |
HAYNES SP—16
(HT 8284) |

o

 

Fig. 2.8. Circular-Groove and Open-End Slot Weld-Cracking Test Specimens.

24
tests which have been conducted thus far under
these programs is described below,

Thermal-Convection Loop Tests

A second Inconel loop has completed 1000 hr
of operation under the phase-1 program and has
been examined metallographically, This second

UNCLASSIFIED
ORNL-LR-DWG 22383R

 

 

 

 

Fig. 2.9. Weld Test Plate Design.

 

PERIOD ENDING OCTOBER 31, 1957

loop circulated fuel mixture 123, NaF-BeF,-UF,
(53-46-1 mole %), and operated in accordance with
the temperature schedule given in Table 2.3
Hot-leg sections of the loop showed light-to-
moderate surface roughening and surface pits to
a depth of 0.00025 in., as shown in Fig. 2,16,
The appearance of the cold-leg specimens was
similar to that of the hot-leg specimens, and they
were entirely free from layers of metal crystals.
The results of chemical analyses of samples of
the fused salt taken before and after the test for
the metal constituents are given in Table 2.4,

Other Inconel thermal-convection loops are now
being operated to complete an initial evaluation
of all the salt mixtures listed in Table 2.2, An
INOR-8 loop is also being operated with the salt
mi xture NoF-Ber-UF4 (53-46-1 mole %), and
additional INOR-8 loops are being fabricated for
operation with the remaining fluoride mixtures

listed in Table 2.2.

UNCLASSIFIED
Y-24048

Fig. 2.10. NiMo Test Plote No. 36.

25
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

 

 

 

 

' _ " UNCLASSIFIED
L Y-24179
AL A L A o
. L . -
# " 4 :
[ 5]
&
- .'_ p :
" - .02
¥ ®
- ¢ - e -
/7 """ - -
» _‘ . - 4 =
) to T Foa
o ~ ; - oamm— | . =
- e ¢ . L N
s . / /
- e -~ . - - .
- i ’ o 1
» .
" —ad g ) T ) ; -
" g _“ o
- . 4 . . »
- - " » - " po—u
oA 2

Fig. 2.11, Weld-Metal Cracks in Haynes SP-16 Weld Metal Welded to SP-16 Base Plate. Etchant: chromic ocid
and HCI. 100X,

UNCLASSIFIED
Y-24180

,
e

INCHES

 

 

 

 

 

Fig. 2.12. Base-Metal Cracks in Haynes SP-16 Welded with SP-16 Weld Metal. Etchant: chromic acid and HCI.
100X.

26
PERIOD ENDING OCTOBER 31, 1957

UNCLASSIFIED
Y.-24049

 

 

1

Fig. 213, NiMo Test Plate No. 37.

. - . ' UNCLASSIFIED
z , Y.24182
, ' ,
. . {
i L i ' : >
/ ' . :
r . . v . *
\ .
8 \ . - ‘ oo
. . *
\'
- ‘fl-\‘. . -, . . v\‘! : ‘ .
. ~ 1 '
vy
‘ L . .
.
Cs \ )
. o’ o "\ ' ¥ «
. + T ‘ ‘ 1\ .
. v \ 1

. '
-8 . - Y

 

1 Y r
INCHES jo E I
1 1 o -

3

 

 

 

 

01

 

Ole

 

013

 

 

 

e

Fig. 2,14, Base-Metal Cracks in Haynes SP-16 Welded with Hastelloy W Filler Metal. Etchant: chromic acid

and HCL. 250X.

27
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

 

o : o 'UNCL ASSIFIED :
’ L o : : : | Y-24181
/ : ; . , .7 J :
/ ‘ Ir b ‘
;' ¥ ’ ” . -‘ .
/ ! /, A ‘e I s
P ! 4 ’
, ] ' ‘ ? { p " .
y i ff ,L;. .} A /] : Ee *
ey T R I N ! ' B3
' .‘b ] ' } [ ¥ v i b
f ' ok P 2
. v { ‘ ! -
! ' J v / . ' .
‘ i v 4 i
* {
l ; . . . . \ ; l. 3
v’, . .
i }‘i .., . ’ 4 %
. . . ; e 02
oo . . r
) A
1 - » v_
. ’ jo.o2
K ‘

 

 

Fig. 2.15. Fusion Line Porosity of Type Found When SP-16 Base Metal Was Welded with Either SP-16 or

Hastelloy W Filler Metal. Etchant: chromic acid and HCI.

Table 2.2, Compositions of Molten Salt Mixtures to be Tested for Corrosiveness in Inconel
and INOR-8 Thermal-Convection Loops

 

 

 

Salt Composition {mole %)
Designation NaF LiF KF ZrF4 BeF, UF, ThF4
Fuel and Blanket Salts
122 57 42 1
129 55,3 40.7 4
123 53 46 1
124 58 35 7
125 53 46 0.5 0.5
126 53 46 ]
127 58 35 7
128 71 29
130 62 37 1
131 60 36 4
Coolant Salts
12 1.5 42
84 27 38

 

28
Table 2.3, Operating Temperatures to be Used

in Yarious Phoses of Corrosion Testing

 

Operating Temperatures {°F)

 

Difference Between

 

Maximum Hot and Cold
Sections

Phase 1

Fuel mixtures 1250 170

Coolant salts 1125 125
Phase 2

Fuel mixtures 1250 to 1350 170

Coolant salts 1050 to 1250 170
Phase 3

Fuel mixtures 1300 200

Coolant salts 1200 200

 

Forced-Circulation Loop Tests

An Inconel forced-circulation loop has been in
operation for some time in order to determine the
corrosion rate of this alloy in a zirconium-base
fluoride and in sodium over a pericd of one year,
The loop design employed provides for the oper-
ation of separate sodium and fluoride circuits
connected through a U-bend heat exchanger, The
maximum fluoride interface temperature in this
foop is 1250°F, while the maximum sodium temper-
ature is 1150°F. The loop, as of October 15,
had accumulated 4900 hr of operation. Con-
struction of additional Inconel forced-circulation
loops, as well as INOR-8 forced-circulation loops,
is now under way for testing under the phase-3
program.

PERIOD ENDING OCTOBER 31, 1957

 

UNCLASSIFIED
T-13379.

 

 

 

 

 

 

 

 

 

 

 

 

 

 

st
[ 009
¢ 4
-
' I'-\,
- ) ) 10
e *
. L ‘ L. . \ - . -
Tl . Lo Lt ~
. . .C" A ;\\ " o 00
. S . . o
o T ' s * "
» ST . .
2 T ‘ 012
o O e A
L o TE T
- P -
s r
3 . 013
. . .
¥ v Sty :
%, . o .
T T L 014

 

Fig. 2.16. Hot Leg of Inconel Thermal-Convection
Loop 1163 Which Circulated Fuel Mixture 123, NaF-
BeF,-UF, (53-46-1 Mole %), for 1000 hr at a Maximum
Temperature of 1250°F. 250X.

Table 2.4, Metal Constituents of Fused Salt Before
and After Circulation in Loop 1163

 

Major Minor
Constituents

{ppm)

Constituents

(wt %)

 

U Be Ni Cr Fe

 

Before-test sample 3,52 8.03 75 375 185

After-test sample
Hot leg 3.43 8.32 30 465 180
Cold leg 3.50 8.32 30 435 180

 

29
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

RADIATION DAMAGE

G. W. Keilholtz

IN-P{LE LOOP TESTS

W. E. Browning R. P, Shields
H. E. Robertson J. F. Krause

An in-pile thermal-convection loop fabricated of
an INOR-8 alloy is being prepared for operation in
the LITR with a lithium-beryllium-uranium fuel
mixture. The purpose of the experiment is to
demonstrate the reliability of this combination of
materials and to determine the effects of radiation
on corrosion of the container material by fluoride
fuel in the MSR design temperature range. The
fuel container will be examined metallographically
for evidence of corrosion after irradiation, ond an
analysis will be made of the fission products in
the fuel., Design features and operating con-
ditions for the loop are described below:

Fuel composition Ber-LiF-UF4 {37-62-1 mole %)
with Li” and U233

Power density 50 w/cm® or higher as necessary

to achieve desired temperature

differential
Maximum temperature  1250°F
Axial temperature 150°F

differential

Up to 10,000 hr

Duration of in-pile

operation

The loop will be air-cooled and will have
external heaters. It will also have a charcoal
trap for removing the xenon gas evolved from the
fuel. The mechanical details are similar to those
of the miniature, vertical, dynamic corrosion loop
recently operated in the LITR. This loop can be
converted to a forced-circulation loop with a
minimum of modification after suitable pump
bearings have been developed.

The parts which will be assembled to form the
loop are shown in Fig. 2.17. It may be noted that
the loop consists essentially of two parallel tubes
with connecting bends at the top and bottom. The
loop will be cooled by air passing through an air
annulus which will, in general, follow the loop
all the way around. The air will travel in the
same direction as the fuel.

30

UNCLASSIFIED
- PHOTO 41790

 

CHARCOAL
TRAP

  
  

    
 

LIPS v )

 

| — FUEL LEVEL

 

fo = i k] 8
AHCIYHOET™ TENOILTN

COOLING
AIR

Fig. 2.17. Parts for INOR-8 Thermal-Convection Loop
To Be Operated in the LITR.
Details of the design and planning of the con-
struction and operation of this loop have been
reported, ! The fuel tube is 30 in. long, % in, in
outside diameter, and will contain 42 ecm? of fuel.
The total power will be approximately 2 kw.

A full-scale mockup of the in-pile loop is being
prepared, with electrical resistance heating of
the fuel tube to simulate fission heating. The
purpose of this out-of-pile test will be to demon-
strate the workability of the design and to provide
operation information that will aid in an evaluation
of hazards in the reactor. The parts shown in
Fig. 2.17 have been fabricated for the bench test
and are being assembled. Practice welds have
been made for the fuel tube by using INOR-8 tubes
and INOR-8 welding rod. Fuel for the bench test
has been requested.

An improved design has been worked out for the
thermocouple leads in the cooling-air annulus that
should permit greater movement of the fuel tube
without plastic deformation of the thermocouple
lead wires. Preparations are being made to
conduct a thermocouple mechanical-mockup test
of this improved thermocouple installation tech-
nique.

Calculations have been made for the equilibrium
distribution of fission gases in the system. It is
estimated that 9 x 1074 moles of stable xenon
and 1.5 x 10~% moles of stable krypton will
accumulate during 10,000 hr of operation at 2 kw.
This amount of xenon and krypton will be absorbed
in the 500-g charcoal trap to such a degree that

 

]W. E. Browning, Proposed In-Pile Convection Loop
for Molten-Salt Reactor, ORNL CF-57-11-20 (in prepara-

tion).

PERIOD ENDING OCTOBER 31, 1957

the remaining partial pressure of xenon will be
600 u Hg and that of krypton will be 150 u Hg.
These equilibrium partial pressures are sufficiently
low for the adsorption efficiency of the charcoal
to be unimpaired. The combined decay-heat
generation of all the daughters of fission gases
having half-lives greater than 10 sec is only 6 w,
according to the calculations. Cooling of the
charcoal trap will not be a difficult problem.

Parts for the in-pile loop are being fabricated
and will be assembled ofter the bench test has
been started. Fluoride fuel containing enriched
U235 and Li7 has been requested for the in-pile
test, The INOR-8 tubing required for both the

bench and the in-pile apparatus is available,

STATIC IN-PILE CAPSULES
H. L. Hemphill

Preparations are being made to irradiate lithium-
beryllium-uranium fluoride fuels in INOR-8 cap-
sules in the MTR. Tests simulating in-pile
conditions have been conducted on INOR-8 tubing
Thermal cycling in
air at various temperatures up to 1250°F did not

W. E. Browning

suitable for use in capsules.

damage the adherent oxide coating which protects
INOR-8 from atmospheric corrosion. Further tests
are to be conducted in high-velocity air streams
similar to those that are used in-pile, and equip-
ment for testing the durability of thermocouples
attached to INOR-8 capsules is being prepared.
Fuel material has been requested for these tests.

31
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

CHEMISTRY
W.R. Grimes

PHASE EQUILIBRIUM STUDIES

C. J. Barton R. E. Moore
H. F. Friedman R. A. Strehlow
H. Insley R. E. Thoma

R. E. Meadows C. F. Weaver

An extensive and systematic study of phase
behavior in fluoride systems has been conducted
in the Chemistry Division of the Oak Ridge
National Laboratory. In recent months attention
has been directed toward those systems of
possible use as fuel solvents, fuels, breeder
blankets, and secondary coolants. Although the
final choice of a fuel system for even the simplest
of such reactor embodiments is far from certain,
it appears that the systems of most promise are
those which contain Bel:2 with LiF and/or NaF.

The present status of the knowledge of phase
behavior of these various systems is described
briefly here, and references are given to more
detailed reports.

Solvent Systems

Melting Point of BeF,. — Several investigators
have examined the melting point of BeF,, and
widely differing results have been obtained.
Quenching experiments by Roy, Roy, and Osborn?
established a value of 545°C, which was confirmed
by Eichelberger et al. of Mound Laboratory.
Recently, however, Kirkina, Novoselova, and
Simanov,3 who studied the system BaF,-BeF, by
thermal analysis and x-ray diffraction, stated that
BeF, crystallizes in a new modification that is
stable above 545°C when BaF, is present. They
claim that 545°C is o transition temperature
between the two forms of BeF, and cite the
existence of a eutectic at 600°C between BaBeF ,
and BeF, as proof.

Recent quenches of mixtures containing BeF,
and up to 5 mole % BaF, have not confirmed the
Russian work. When the samples were equilibrated
at and quenched from temperatures above 550°C,
only isotropic material and quench growth were

 

2p, M. Roy, R. Roy, and E. F. Osborn, J. Am Ceram
Soc. 33, 85 (1950).

3D. F. Kirkina, A. V. Novoselova, and Y. P. Simanov,
Zhur. Neorg., Khim. 1, 125 (1956).

32

found by petrographic examination. X-ray dif-
fraction of the specimens showed none of the
commonly observed form of BeF, but revealed the
existence of lines identical with those presented
by the Russian investigators for BaBeF,. The
ordinary variety of BeF, was found by petrographic
examination of quenches from below 550°C, and
x-ray diffraction revealed ordinary BeF, along
with the compound found at higher temperatures.
Since this material coexists in equilibrium with
BeF, below 550°C, it is, presumably, BaBeF
and it cannot be a high-temperature modification o?
BeF .. These experiments present additional
confirmation of the 545°C melting temperature of
BeF .

The System L.iF-BeF,. — The phase diagram
presented in Fig. 2.18 for the simple system
LiF-BeF2 represents data from Roy, Roy, and
Osborn; 4" Novoselova, Simanov, and Yarembash;?3
and the Oak Ridge National Laboratory. The
low cross sections and generally good nuclear
properties of BeF, and Li’F, along with the
very low liquidus temperatures available in
mixtures containing from 35 to 70 mole % BeF,,
make this system of obvious interest as a fuel
solvent or as a coolant.

The System NaF+BeF,. — The phase diagram for
the NaF-BeF, system, shown in Fig. 2.19, was
published by Roy, Roy, and Osborn.® The lowest
temperature in the system is below that in the
LiF-BeF2 system, but the liquidus temperature
rises more steeply as the BeF, concentration is
decreased.  Mixtures that contain between 40
and 65 mole % BeF, may be of interest to the
program.

The System NaF.LiF-BeF,. -~ A preliminary
and partially complete diagram of the ternary
system NaF-LiF-BeF, is shown as Fig. 2.20.
The ternary compound has been identified, with
a fair degree of certainty, as NaF.LiF.3BeF, by
petrographic and x-ray diffraction examinations of

 

4D. M. Roy, R. Roy, and E. F. Osborn, J. Am. Ceram.
Soc, 37, 300 (1954).

5A. V. Novoselova, Y. P. Simanov, and E. |. Yarem-
bash, Zhur. Fiz. Kbim, 26, 1244 (1952).

6D. M. Roy, R. Roy, and E. F. Osborn, J. Am. Ceram.
Soc. 36, 185 (1953).
PERIOD ENDING OCTOBER 31, 1957

UNCLASSIFIED
ORNL~LR-DWG 16426

 

 

 

 

 

 

   
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

900

800 —

700 | ————— —
)
£ 600 |—— LiF + L1QUID —‘fif
)
o
2
5 |
<
.
a
L 500 b —
L
-

400 ( LipBeFy . S

+LIQUID
R
LiF + Li,BeF,
Li,BeF, + BeF, (HIGH QUARTZ)
300 —————»———‘——— W — fi—?——‘Lh—?——jfi—v‘-T
o
5 ., o |
HeBefa LiBeFy + BeF, (HIGH QUARTZ)
: @ | |
200 (LiBeFy 4 [iBeF3 + BeFp (LOW QUARTZ) |
LiF 10 20 30 40 50 60 70 80 90 BeF;
BeF, {mole %)

Fig. 2.18. The System LiF-BeF,.

previously equilibrated quenched samples con-
taining 60 mole % BeF,-20 mole % LiF. The
compound melts incongruently to BeF, and liquid
at about 287°C. Only single-phase material was
found just below 287°C, but the solid-phase
reaction in which NaF.LiF.3BeF, is formed
proceeds very slowly at only a few degrees lower.
The primary phase field of the ternary compound
is small and lies near the composition 50 mole %

BeF2—25 mole % LiF.

Exominations of quenched samples of a number
of compositions containing more than 40 mole %
BeF, have led to the conclusion that the following
are compatibility triangles:

NaF.LiF-3BeF, ~ BeF, — 2LiF-BeF,
NaF.LiF-3BeF, — 2LiF.BeF, — 2NaF.LiF.2BeF,

NaF-LiF-BBeF2 - 2NoF-LiF-28eF2 - NOF-BeF2
l\lcnlz-Lil:-.'.’:BeF2 - BeF2 - Nc:F-BeI:2

Although exact determinations of ternary eutectics
and peritectics hove not yet been completed, it
is apparent that a region with melting points below
300°C exists in the neighborhood of the primary
phase field of NaF.LiF.3BeF,. Compositions
in this region (within 5 to 10 mole % of 50 mole %
BeF,-25 mole % LiF) may be of interest as low-
melting reactor coolants or fuel carriers.

Systems Containing UF4

The binary systems LiF-UF, and NaF-UF,

have been well known for some years.” Neither

 

7C. J. Barton et al., "*Phase Equilibria in the Systems
LiF—UF4 and NaF-UF4," J. Am. Ceram. Soc. (in press).

33
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

800

800

NaF + LIQUID

700

UNCLASSIFIED
ORNL-LR-DWG 16425

ORNL DATA
HIGH QUARTZ
= LOW QUARTZ

@-Na,BeF, + LIQUID

 

BeF,(HQ) + LIQUID

B'~NaBeF3
+ LIQUID

BeF,{HQ) + 5~ NaBefy

B-NaBeF3 + LIQUID
BeF,(HQ) + B-NoBefF3
BeF,(LQ) + 8- NaBefy

50 60 70 80 20 Bef,

Bef, (mole %)

Fig. 2.19. The System NaF-BeF,.

o
- 600
Lt
o
o
-
<
o
o
o
= 500
W
—
a- Na,BeF, + NaF
400
a -Na,BeF, + 8'~NaBeF;
300 ‘a’—NozBeF4 T NaF
B -NaBeF3 + y-No,BeF,
Y- F4 + NaF
200
NafF 10 20 30 40
UNCLASSIFIED
ORNL-LR-DWG 36151
E= EUTECTIC
T= TERNARY EULTECTIC Befo
TC= TERNARY COMPOUND 542

   

P= PERITECTIC

SUBSQOLIDUS COMPOUNDS SHOWN
IN PARENTHESES

LIQUIDUS TEMPERATURES ARE IN °C

{SNF-LiF- 3BeFy)
320

NaF- BeFy

LiF
B44

NoF
990

 

E649

Fig. 2.20, The System NaF-LiF-BeF,.

34

of these systems has sufficiently low liquidus
temperatures at low wuranium concentrations to
serve as a practical reactor fuel. Liquids suitably
rich in UF, for enriching an operating reactor are
available, however, with either system. The
NaF-LiF-UF , ternary system, although it shows
considerably lower liquidus temperatures than
those of either binary system, is likewise not
useful directly as a fuel.

The binary system BeF,-UF, shows no inter-
mediate compounds and has a eutectic at a concen-
tration of about 1 mole % UF, and a temperature
of 530°C. The fernary systems LiF-BeF,-UF,
and NaF-BeF,-UF ,, which have been examined in
detail at the Mound Laboratory, are shown in
Figs. 2.21 and 2.22. It is evident from inspection
of these systems that low-melting mixtures of
interest as reactor fuels are available over wide
composition ranges.
PERIOD ENDING OCTOBER 31, 1957

UNCLASSIFIED
MOUND LAB, NO.
56-11-29 (REV.)

   
 
  

ALL TEMPERATURES ARE IN °C

E = EUTECTIC
P = PERITECTIC LiF - 4UF,
UF,| = PRIMARY PHASE FIELD

 

 

LiF

Fig. 2.21. The System LiF-BeF,-UF ,.

35
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

  
 

NaF-2 UF,

7 NaF-6 UF,

SNaF-3UF,

£
/\
2 NuF-UF4/'

7 '
3NaF-UF,

A S S

NS

   
 

\ \ R
NoF £ / £
2NaF-BeF,

UNC LASSIFIED
MOUND LAB. NO.
56-11-30 (REV.)

AlLL TEMPERATURES ARE IN °C

£ = EUTECTIC
PERITECTIC
PRIMARY PHASE FIELD

A
il

 

NaF-BeF,

Fig. 2.22. The System NdF-Ber-UF4.

Systems Containing 'I"|1F4

A number of ThF ,-containing systems are being
actively investigated, but virtually complete phase
diagrams can be+ presented at present for only

the BeF ,-ThF ,, LiF-ThF ,, and NaF-ThF , systems.
The System BeF,-ThF,, - The BeF,-ThF,

binary, shown in Fig. 2.23 is very similar to the
BeF ,-UF , system. A single eutectic, that melts
at 530°C, was found which contains 1.5 mole %
ThF4. No intermediate compounds are formed in
the system.

36

The System LiF-ThF,. — Four intermediate
compounds occur between LiF and ThF . The
compound 3LiF.ThF, melts congruently at 580°C,
forms eutectics with LiF and 7LiF-6ThF ,, re-
spectively, at 570°C with 22 mole % ThF, and
at 560°C with 29 mole % ThF,. The compounds
7LiF-6ThF, LiF-2ThF,, and Lil:nfi']'hfi4 melt in-
congruently at 595, 755, and 890°C. The phase
diagram, which is presented in Fig. 2.24, shows
the LiF-ThF, system to have somewhat higher
liquidus temperatures than those of its uranium-
containing counterpart. A

narrow range of
TEMPERATURE (°C)

TEMPERATURE (°C)

1200
1100
1000
200
800
700
600

500
BeF,

1100

1000

900

800

700

600

500

400

10

20 30

40 50

60
ThF, (mole %)

70

UNCLASSIFIED
ORNL-LR-DWG 24554

 

80 90

Fig. 2.23. The System BeF,-ThF ,.

ThE,

PERIOD ENDING OCTOBER 31, 1957

LiF-ThF, compositions melt below 1100°F,
however, and such mixtures would seem, in
principle, to be possible breeder blanket materials.

The System NaF+ThF,. — The phase diagram of
the NaF-ThF, system obtained from studies in
this Laboratory and shown in Fig. 2.25 differs
in several respects from one recently described
by Emelyanov and Eustyukhin.8 As may be seen
in Fig. 2.25, Nan-ThF4 melts incongruently to
NaF2ThF, and liquid, and there is only one
crystalline form of 2NaF.ThF,. The Russian
diagram, in which the liquidus temperatures are
in general similar to those of Fig. 2.25, shows
NaF.ThF, to be a congruent compound and gives

 

By, s. Emelyanov and Evstyukhin, J. Nuclear Energy
5, 108-114 (1957).

UNCLASSIFIED
ORNL—LR—DWG 26535

 

//

 

-

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

\ /
A\

s <

L_g’ = 'E L.g —

= © & &

W e L 0

: : 50

LiF 10 20 30 40 50 60 70 80 90 ThF,

ThE, (mole %)

Fig. 2.24. The Sys"em LiF-ThF4l

37
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

UNCLASSIFIED
ORNL—-LR-DWG 18964

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

1100 /
1000 //
oo //
W
7~
% 800 /
=
x
w
a /
= i
= 700 \ ,/ \[
l
600 | ' -1
U_cr
= ‘ o+
500 — L <+ ! U
. = [ -
= — = r—-
S L : >
< o Ll6 Llc-
(% =z =z
400 & ‘
NaF 10 20 30 40 50 60 70 80 20 Thf,
ThF, (male %)

Fig. 2.25. The System NaF-ThF .

two crystalline forms of 2NaF-ThF ,. The ORNL
diagram  (Fig. 2.25) shows that 4NaF.ThF,
decomposes at 568°C. It appears that the Russian
workers have interpreted these thermal effects as
a transition between two forms of 2NaF.ThF ,,
which were suggested by Zachariasen;? one of
these two forms of 2NaF:ThF, which may be
prepared by Zachariasen's technique, has been
shown in this Laboratory to be metastable.

The System ThF,.UF,, — The compounds ThF,
and UF, form o complete solid solution series for
which liquidus and solidus temperatures are not
yet completely defined. It appears that the
minimum melting point occurs at a concentration of
60 mole % UF4 and a temperature of 1010°C.

Ternary Systems Containing ThF4. — None of
the ternary systems containing ThF, can be

 

9W. H. Zachariasen, J. Am. Chem Soc. 70, 2147
(1948).

38

considered to be complete, but sufficient study of
the systems has been made to support the following
statements. The system LiF-ThF4-UF4 has
neither ternary compounds nor ternary eutectics.
Most of the area on the phase diagram is occupied
by the primary phase fields of the ThF -UF,
the LiF.4ThF ,-LiF.4UF ,, and the 7LiF.6ThF ,-
7LiF-6UF, solid solutions. Liquidus temper-
atures, in general, decrease toward the LiF-UF,
edge of the diagram.

The LiF-BeF,-ThF, and the NaF-BeF,-ThF,
systems are remarkably similar to their UF ,-
containing analogs. In the LiF-BeF -UF , system,
for example, two ternary eutectics occur that melt
at 350 and 435°C and contain 1 mole % UF ,-52
mole % BeF2 and 8 mole % UF4--26 mole % BeFZ,
respectively, and the ternary eutectics for the
ThF ,-containing system are nearly identical in
composition and temperature. The very great
similarity of the LiF-BeF,.UF, and LiF-BeF,-
ThF , ternary systems indicates that a relatively
wide choice of useful breeder-fuel compositions
should be availablie,

Systems Containing Plutonium Fluorides

The available thermodynamic data suggest that
PuF ; should be more stable toward oxidation and
disproportionation than UF , is at elevated temper-
atures.  Conversely, F’UF4 is more strongly
oxidizing and probably is more corrosive than
UF,. Previous studies indicated that the solubility
of UF, in BeF ,-containing melts was less than
2 wt % at 600°C, and similar low solubilities were
anticipated for PuF,. [t seemed desirable,
therefore, to determine initially the solubility of
PuF , in mixtures of interest for long-lived power
reactors,

The first solubility determinations were made
with the 57 mole % NaF—-43 mole % Bef:2 mixture,
which is reported to melt at 360°C (ref 10). The
experiments were performed in a stainless steel
glove box supplied with low-humidity air at a
pressure slightly below that of the laboratory.
Nickel filtration apparatus similar to that de-
veloped for obtaining UF3 solubility data was
used, but it was slightly modified to permit use of
smaller amounts (6 g) of fused salts. Mixtures
prepared by holding weighed portions of purified
solvent and PuF_ in a mixed atmosphere of argon,
hydrogen, and HF for about 1 hr at the desired
temperature were filtered by application of vacuum
and inert-gas pressure. The nickel filter, which
was welded into the end of a long piece of :?B-in.
nickel tubing, was kept out of the melt prior to the
actual filtration operation. The data obtained in
the first experiments are shown in Table 2,5,

 

105, M. Roy, R. Roy, and E. F. Osborn, J. Am
Ceram. Soc. 36, 185 (1953).

Table 2.5, Solubility of PuF, in NaF-BeF2
(57-43 mole %)

 

 

 

Temperature Filtrate Analysis
(°C) Py (wt %) PuF 4 (mole %)
538 1.10 0.20
600 1.32 0.25
652 2.16 0.41

 

PERIOD ENDING OCTOBER 31, 1957

The result obtained ot 600°C was lower than
was expected from the results obtained at 538 and
652°C, and therefore the filtration was repeated at
600°C aofter a longer equilibration time. The
analysis of this filtrate and those obtained at the
same temperature with two other NaF-BeF , compo-
sttions (64-36 and 50-50 mole %) have not been
completed.

Several of the mixtures were also examined
spectrophotometrically for the presence of Pu4*
in dissolved portions of the fused salt, and nearly
all samples have been examined with a petrographic
microscope mounted in a glove box adjacent to
the stainless steel processing box. The plutonium
in all the samples examined to date appeared to be
in the trivalent state, but the plutonium compounds
present in the mixtures have not as yet been
positively identified.  The compound PuF, is
probably stable in contact with nickel under the
conditions of these experiments.

CHEMISTRY OF THE CORROSION PROCESS

F. F. Blankenship L. G. Overholser
R. B. Evans G. M. Watson

A reactor fueled with a molten salt and operated
at a high temperature will be free from corrosion
only if the system defined by the fuel and its
container is thermodynamically stable under oll
conditions of reactor operation. True thermo-
dynamic stability of such a system would not seem
to be realizable in practice. The rate at which
the system can attain equilibrium can be made to
be very slow for many cases, however, and reliable
long-term compatibility can be obtained through
the use of certain flyoride fuels and nickel-base
alloys, such as Inconel or the INOR alloys (de-
scribed above under the heading ‘‘Metallurgy’’),
which contain molybdenum and chromium. Since
most of the experience with molten-salt corrosion
behavior has been gained with Inconel, this
material will be used as the example in the
following discussion.

When Inconel is exposed at a single uniform
temperature to a molten UF | solution, selective
leaching of chromium from the alloy occurs. Re-
actions with traces of oxidants, such as FeF,
or NiF,, that were originally present in the melt
or were formed by reactions of the melt with oxides
on the alloy are partially responsible. A more
important contributor to the attack is the attainment

39
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

of the equilibrium

2UF, + &r—=—0CrF, + 2UF,

In simple isothermal tests no chromium is removed
from the metal after chemical equilibrium is
attained; diffusion within the alloy presumably
proceeds until the chromium activity is equal
throughout the metal wall. Such attack can be
avoided in principle and coan be minimized in
practice by careful purification of the melt,
deoxidation of the alloy, and the addition of UF,
and CrF, at equilibrium concentrations to the
melt.

Even when the required purification, deoxidation,
and pre-equilibration are perfectly attained, how-
ever, a corrosion mechanism still exists when
UF4-contoining solutions are circulated, as in a
reactor, through Inconel systems which contain
temperature gradients. If the iron and nickel
constituents of Inconel are considered to be inert
diluents for chromium (as is true to a reasonable
approximation), the long-term corrosion process
can be described simply. When the fuel mixture
is circulated rapidly through the polythermal
metal system, uniform concentrations of UF,,
UF,;, and CrF, are maintained throughout the
fluid. The concentrations are those which will
satisfy the equilibrium constant

VCer’VfiFs NCer'NaF

Ka = Ky‘KN = -

3

VCr'YElF4 NeoNoe

at N., of the unaltered alloy and at an inter-
mediate temperature which depends on the geometry
and the temperature profile of the system. In
this expression, K_ is the equilibrium constant
based on free-energy data, K_ is the activity
coefficient quotient, Ky is the equilibrium concen-
tration quotient based on mole fractions, y is the
activity coefficient based on pure substances as
the standard state, and N is mole fraction. The
system must, however, at ultimate equilibrium,
satisfy K_ at all temperatures and locations in
the system with the fixed concentrations of CrF,,
UF3, and UF . Since Ky for the reaction shown
increases with increasing temperature, the
chromium concentration in the alloy surface is
diminished at high temperatures and is augmented
at low temperatures. These surface concentrations
are maintained at the values required by the

40

equations by diffusion of chromium from or into
the bulk metal. The equilibrium concentration of
chromium at the surface of the alloy is accordingly
fixed by the temperature drop in the circuit and
the temperature dependence of the equilibrium
constant for the chemical reaction. The rate at
which chromium is transferred from hot to cold
sections of the apparatus is determined by the
rate of chromium diffusion in the alloy; this rate
depends strongly on the absolute temperature of
the metal. The flow rate of the fluid cannot offect
the rate of this diffusion-controlied process.

If the equilibrium constant changes sufficiently
with temperature, the temperature drop is suf-
ficiently large, and the activity of chromium
in the alloy is sufficiently high, the activity of
chromium necessary to satisfy the equilibrium
constant at low temperatures rises to near unity,
and crystals of chromium can form in the low-
temperature regions of the system at the expense
of depletion of chromium in high-temperature
regions of the system. Inconel systems with
high- and low-temperature regions at 1600 and
1200°F in which NoK-KF-LiF-UF , mixtures are
circulated and Hastelloy-X systems operated at
similar temperatures with the fuel mixture NaF-
ZrF ,-UF , belong to the class of systems in which
chromium deposits occur. The deposited dendritic
crystals of chromium tend to plug the equipment at
a rate controlled by diffusion of chromium from
high-temperature regions of the system. Such
combinations are obviously unsuited for reactor
application.

If, however, the temperature effect on the
equilibrium constant is small, as is true in many
ZrF ,-containing systems, the equilibrium is
satisfied at all useful temperatures without the
formation of crystolline chromium deposits. The
corrosion occurs, without restricting the flow
passages, at a rate controlled by diffusion of
chromium from the surfaces into the interior of the
cool region,

Measurements of the rate of loss of radiotracer
chromium from on inert fluoride melt exposed to
Inconel showed the over-all diffusion rate for
chromium into Inconel to be 3.7 x 10~15 ecm?/sec
at 700°C; the activation energy!! of the process
is 66 kcal/g-at. These numbers are, of course,
independent of the fuel used.

 

”G. M. Watson and R. B, Evans {unpublished data).
Relatively little information is available re-
garding the effect of temperature on the equilibrium
constant in BeF,-containing systems, but pre-
liminary values obtained for an LiF-BeF, (48-52
mole %) solvent are compared in Table 2.6 with
those for Nc:F-ZrF4 solvents for which many
corrosion tests have been made. The data indicate
that the temperature dependence of K, is sig-
nificantly greater for LiF-BeF, systems than for
the NaF-ZrF, systems which do not result in
chromium deposition in cool zones of dynamic
systems. |t appears virtually certain, however,
that with small chromium activities, such as those
in Inconel and the INOR alloys, and small temper-
ature drops, such as those contemplated for most
long-term power reactors, chromium deposition will
not result if fuel mixtures based on the BeF,-
containing system are used. Investigations of
the effect of fuel composition on the response of
Ky with temperature in BeF,-containing systems
of reactor interest are continuing.

FISSION-PRODUCT BEHAVIOR

J. H. Shaffer R. A. Strehlow
N. V. Smith W. T. Ward
G. M. Watson

When fission of uranium occurs in a molten
fluoride solution, the fission fragments must
originate in energy states and ionization levels

PERIOD ENDING OCTOBER 31, 1957

very far from those normally encountered. These
fragments must, however, quickly lose energy as a
consequence of collisions in the melt and come to
equilibrium as common chemical entities. The
valence states which they will ultimately assume
are determined by the requirements of redox
equilibria among components of the melt and the
alloy container and by the necessity for cation-
anion equivalence in the melt.

The equilibrium valences of several of the
fission products are not yet well established, but
it seems certain that xenon and krypton will
appear as elements and that the alkalies, the
alkaline earths, zirconium, and the rare earths
will appear as ions of their customary valence.
A systematic study of the solubility of these
materials in NaF-ZrF, and some other mixtures
has been made, and the status of knowledge
concerning behavior of these materials is summa-
rized below,

Solubility of the Noble Gases

The solubilities of the noble gases in molten
NaF-ZrF, (53-47 mole %) obey Henry’s law,
increase with increasing temperature, decrease
with increasing atomic radius, and are not
appreciably affected by the presence of 10 wt %
UF,. The Henry’s law constants at 600°C (in
moles of gas per cubic centimeter of NaF-ZrF,

Table 2.6. Equilibrium Quotients for the Reaction ZUF4(a’) + Cr{c) fi ZUFs(d) + Cer(d) in Molten Fluorides

 

 

Reaction Medium Temp:rdture Initial UF4 Concentration KN

(- C) (mole %)

NaF-ZrF , (53-47 mole %) 600 4.0 (1.3 + 0.1) x 10~4
800 4.0 (2.9 +0.1) x 104

NaF-ZrF , (50-50 mole %) 600 4.1 (3.2 +0.8) x 10~4
800 4.1 (5.9 + 1.8) x 104

NaF-ZrF , (59-41 mole %) 600 3.7 (1.5 +0.2) x 10-3
800 3.7 (1.7 + 0.7) x 10~°

LiF-BeF, (48-52 mole %) 550 1.5 2 x 10~6+
650 1.5 1.9 x 10=5+
800 1.5 7 x 10~

 

*Preliminary values.

4]
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

mixture per atmosphere) are 21.6 x 10-8,
11.3 x 108 5.1 x 10-8 and 1.9 x 10-8,
respectively, for helium, neon, argon, and xenon.
The heats of solution are, in the same order,
6.2, 8.0, 8.2, and 11.1 kcal/mole. The increase
in solubility with increasing temperature is in
marked contrast to the behavior of HF in this
solvent; the Henry's law constant for HF at 600°C
is 1.23 x 10=3, while the heat of solution is
—4.5 kcal /mole.

The solubility of krypton has not been examined,
but there is little reason to doubt that it dissolves
in NaF-ZrF, (53-47 mole %) at 600°C to the
extent of 3 £1 x 10~8 moles/cm3/atm with a
heat of solution of 9.5 £ 1.5 kcal/mole.

Although experimental determinations are under
way, no data have been obtained as yet regarding
the solubility of the noble gases in BeF,-con-
taining systems. It is anticipated that the
solubility of xenon will depend to some extent on
solvent composition, (Helium is roughly half
as soluble in the LiF-NaF-KF eutectic as in the
NqF-ZrF4 mixture described above.) However,
it is most unlikely that the solubility will be so
different from those listed above as to prevent the
removal of krypton and xenon from an operating
reactor fueled with a BeF ,-containing mixture.

Solubility of Rare-Earth Fluorides

The rare-earth trifluorides are sparingly soluble
in NaF-ZrF, (53-47 mole %) mixtures. In this
solvent the solubility of CeF, is 2.3 mole % at
600°C and 4.3 mole % at 800°C. The product of
the solubility (in mole %) and the third power of
the cationic radius is constant to within +3% for
CeF, LaF, and SmF,. While the solubility
changes with the NaF-to-ZtF , ratio, the saturating
phase is apparently the simple rare-earth fluoride
- (that is, CeF,) over a relatively wide NaF-ZrF
concentration range.

The solid phase in equilibrium with a saturated
solution of two rare-earth fluorides in the NaF-ZrF
(53-47 mole %) mixture is a nearly ideal SOFié

 

solution of the rare-earth fluorides. The equality
N x N s0
C°F3(d) L°F3(.s's) CoF3
LaF X Neor SE F
3(d) *T3(ss) |

{(where N is mole fraction, the subscripts d and ss
refer to liquid and solid solution, and $9 is
solubility of the single fluoride) is valid at each

42

temperature. This behavior of rare-earth mixtures
indicates that partition of fission-product rare
earths between the liquid and solid CeF_ in a
cooled external circuit should deplete the ?uel of
nuclearly harmful species. The reactor fuel would,
of necessity, be saturated with CeF, at the
reprocessing temperature. |f the fuel were
NaF-ZrF4-UF4, considerable CeF, (perhaps
2 mole %) would be carried in the circulating
stream.

Preliminary experiments on the solubility of
CeF, in an NaF-BeF, (57-43 mole %) mixture
produced the data presented in Table 2.7. A
comparison of the values for the NaF-BeF, and
NaF-ZrF, mixtures, as read from smooth curves
through the experimental points, is given in
Table 2.8, It is obvious that the solubility of
CeF, is considerably less in the BeF ,-containing
mixture and that it is more temperature sensitive.
The saturating phase has been shown to be the
simple trifluoride CeF,.

No data have yet been obtained with mixtures
of rare earths, but it is anticipated that they will

Table 27. Solubility of CeF, in NaF-BeF,
(53-47 mole %)

 

 

 

Temperature Solubility of CeF
°0) Ce™™ (Wt %) CeFy (mole %)
718 2.58 0.83
639 1.29 0.41
517 0.39 0.12
434 0.15 0.047

 

Table 28. Comparison of Solubility of CeF3
in NuF-ZrF4 (53-47 mole %} and in NuF-Ber
(53-47 mole %)

 

 

 

TempeerUre SO'UbIIH’Y (mole % CEF3)
(°C)
In NaF-ZcF In NaF-BeF,
750 3.2 1.1
700 2.7 0.65
600 1.9 0.29
550 1.7 0.17

 
exhibit the nearly ideal solid solution behavior
found in the NaF-ZrF, system. Such behavior,
along with the much reduced solubility in the
BeF, system, should make the fission-product
partition process quite attractive. Further studies
with NaF-BeF, (53-47 mole %) and other BeF,-

bearing systems are under way.

PRODUCTION OF PURIFIED MOLTEN SALTS

J. P. Blakely C. R. Croft
F.A. Doss

During the past quarter, 23 batches (described
in Table 2.9), including eleven different compo-
sitions, were prepared. It is likely that some
increase in process efficiency or in equipment
capacity will become necessary if the demand for
these materials continues as at present.

METHODS FOR PURIFICATION
OF MOLTEN SALTS

J. P. Blakely C. R. Croft
F. A. Doss
Production efforts have been devoted, in-

creasingly, to the preparation in pure form of the

various BeF ,-containing mixtures needed for

PERIOD ENDING OCTOBER 31, 1957

engineering tests in corrosion loops and similar
apparatus. The melts have been processed by
using slight modifications of the procedure used
to purify NaF-ZrF, and NaF-ZrF ,-UF , mixtures.
In this method the blended raw materials are
charged into reaction vessels of copper-lined
stainless steel that have capacities of 5, 10,
or 50 tb. The materials are melted and brought to
a temperature of 800°C under a flowing atmosphere
of HF to remove water with @ minimum of hy-
drolysis. The 800°C melt is sparged with dry
hydrogen to reduce penta- and hexavalent uranium
compounds, sulfate, and extraneous oxidants.
Subsequent sparging with anhydrous HF serves to
volatilize the HCl and H,S and to convert to
fluorides any oxygenated compounds in the mixture.
This HF treatment, however, contaminates the
melt with some copper fluoride through reaction
with the container metal. A final 15- to 24-hr
sparging with hydrogen serves to remove the
CuF,, along with any FeF, and NiF, that was
originally present in the melt, by reducing the
compound to metal,

The purified melt is then forced under helium
pressure through a transfer line that contains a
filter of sintered nickel and a sampler into a clean

Table 2.9, Materials Processed During the Quarter

 

 

 

Composition Constituents (mole %) Number
No. UF, ThF, BeF, LiF NaF KF ZeF, of Batches
C-12 46.5 1.5 42.0 3*
C-84 38.0 35.0 27.0 2%
C-122 1.0 57.0 42.0 1
C-123 1.0 46.0 53.0 2%
C-124 7.0 35.0 58.0 1*
C-125 0.5 0.5 46.0 53.0 2%
C-126 1.0 46.0 53.0 4*
C-127 7.0 35.0 58.0 1
Cc-128 29.0 71.0 3
C-129 4.0 55.3 40.7 3*
C-130 1.0 37.0 62.0 1
Total 23

 

*Includes one large batch (> 13 kg); all other batches were small (<6 kg).

43
nickel receiver vessel. By manipulation of
appropriate valves and connections the cold
receiver is detached from the production assembly
and attached for storage to a manifold containing
helium under a slight positive pressure.

The BeF ,-containing mixtures are substantially
more difficult to purify than ZrF ,-containing
melts. The difficulties are, in general, a conse-

44

quence of the fact that BeF, contains up to
2000 ppm S, as sulfate, and up to 600 ppm Fe,
probably as FeF;. The high sulfur content causes
numerous equipment failures, especially failures
of the nickel or nickel-alloy dip lines and transfer
lines. Studies of improvements which can be
made readily in the purification process are under
way.