ORNL-CF-60-6-97.txt

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OAK RIDGE NATIONAL LABORATORY
Operated by
UNION CARBIDE NUCLEAR COMPANY

Division of Union Carbide Corporation

uec

Post Office Box X
Oak Ridge, Tennessee

July 25, 1960

Molten-Salt Reactors: Report for
1960 Ten-Year-Plan Evaluation

R. W. Ritzmann
H. G. MacPherson

NOTICE

This document contains information of a preliminary nature
and was prepared primarily for internal use at the Odk Ridge
National Laboratory. It is subject to revision or correction
and therefore does not represent a final report. The information
is not fo be absiracted, reprinted or otherwise given public
dissemination without the approval of the ORNL patent branch,
Legal and Information Control Department.

DRAFT

/0 =c/w M

Distribution Limited %o
Recipients Indicated
External Transmittal
Authorized

ORNL
CENTRAL FILES NUMBER

60-6-97
Revision 1.

 

 

 

 

copY No. /& A
MOLTEN-SALT REACTCRS: REPORT FOR 196C TEN-YEAR-PIAN EVALUATICON

Classification

For purposes of this evaluation, the molten-salt reactor is considered
as an advanced concept. It is considered not to have a status of current tech-
nology adequate to allow the immediate construction of large-scale power plants,
since no power reactor has been built or even designed in detail. As a result
there can be no estimate of present cost of power;, and the projection of power
costs to later years is necessarily based on general arguments rather than

detailed considerations.

Definition of Concept

 

Molten-salt reactors utilize molten fluoride salts as the solvents for
both fuel and fertile materisl. Although the fluoride salts themselves have
about half the slowing-down power of graphite, and therefore the reactors may
be homogeneous with only self-moderation by the salt, most present-day designs
call for graphite as the modera?or, unclad, and in contact with the salt.
Typically, the salt will occupy>8 to 25 vol % of the reactor core, with the
graphite occupying the remainder. The metallic container material is a nickel-
‘mclybdenum-base alloy which is compatible with both the salt and the graphite
to IBOOOF° A large variety of reactor types can be constructed by using these
basic materials of construction and utilizing U-233, U-235, or Pu fuel and Th

or U-238 as fertile material.

Background

Molten-salt fuels were conceived originally as a means of satisfying
the requirements of very high temperature and extremely high power density

necessary for nuclear aircraft reactors.l A very large amount of work on the
physical, chemical, and engineering characteristics of uranium- and thorium-
bearing molten fluorides was carried out as a part of the ANP program. In
1954, the Aircraft Reactor Experiment was operated as a 2.5-Mw circulating-
fuel molten-salt reactor. More than 90,000 kwhr were logged during the planned
experimental program, with outlet temperatures as high as 1650°F.

Although the molten~salt fuels were originally proposed for aircraft
use, their potential usefulness for power reactors was recognized from the
start. The features that attracted initial attention for civilian use were
the high temperature of the fuel combined with its low vapor pressure, the
stability under radiation of the halide salts, and the usual features that a
fluid fuel provides. These include a negative temperature coefficient éf re-
activity, no limitation to burnup reafi%&ing from either radiation damage or
loss of reactivity, the absence of a complicated structure in the reactor

core, and the potential for a low-cost fuel cycle.

Reactor Types

Two general types of reactor are now considered most attractive for
power applications. One is a single-region graphite-mcderated reactor; the
other is a two~region reactor with a graphite-moderated core and a thorium-
bearing molten-salt blanket.

The single-region reactor is the simpler type and is probably most
suitable for small-sized power stations because it will be cheaper to construct
and operate. The physical size of such a reactor will vary from about h»l/E ft
in diameter and height for the lowest power leveis up to perhaps 12 £t in diameter
and height for a 300-MwE size. The internal structure of the reactor consists of

graphite bars separated by graphite spacers to provide fuel channels. The
3

molten-salt fuel, occupying from 8 to 25% of the reactor volume, flows longitudi-
nally in the slots that surrcund the bars and is pumped from the top of the reactor
through a heat exchanger and back to the refictor. The fuel enters the reactor tank
again in an annulus arcund the top and flows down to the bottom inside the reactor
vessel to cool it. All reactor core components are simple in geometry and easily
fabricated.

T

F and BeF2

point of about BBOGF. A typical fuel for the single-region reactor would contain

The basic fuel-solvent salt is a mixture of Ii having a melting
4 to 15 mole % ThFh, depending on reactor diameter, and less than 1.0 mole % UFh;
the maximum ThFh addition would increase the melting temperature to about 950°F.
The conversion ratioc of this reactor will vary with the frequency of chémical
reprocessing and with the physical size of the reactor, which in turn determines
the leakage of neutrons. TFor the conversion ratio to be as high as 1.0, the
reactor would have to be about 20 ft in diameter and 20 £t high. Such a large
reactor would be capable of producing more than 1000 MwE in a single unit, but
it is doubtful that such & large unit would be built toc achieve a conversiocn
ratio of unity when a much smaller two-region reactor will accomplish the same
purpose .

In the two-region reactor the fuel salt (which contains little, if any,
fertile material) passes through the reactor core in graphite tubes. In the core
region, the fuel tubes are surrounded by moderator graphite. A blanket salt con-
taining about 15 mole % ThFh surrounds the core region on all sides to a thickness
of about 30 in. A small amount of the blanket salt passes through the core region,

providing some internal conversion and also cocling the moderator graphite.
A two-region reactor with the core about 7 £t in diameter and 7 £t high
should be capable of generating 300 MwE. When this éore is surrounded by a *0-in.
molten-salt blanket on all sides, nearly all the neutrons are usefully absorbed
and conversion ratios in excess of 1.0 are possible. In comparing the single-
region and two-region reactors, the simpler construction and lower capital cost
of the former must be balanced against the better neutron economy of the latter.
In'large power installations, the high conversion ratio of the two-region reactor
results in lower fuel-cycle costs.

The heat transfer systems of both types of reactor are similar. Typically,
the fuel salt leaves the reactor at about lBOOoF, passes through a heat exchanger,
and returns to the reactor at llESoF, The preferred intermediate coolan% fluid
is a similar salt which contains no uranium or thorium. Again, typically, the
intermediate coolant would enter the heat exchanger at 10000F and leave at 1175°F.
The coolant salt would be used to genergte and superheat the steam. Under near-
optimum economic conditions, the steam throttle condition would be IOSQOF and
1800 psia, yielding a net fihermal efficiency for the plant of about 42%.

| Auxiliary equipment required for the fuel system includes drain tanks
for the fuel, an off-gas system to remove volatile fission products from the
fuel on a continuous basis, means of extracting fuel samples and adding enriched
fuel mixture, means of heating all equipment to the melting point of the fuel,
and preinstalled equipment for replacing failed components.

Two general types of layout for the primary-system equipment are being
considered. One is a so-called disjointed layout, with individual pieces such
as‘reactor, pump, and heat exchanger separated in space and joined by mechanical

flanges. The other type of layout is more compact, with reactor, pump, and heat
5

exchanger within a single large vessel. Although the two systems afe laid ocut
differently, maintenance equipment developed for one system is generally appli-
cable to the other. It is believed that early reactor expefiments will be of
the disjointed type to provide greater versatility, but that later power reactors

may be of the more compact arrangement to reduce capital costs.

Fuel Cxcle

Since enriched uranium can be added to the reactor during operation,
the only need for reprocessing the fuel is to remove nonvolatile fission-product
poisons. The amount of reprocessing that is done will depend on the neutron
economy desired, the type of reactor, and a balance of economic factors. With
a one-region reactor there will probably be a tendency to use the simpleét
possible fuel cycle. This will consist of leaving the fuel in the reactor for
a long period of time, with removal of fission-product gases as the only on-site
process. UF,4 would be added as necessary to replace burnup of fuel and to com-
pensate for fission-product poisons.  Barring accidental contamination of the
fuel, it would be left in the reactor until the cost of the inereased inventory
and burnup charges for uranium economically overbalanced the charges associated
with réplacing the fuel. In the reactor considered by the Fluid Fuels Task
Forcez, this period was about nine years for a 318-MwE (net) reactor. At the
end of this time, the fuel would be drained into sealed and cocled shipping
flasks and sent to some central processing plant for recovery of the uranium.
In the meantime, a new charge of salt and uranium would be installed in the
reactor.

On the other hand, two-region reactors would have a relatively simple

on-site chemical plant. The blanket process planned for this plant would be the
6

removal of the bred uranium by UTB volatility on a frequent schedule. The treat-
ment of the fuel salt would first involve removal of the uranium by the volatility
process and then treatment of the solvent salt to remove rare-earth fission pro-
ducts. The purified solvent salt would then be recombined with UFh and returned

to the reactor.

Justification for Pursuit of the Concept

Since there is no detailed design available and therefore no reliable
cost estimgte for a molten-salt reactor, the best justification for working on
this concept is based on rather general grounds. The Ad Hoc Advisory Committee
on Reactor Policy gave five criteria that should be met by a reactor that is to
achieve economic power, and the following discussion is directed to these points.

High Neutron Economy. -- As discussed in another section, molten-salt

 

reactors are capable of good performance as breeders. It appears that, even
when optimized for low power costs, two-region molten-salt reactors will have
conversion ratios in excess of 0.90.

Low Fuel-Cycle Cost. =-- Molten-salt reactors should have lower fuel-cycle

 

costs_than any other reactor for four reasons: (1) As in all liquid-fuel reactors
the burnup of the fuel is not limited either by radiation damage or by reactivity
loss, so that chemical reprocessing may be more infrequent than for solid-fuel
reactor systems. (2) A simple method exists for chemically reprocessing molten-
salt fuels. This consists of the fluoride volatility process combined with the
solution df the major constituents of the core salt in HF. (3) Reconstitution
of fuel and blanket involves only dissolving UFu or ThFh in the salt, with no

metallurgical, ceramic, or mechanical steps. (4) The thermal efficiency is high.
High Thermal Efficiency. =~ The molten-salt system delivers steam at

 

lOSOQF and 1800 psia, or higher if desired, and so fits modern econcmical
steam cycles.

High Power Density and Specific Power. -- The high specific heat of the
molten salts, combined with the large temperature range in which they can operate,
ylelds high specific powers. In two-region reactors specific powers of 1.2 MwE
ver kilogram of fissionable uranium in the entire reactor and chemical-processing
system are feasible. This is many times higher than that for a pressurized- or
boiling-water reactor, for example.

The fact that all the heat transfer is between liquids of good heat
transfer properties and that there is ample excess temperature of the heat source
makes the heat transfer equipment small physically and leads to the possibility
of compact reactors yielding very high power densities.

Simplicity and Reliability of Plant Design. -- There is disagreement
as to whether or not the moliten-salt reactors meet the objectives of simplicity
and reliability. The matter will be resolved in part by a reactor experiment.
Within the scope of this factor, as outlined by the Ad Hoc Committe, the molten
salt is open to question on the point of using expensive materials, INOR-8 and
fluoride salts. The total cost of these materials at their expected price
($3/1b for INOR-8 and about $2000/cu ft for the base salt) ranges from about
$10/kw to $£23/kw of installed capacity. Furthermore, the molten-salt reactor
system requires an intermediate heat transfer system, at least at the present
stage of the technology. Balanced against these factors tending to higher

capital costs are those of compact design mentioned above, the use of a
low-pressure primary system (no pressure shell); the omission of almost all
control rods for the reactor; the absence of a complicated internal reactor
structure; and the simplification of the required containment for a low-pressure

nonvolatile liquid system.

Power Generation Costs of g 300-MwE Plant

Although, as mentioned above, no detailed design of a power plant has
been made, the Fluid Fuels Task Force2 attempted to estimate in so faf as possible
the direct material and labor costs for a molten-salt power plant. Their basiec
numbers have been used in part A of Table I. Part B of Table I indicates changes
that are considered reasonable as a result of additional design studies made since
the Task Force met.

The Task Force reactor was a single-region reactor, whereas the reactor
in parts B and C of Table I is a two-region reactor, in which use is made of the
compactness possible in the molten-salt system. Compactness was emphasized in
the aircraft reactor work to reduce shielding requirements; it is used now to
reduce capital cost of the reactor system. A reactor core 7 ft in diameter by
7T £ high is capable of producing enough heat for a 300-MwE station; the reactor
core and blanket together will fit in a 13-ft-diameter reactor vessel. The fuel-
' salt-to-coolant heat exchanger may be arranged for vertical removal and placement
of the tube bundle; one arrangement would have the fuel pump and expansion volume
directly above the heat exchanger. With this arrangement, the maintenance opera-
tions become comparablé to those required for sodium-cooled reactors. The follow-
ing table gives a few pertinent size and complexity factors in comparison with
sodium-cooled reactors. The evident reduction in sizes of components resulits in
part from the fact that the molten salt has a volumetric heat capacity_h.E times

that of sodium, and in pért from a higher source temperature.
MSR5 , Sodium-Cooled Reactors
i Advanced
Hallam __P/604 Fermi _ Fast Reactor

Fuel tubes per MwE 0.17 58 28 140 264
Fuel tube ID, in. 3,75 0.50 0.65 0.148 0.122
Heat exchanger, sq ft

per MwE 41 93 170 160 92
Steam generator, super- 92 - 237 214 345 200

heater, and reheater
surface, sq ft per MwE

Coolant flow data (avg)
Bulk AT assumed, CF 175 338 275 250 350
Flow per MwE, gal/min 90 oL 263 318 214

 

As a result of this comparison it is believed that the reduction in build-
ing size claimed in note 4 of Table I is justified.

After the Task Force met, a partial cost estimate of a molten-salt repro-
cessing plant was made by Weinrich and Associates. Although a more complete cost
estimate is to be made by them soon, it appears probable fhat the fuel charges
shown in section B of Table I are at least adequatevfdr-a 00-MwE plant.

A comparison‘of the problems of maintaining a molten-salt reactor of
compact geometry with those of a fast sodium-cooled reactor would indicate little
difference in difficulty in so far as can be ascertained at this stage. Therefore
it was assumed rather arbitrarily in sections B and C of Table I that the operation
and mgintenance costs need be nc higher than for the fast reactpr.

| The reductions in capital costs shown in part C of Table I are fairly
obvious ones that might be expected after some experience with the operation of
such plants. The resultant projected power cost for a 300-MwE plant is about

6-1/2 mills/kwhr.
Table I. Molten-Salt Reactor Costs, %00 MwE (Net)

 

Costs Notes

 

A, Tluid Fuels Task Force Reactor (TID-8507)

Capital charges 5.9% mills/kwhr 1. Task Force estimates of direct material and labor;
ten-year plan schedule of indirect construction
costs and "land and land rights" charge. A main
transformer is added, and the plant is scaled to
300 Mw.

Fuel charges 2.95 mills/kwhr 2. Task Force number. Includes 1.54 mills/kwhr for
operation and capital charges on an arbitrarily
imposed $11,800,000 on-site chemical plant.

Operation, maintenance, and insurance 3. Task Force method of calcul?tion, including annual
charge for maintenance at 3% of total capital cost.
L.77 midls/khr Insurance according to ten-year plan schedule added.

 

total 10.65 mills/kwhr

B. With improvements in reactor design and plant
layout made since Task Force study; still
first-reactor basis.

Capital charges 5.29 mills/kwhr 4. (a) Replacement of Loeffler boiler system with
special design of combined boiler and super-
heater; savings - 0.3% mill/kwhr. See Task
Force report, p 47, note 2.

(b) Shift to compact two-region reactor design.
Together with 4(a), this reduces building
size by a large amount. Number of fuel and
coolant pumps reduced from eight to four,
and reduction in fuel piping; savings -

. 0.57 mill/kwhr.

(¢) Additional contingency due to first-plant basisg:
20% of "Reactor and Steam Generator Plant" adds
0.27 mill/kvwhr. -

ot
Table I.

(Continued)

 

Costs

Notes

 

B. (continued)

Fuel charges 1.22 mills/kvwhr

Operation, maintenance, and insurance
1.05 mills/kwhr

total 7.56 mills/kwhr

C. Future Costs

Capital charges 4.32 mills/kvhr

Fuel charges 1.22 mills/kwhr

Operation, maintenance, and insurance
1.02 mills/kwhr

 

total 6.56 mills/kwhr

7.

8.

Small batch-process on-site chemical plant;
cost estimate on plant derived from Weinrich
and Associates report. Conversion ratio 1.00.
Cost includes capital charges on chemical plant;

- operation of chemical plant; use charge on U;

capital charges on fuel and blanket salts.

Assumes same operation and maintenance as fast
breeder reactor.

(a) Predicted decrease in cost of INOR-8, with
volume use, from $6/1b average price to $3/1b
average price: 0.20 mill/kwhr.

(b) Elimination of hot cell for examination and
repair of failed components: 0.25 mill/kwhr.

(c) Reduce spare-parts inventory to 31,000,000;

0.25 mill/kwhr.
(d) Eliminate extra contingency: 0.27 mill/kvwhr.

Slightly decreased insurance due to lower
capital cost.

 

T
12

Breeding Potential

 

5

A recent comparative study” of thermal-breeder reactors has indicafied
that the molten-salt reactor is second to the agqueous homogeneous reactor in
breeding potential. An advanced type two-region molten-salt reactor is capable
of a conversion ratio of about 1.06 with a specific power of 1.2 MWE/kg of fuel,
and a doubiing time of 13 years. When operated to yield this performance, the
fuel-cycle cost is estimated at O.7 mill/kwhr, based on grouping reactors to

provide 1000-MwE capacity at a single site.

Status of Present Technology and Development Program

(1) Chemistry of the Salts. -- Phase studies have been made for a wide
variety of mixtures of fluoride salts,6 seeking those mixtures'having low melt-
ing points, low viscosity, low neutron absorption, low vapor pressure, and the
high chemical stability that prevents excessive corrosion. The system of greatest
interest involves the components LiTF, Bng, UFM’ and ThFh, and this system has
been thoroughly explored. Fortunately, UFu and ThFh behave similarly so that one
may be substituted for-the other with little change in properties. There is also
some potential interest in the substitution of NaF for IiF and Zth f‘or"il.‘hF)+ in
special reactor salt mixtures, and these systems are currently under continued

investigation. Sufficient studies7of the solubility of PuF._ have been made to

3
assure operability of a plutdnium«fueled reactor. Similarly, the solubility of

rare-earth fission products has been determined in a number of mixtures.

7

Solubility studies of noble gases in molten salts' have established an

adequate technology for the design of off-gas systems that will continuously remove
135

Xe135 from the reactor fuel. The removal of Xe was actually demonstrated by an
13

experiment carried out in the ARE.l‘ In connection with the purification method7
adopted for the salts, the solubility of HF has been established for several
salt compositions.

A subject of current research is the determination of the sensitivity
of the molten-salt mixtures to contamination with moisture and oxygen, and a
search for methods of reducing this sensitivity. When excess oxidfi appears in

the fuel, it selectively precipitates U0 Salts containing appreciable per-

2.
centages of ThFh or Zth are less susceptible than those containing only IiF,

BeF,., and UFh' Although calculations and experiments indicate that no trouble

2.’

will result from UO, precipitation if reasonable precautions are taken to keep

2
the cover-gas system pure, attention is focused on this problem because it is
the most troublesome chemical problem that is presently known.

(2) Physical Properties of the Molten Salts. -- The density at 1200°F
of the LiF-BeF,-UF) -ThF) salts ranges from 1.9 g/cc for a composition with no
heavy elements to about 3.5 g/cc for alsalt containing about 15 mole % of ThFh

and UFh combined. Specific heats have been measured accurately,9

and the product
of the specific heat and the dehsity does not wvary much with the variations in
composition to be expected in reactor fuels and blanket materials. Thus the heat
capacity per unit volume is about 1.0 cal/ecc-"C. Viscosities of the salts are
dependent on temperature, but generally range from 7 to 15 cp in the temperature
range of 1100 to 15000F, Thermal conductivities have not been measured for the
composition of greatest interest as yet, but the value presumed from other measure-
ments'" in salts is 2.5 Btu/br-£t-CF.

Heat transfer properties have been measured}l both with the salt on the

tube side and the salt on the shell side of salt-to-sodium heat exchangers. The

results correlate quite well with Reynolds number, showing small deviations from
1k

the Dittus-Boelter relationship. Méasurements of the stability of the film
coefficient as a function of time are under way. After 5000 hr, there has been
no significant trefid of the heat transfer characteristics.

(3) Metallurgy of the Container Alloy. -- The alloy INOR-8 (Hastelloy N)
was developedl2 té meet the special needs of corroéion resistance to fluoride
salts, high-temperature strength, resistance to air oxidation, and ease of fabri-
cation. It is a nickel-base alloy containing about 17% Mo and 7% Cr. The former
imparts high-tempersture strength and the latter good resistance to air oxidation.
Several production lots of the alloy have been made, and plate, tubing, rod, and
wire stock have been formed. The welding properties have been thoroughly explored
and found good. Casting techniques have been partially explored and the results
appear promising. A thorough examination of the mechanicsal properties has been
made, including creep data for times as long as 20,000 hr. Design-strength
criteria have been established based on creep strength, and the high-temperature
design strenéth lies between those of %04 and 316 stainless steel. Thermal
expansion coefficients have been established, and some thermal conductivity
measurements have been made. At present, the thermal‘conductivity at temperatures
from 900 to 15000F is being determined more accurately. Brazing alloys have been
developed for back brazing tube-to-header joints, and a fair amount of experience
has been gained in fabricating typical components. In general, the alloy has
been considered "commercial"” for more than a year.

(%) Compatibility of Salt and Comtainer Alloy. -- Early corrosion-loop
work, under the suspices of the ANP, with Inconell3 showed that chromium was the
alloy constituent most susceptible to attack and that it tended to be leached

from the hot portions of a corrosion loop and to appear in the tube walls of the
15

colder legs as a chromium-rich alloy. The chromium conecentration in the salt

does not become high enough to allow deposition of fiure dendritic chromium in

the cold leg. The depletion of chromium from the hot leg causes surface roughen-
ing and the appearance of subsurface voids. Thus the type of attack seen with
Inconel does not result in a reduction of tube-wall fihickness or in plugging of
tubes, but does result in a gradual'weakening of the metal in the high~temperature
regions.

INOR-8 was specifically developed to yield improved resistance to this
weakening attack, and the corrosion results that have been cbtained indicate
that it is at least a factor of 10 better than Inconel. An extensive testing
p»:c'og:t"aazmll+ has involved a number of salt compositions and a range of temperatures.
Nine thermal-convection loops have been run for one year, with peak temperatures
of 1250 and 1350°F and temperature differences in the loop of 200°F. TFive of
these showed no appreciable attack on the INOR-8, with a maximum depth of surface
roughening and pitting of 1/2 mil. The maximum depth of attack of any of the
1oops was 1—1/2 mils. DNo attack or deposits were found in the cold legs. Four
pumped loops have been examined to date, and at 13060F the average éttack rate
is less than l/h mil per year. OCne loop which had been repeatedly opened to
the atmosphere showed heavier attack, however, with pitting to a maximum depth
of 1-1/2 mils in 14,500 hr. In this loop there was also evidence of oxidation
~of the salt, and experience with this loop indicates the strong desirability of
keeping strong oxidants, such as the HF produced from HEO, out of the system.

In some of the loops that have operated a year or more, a thin adherent
film of a molybdenum-rich alloy has been observed in both hot and cold seections.

No adverse effects of this film have been detected.
16

A number of pumped corrosion test loops are still in operation to lock
for longer-term effects and té try tc ascertain just what the upper temperature
limit is for the system.

(5) Graphite in Molten Salts. -- About thirty different grades of graphite
have'been tested in a variety of molten fluoride-salt compositions. The graphite
is completely inert to the salts and is not wetted by it. Since graphite is porous,
the salt can be made to penetrate the graphite under pressure. The extent of pene-
tration is determined by the pressure applied and by the pore-size distribution of
the graphite.

At 150 psi and 15000F the salt penetrates a normal reactor-grade graphite
(AGOT) to the extent of about 1% vol %. So-called impervious grades are penetrated
to a lesser extent, and seven grades have been found with penetrations of less than
1 vol %. This latter level is expected to be entirely satisfactory‘for reactor use,
and sizes and shapes suitable for reactor construction can be obtained in these
grades. A bundle of rods‘of one of these grades was tested in a circulating-salt

loop at lBOOOF for one year.l5

There was no liquid penetration of the graphite,
no attack at all on the surface of the graphite, and no carburization of the INOR-8
loop. The graphite showed a 0.01% weight loss, explained by removal of adsorbed
gases, and showed an average pickup of 15 ppm of U and 93 ppm of Be as a result of
vapor transfer of UFu and BeFa.
A search has been made for fission-product-type additives that will make
the salt wet the graphite or that will cause attack on the graphite by intercala-
tion.l6 So far, no reasonable constituent of a fission-product mix has been found

that will do either of these things. Tests of this kind will be continued to

learn the effects on irradiated graphite.
17

All graphite normally contains adsorbed oxygen, and it has been found
that this will come off in the presence of a molten salt and, if the graphite-
to-salt volume ratio is excessive, will precipitate UO2 from the salt. The
oxygen can be cleaned off from the graphite by a 20-hr treatment with a flush

T and also by pretreatment with ammonium bifluoride (gas) at 1300°F.

saltl
Research on other ways for cleaning up the graphite is continuing, although

the flush-salt technique; required anyway to remove oxides from the metallic
partsTofwfihe system, seems.adeqfiafe. - N

(6) In-Pile Tests. -= A number of circulating-loop in-pile tests of
Inconel corrosion were made in the ANP program. In general, no acceleration
of corrosion was found to result from carrying out the tests uhder radiation,
and because of the nature of the corrosion process, none was expected. These
tests were carried out at high temperatures (1500 to 1650°F) where Inconel
undergoes reasonably rapid attack (7 to 10 mils/1000 hr), and so some attack
was always observed.

Two circulating-loop in-pile tests using INOR-8 at 12500F, each of about
fO0-hr duration; have been completed, and in neither case was any corrosive effect
found. |

Two small-capsule tests of graphite samples immersed in salt and encased
in INOR~8 have been run.18 These were taken to rather high burnups with about
5/h mole % U consumed. by fission. The graphite did not appear to be attacked
or damaged in any way. There was a slight indication of a lowered interfacial
tension for the salt to graphite, since two small holes that had been drilled
in the graphite became filled with salt in the in-pile test but did not in the

out-of-pile control test.
18

Two more capsule tests, each'involving four capsules containing molten-
salt fuel and graphite, have been irradiated and are awaiting hot-cell examination.
A third such test is planned, and it will probably conclfide the in-pile testing
prior to operation of an experimental reactor.

(7) Chemical Processing. -- The fluoride volatility processl9 is directly
adaptable to use with molten-salt fuels. In the case of molten salts, the dissolu-
tion Step using HF for dissoiving fuel elements is omitted. F

2
the salt, oxidizing UFh to UF,, which is swept out as a gas. The UF6 may be puri-

is bubbled through

fied by absorption in NaF beds and finally condensed as liquid UF6. The UF6 can
be reduced to UFlL in a flame reducer20 and then introduced into the fuel carrier
salt. In a continuous on-site chemical plant where complete decontamination is
not important, some of these steps can perhaps be eliminated, with reduction of
the UFB to UFh being effected as it is introduced into the salt. In any case,
the fluoride volatility process provides a way of readily recovering uranium
from a molten salt. As such, it is the only processing step required for a
blanket salt, but for a fuel salt it does not remove the fission products from
the carrier salt. Thus a separate process is required to purify the fuel carrier
salt from fission products, unless one is willing to throw away the carrier salt. |
A way of recovering the base salt and separating it from at least the
rare-earth fission products is to dissolve it in 90% HF.El On the basis of
laboratory tests, this method appears suitable for recovery of the LiF and BeFe,
but not any ThFh that may be present.

Another possible method of removing high~cross-section rare earths, that

of displacement with low-cross-section rare earths such as CeF
22

3 2 has been explored

in laboratory tests.
19

A different approach tc the chemical processing of molten salts is the
selective précipitatidn of certain ccnstituents. Particularly promising is the
selective precipitation of Pa and uranium from thorium-bearing fluoride salts
by additions cof BeO. These procedures will be exploited in future development
work, with particular emphasis on means of removing sclid precipitates from
molten-salt streams.

Although there are several promising lines for future development in
the chemical processing of melten salts, all cost studies tc date have been
made using only the flucride veclatility and HF disscluticn processes, both of
which have been fairly well explored chemically.

(8) Engineering Compcnents. -- Sump-type centrifugal pump323

ranging
in capacity from 2 tc 1500 gpm have been develcoped for circulating molten salts
at temperatures up to 1500°F. Eight different mcdels have been manufactured and
tested, and approximately 500,000 hr of non-nuclear operation have been accumulated
in the temperature range éf 1100 te 1500°F. The longest single pump test is still
in cperation after three years at 1200°F. All but one of these pumps have the
impeller at the end of an overhung shaft and utilize cil-lubricated bearings.
Iimited tests have been made on a pump employing a lower bearing that cperates
in and is lubricated by the salt. All of the pumps utilize an oil-lubricated
face~-type gas seal.

One of the cil-lubricated pumps has]had extensive radiatiocn testing,
with exposure te lOlo r of gamma radiation in the bearing and seal region. This
amount cf radiation did not interfere with the functioning of either the bearings
or the seal.

It is believed that no further development is needed for pumps with capaci-

ties tc 1500 gpm and having cil-lubricated bearings. Any punp manufactured for a
20

reactor must receive shakedown tests. Testing is also required for features
incorporated in the basic design for a specific reactor application, e.g., use

of the pump sump as system expansion tank and as the vessel from which to purge
evolved fission-product gases is a typical example of a special problem associated
with pump design.

Pumps having at least one salt-lubricated bearing offer the promise of
lower pressure of cover gas and possible simplification of some auxiliary systems;
but further development of thig type is required before it can be used in a reaétor.
No insuperable difficulties are foreseen in developing the technology of either of
these two types to larger sizes.

Pumps having bearings and drive-motor rotor submerged in molten salt
appear to offer the most versatile and simple solution for molten-salt reactor
application. This type would require development of salt-lubricated thrust bear-
ing and elevated temperature electric motor.

Nineteen heat exchangersah of 1/2- to 1-1/2-Mw capacity have been cperated
for an accumulated total of 25,000 hr with molten-salt temperstures ranging from
1100 to 1600°F. The operating programs included both steady-state operations and
continuous thermal cyecling between no load and full locad. The maximum operating
time at 12000F or above for a single unit was 2574 hr, and the maximum number of
thermal cycles from full load to no load was 194%. Heat-transfer correlations
and pressure-drop data were obtained for molten salt flowing both in the shell
side and tube side of the heat exchangers. No further developfient work is
expected to be required on salt-to-salt or salt-to-sodium heat exchangers.
Development work is needed, however, to provide steam generators heated by

salt. This is expected to be g fairly extensive program.
21

Many of the miscellaneous réactor parts such as enriching system, samplers,
and freeze valves will be designed for specific reactors, and they will require test-
ing and perhaps some development before reactor use.

25

A remote-mgintenance facility ~ incorporating a high-temperature (lBOOOF)
mockup of s 20-Mw (th) molten-salt-fueled reactor has.beenxconstructed. Techniq#es
and procedures have been developed for removing and replacing all major reactor
compdnents, including heat exchangers, the primary fuel pump, the reactor core
vessel, the fill and drain tank, and major piping éections. All maintenance
operations are performed by a single operator from a refiotely locgted control
cenfier, using closed-circuit television as the only means of viewing.

Maintenance is such an important aspect of molten-salt reactor work

that sdditional development work on maintenance techniques is desirable.

(9) Molten-Salt Reactor Experiment. -- Design for s molten-salt reactor

 

experiment was initiated‘about May 1, 1960. The reactor will be & single-region
graphite-moderated reactor, approximstely a right cylinder 4-1/2 ft in diameter

and 5—1/2 ft high. The container will be an INOR-8 vessel; the grgphite 8 fine-
grained extruded impervious type that should absorb less than 1.0 vol % of salt.
The fuel will occupy between 5 and 10% cf the reactor core volume.

The resctor, together with the fuel-circulating pump, the primary heat
exchanger, the off-gas system, and the preheating system, will be contained in
the existing 7503 building at ORNL. The facilities already available include
emergency power, cooling water, air blowers and a stack for the heat dump, a
control room, pits for drain tanks, an inert-gas supply, and overhead cranes.

It will be necessary to provide a new top cover for the reactor cell, so that

equipment for the maintenance of the radiocactive system can be installed.
22

The heat transfer system inciudes a secondarj salt loop, transferring
heat from the fuel salt to an air radiator. |

The major cbjective of the experiment is the determination of the depend-
ability, serviceability, and safety of a molten-salt reactor in so far as the small
size of the reactor will permit. The dependability will be determined by several
attempts to provide continuity of operation over periods of the order of four to
six months each. The serviceability will be determined by the success in keeping
the reactor operating during these periods with only short shutdown periods, and
by intentional component replacement programs between the long operating runs.
The safety will, of course, be indicated by the operating experience.

Many secondary objectives will be served by the reacfior. Most important
will be the opportunity of testing materials and components for long periods of
time under actual reactor conditions of radiation and fission-product generation.
Included will be tests of graphite and INOR-B in the reactor core. From the
behavior of graphite with respect to absorption of fission products, an important
point with respect to the feasibility of breeding in a molten-salt reactor can
be resolved.

Many auxiliary components such as the off-gas system, the sampling and
enriching devices, instruments, and auxiliary heating unite will be tested under
reactor conditions.

The fate of many fission products will be determined. This is particularly
important with respect to noble metals, since they are expected to plate out, and
it is important to know where this will occur.

Fuel salt contaminated with fission products will be available from the

reactor so that trials can be made of various chemical reprocessing methods.
25

Future Research and Development Costs

 

It must be recognized that research and development programs can vary
widely in the breadth and depth of the work and scope, and this variation will
be reflected in the degree to which the power reactor constructed has been opti-
mized. The program outlined here is believed to be on the frugal side, but should
nevertheless be adequate for reaching the goal of being prepared to build g 300-MwE
thermal-breeder plant with complete confidence that it will work, and with a good
background as to what the costs for it will be.

The program 1s outlined with respect to three reactor construction programs.
The proposed timing of these is rather close, with construction periods overlapping
slightly, in accordance with the guide lines expressed for the original ten-year-plan
study. The reactors propoged are as follows:

(a) Molten-Salt Reactor Experiment. -- This experiment is described

 

adequately in the preceding section.

(b) Two-Region Molten-Salt Reactor Experiment. ~-- The hext logical step
would be an experiment to demonstrate the features of a two-region molten-salt
concept. Impervious graphite would be used as the separation medium and thorium
as the fertile material in the blanket. The thermal power should be in the 30-
to 4O-Mw range, with the generation of electricity not a necessity. Possibilities
for breeding would be‘explored. |

(¢) Molten-Salt Reactor Prototype. -- The prototype size is visualized

 

as being in the range of 150 to 200 Mw of thermal power. Direct extrapolation
from this reactor prototype to a plant capable of 300 MwE should be possible
with a high degree of confidence, since most of the components of the prototype

would be of the full size required for the 300-MwE reactor.
Schedule for the Molten-Salt Program

24

 

 

 

 

 

 

 

Design | Reactor Program
Start  Criticality Completion
Molten-Salt Reactor May 1960 Mey 1963 May 1964
Experiment (5-10 Mw) (under way)
Molten-Salt Two-Region July 1961 Jan. 1965 July 1966
Experiment (40 Mw)
Molten-Salt Prototype Jan. 1963 Jan. 1967 Jan. 1969
(200 Mw)
Plant Construction Estimates Including
Research and Development Costs (millions of dollars)
Fiscal Year
61 62 63 64 65 66 67 69
*MSRE 1.5 2.2 1.0 1.0 0.2
Rand D 2.0 1.8 0.9
*TW‘O-Region 205 500 2.5 1'07 195 005
Experiment
Rand D 0.2 3.0 2.0 1.5 0.5
*Prototype 1.0 3.5 12.0 16.0 5.0 4.0 3.5 3.5
Rand D 2.0 3.5 3.5 2.0 2.0 2.0 2.0 1.5 0.5
3.7 12.5 15.9 20.5 20.4 8.5 6.5 5.5 5.0 0.5

 

* Includes operation costs.
No escalation factor included.
25
Summary of Molten-Salt Reactor Costs and Research and Development Costs
(millions of dollars)

_ Credit Total
Construction Operation for Power Net Cost

Reactor Costs

 

Molten-Salt Reactor Experiment 4,2 1.7 0 5.9

Two-Region Molten-Salt

Experiment 10.0 3.7 0 13.7

Molten-Salt Prototype 5.0 13.5% 18.2%x 30.3
49.9

Research and Development Costs

Molten-Salt Reactor Experiment h.7
Two-Region Molten-Salt Experiment _ . T.2
Molten-Salt Prototype 19.0
30.9

Grand Total 80.8

 

¥ Includes chemical plant operation.
*#% $200/kw capital cost plus 2 mills/kwhr fuel cost for two years.

 

Small-Size Molten-Salt Reactors

12.65-MvE (gross) Reactor. -- For a reactor as small as 12.65 MwE,

 

simplicity in both construction and in operation must be stressed. A molten-
salt reactor of this electrical capacity would have a thermal power of about
35 Mw, and this amount of power can be obtained without difficulty from the

minimum, practical size of reactor such as the one that is planned as a 5-Mw
experiment. The reactor is a single-region reactor about 4.5 ft in diameter

by 5.5 £t high. The flow of fuel salt would be the same as planned for the
26

experiment, 1250 gpm, but the inlet and outlet fuel temperature would be changed

to 1125 and 1300°F, respectively. Single pumps of 1250-gpm capacity would be used

in both the primary and secondary salt locps. In the primary system, the circuit

would be designed so that afterheat could be removed by natural convection in case

of sudden stoppage of the fuel pump.

direct contact, but the primary system would require remote maintenance.

The secondary system would be maintained by

Since a

specialized maintenance crew would probably not be available except on special

call, the primary system would be designed for maximum simplicity and reliability.

This means that there should be a minimum number of components in the primary system.

The fuel cycle should alsc have a minimum number of operations on-site to

achieve economy.

Thus the only fuel operations routinely carried out on the site

would be tending the off-gas system, adding enriched-UFu once a week, and probably

taking a weekly sample of the fuel for analysis.

With enriched-uranium additions,

the fuel salt should last from 10 to 20 years before it would be necessary to

replace it. The recovery of uranium from the replaced fuel would be accomplished

in an off-site flucride volatility plant.

40 to 50 kg.

The inventory of U-235 would be about

With a single-region reactor this small, there would be a high

neutron leakage so that the conversion ratio would be quite low-~in the neighborhood

of 0.25.

However, with the high thermal efficiency of this reactor, the fuel burnup

cost is relatively low, about 2 mills/kwhr. An approximate analysis of fuel-cycle

costs is given in the following table.

 

 

 

 

Unit Capital Rate Annual
Amount Value Investment per yr Cost mills/kwhr

Uranium 50 kg 17,000 850,000 4% 24,000 0.42
Fuel carrier salt .

with thorium Y0 cu £t 2,000 80,000 16% 12,800 0.16
Fuel burnup 9.15 kg 17,000 - - 155,000 1.92
Cost of recovery - - - , - 12,200 0.15

of U from spent total 2.65

fuel. Sinking fund
27

Whether a molten-salt reactof as small as 12.65 Mw has any legitimate
use depends most strongly on the capital costs and on the religbility of its
operation, since it is clegr that the fuel-cycle cost will not be excessive.
The MSRE will go a long way to answering these questions, but until the design
of the reactor is complete, it is impossible to estimate capital costs within
a factor of 2. Furthermore, any statement with respect to the reliability and
simplicity of operation can only be an opinion until operating experience with
the MSRE has been obtained. |

Lh-MvE (egross) Reactor. -- In this size range there are two possible

 

designs for a molten-salt reactor. One is essentially the same as that described
above for the 12.65 MWE size. The reactor would have a single region, with the
core volume and fuel-pumping rate increased in proportion to the reactor power.
Thus the reactor diameter and height would be about T ff, and the fuel would

be pumped at about 4500 gpm. With this type reactor the fuel handling would

be almost exactly the same as for‘the smaller reactor. The fuel-cycle cost
would be decreased by about 0.25 mill/kwhr as a result of the slightly higher
conversicn ratic. Capital and operating costs would be decreased, of course,

due tc the larger size.

The L4-MvwE size is large enough to allow consideration of a two-region
(breeder) reactor. Such a reactor would require a small on-site fluoride vola-
tility plant. Physically the reacter would be the same as that described for
the 300-MwE size, except that the core diameter would be about 5 £t instead of
T £, the blanket thickness would be trimmed to about 2 ft, and the fuel-managing
scheme would be simpler. |

The only on-site chemical processing visuvalized would be a small fluori-

nator and a small U’F’6-UFh flame reducer. This plant would be used to transfer
28

bred uranium from the blanket to the fuel circuit, and to strip uranium from the

fuel salt, which would then be discarded and replaced with fresh salt.

Both of

these processes would be carried out on a schedule that would provide one com-

plete treatment per year.

No special attempt would be made to achieve a high

conversion ratio, but a minimum-size volatility plant would probably be large

enough to achieve a conversion ratio of 1.0, and thus enable self-sustaining

operation with only thorium feed.

Whether it would be more economical to use a single-region reactor or

a two-region reactor in the 44-MwE size depends on the cost of construction and

operation of the minimum chemical plant.

It is reasonably pcssible that the

minimum chemical plant can be constructed for $2,000,000, and that it can be

operated for $300,000 per year.

Thus the chemical plant adds about 2 mills/kwhr

to the costs, while saving about the same amount in burnup costs. There are

increases in the other costs, however, caused by the presence of the blanket,

as indicated in the following table of approximate fuel-cycle costs for the

Li-MwE two-region reactor.

Uranium

Fuel and blanket
carrier with Th

Fuel-salt
replacement

Chemical plant
investment

Chemical plant
operation

 

| Unit Capital Rate Annual

Amount Value Investment per yr Cost mills/kwhr

100 kg 17,000 1,700,000 L4 67,000 0.23

550 cu ft 2,500 1,300,000 14% 193,000 0.67

B0 cu £t 2,000 - 100% 80,000 0.28

- - 2,000,000 14% 280,000 0.96

- - - - 300,000 1.03

total 5.17
29

Since the one-region reactor.of this size should have fuel-cycle costs
of about 2.4 mills/kwhr, it would appear that the two-region breeder is less
economical. However, it may be of interest to some to pay the premium of

approximately 0.8 mill/kwhr to obtain self-sustaining operating on thorium.
11.

12.

1.3.

1k.

15.

16. .

17.
18.

30
References

R. C. Briant et al., "The Aircraft Reactor Experiment,” Nuc. Sci. and Eng.
2, 797-853 (1957).

Report of the Fluid Fuel Reactors Task Force, TID-8507 (1959).

I. Spiewak and L. F. Parsly, Evaluation of External Holdup of Circulating-
Fuel Thermal Breeders as Related to Cost and Feasibility, ORNL CF-60-5-93
(1960).

S. Levy et al.; Advanced Design of a Sodium-Cooled Thermal Reactor for Power
Generation, 1948 Geneva Conference Paper P/60k.

L. G. Alexander et al., Preliminary Report on Thermal Breeder Reactor
Evaluation, ORNL CF-60-7-1 (July 1, 1960).

R. E. Thoma (ed.), Phase Diagrams of Nuclear Reactor Materials, ORNL-2548
(1959). | »

Reactor Chemistry Division Annual Progress Report, ORNIL-2931 (1960).

W. T. Ward et al., Solubility Relations Among Rare-Earth Fluorides in
Selected Molten Fluoride Solvents, ORNL-2749 (1959). -

H., W. Hoffman, MSR Quar. Prog. Rep. July 31, 1959, ORNL-2799, p 37-39.

S. I. Cohen et al., A Physical Property Summary for ANP Fluoride Mixtures,
ORNL-2150 (Declassified) (1956).

J. C. Amos et al., Preliminary Report on Fused Salt Mixture No. 130 Heat
Transfer Coefficient Test, ORNL CF-58-4-23 (1958).

 

Lane, MacPherson, Maslan (ed.), Fluid Fuel Reactors, Chap. 13, Addison-
Wesley (1958).

W. D. Manly et al., Aircraft Reactor Experiment — Metallurgical Aspects,
ORNI.-2349 (Declassified) (21957).

Metallurgy Division Annual Progress Report, ORNL-2839, p 1k7ff. (1959).
Reference T, p 69.

Reference 7, p T0.

MSR Quar. Prog. Rep. July 31, 1959, ORNL-2799, p 62-63.

Reference 7, » T4.
19.

20.

2l.

22.

23.

2k,

5,

31

References - (continued)

G. I. Cathers, "Uranium Recovery for Spent Fuel," Nuc. Sci. and Eng. 2,
768-777 (1957).

S. H. Smiley and D. C. Brater, "Conversion of Uranium Hexafluoride to
Uranium Tetrafluoride,” Chap. 4, Process Chemistry, vol II, Pergamon
Press (1958). -

D. 0. Campbell and G. I. Cathers, Processing of Molten Salt Power Reactor
Fuel, ORNL CF-59-2-61 (1959).

MSR Quar. Prog. Rep. April 30, 1959, ORNL-2723, p 81-82.

A. G. Grindell et al., "Development of Centrifugal Pumps for Operation
with Liquid Metals and Molten Salts at 1100-1500°F," Nuc. Sci. and Eng. 7,
83-91 (1960).

R. E. MacPherson et al., "Development Testing of Iiquid Metal and Molten
Salt Heat Exchangers,” Nuc. Sci. and Eng. 8, 14-20 (1960), A. P. Fraas,
"Design Precepts for High- Temperature Heat Exchangers," Nuec. Sci. and Eng.
8, 21-31 (1960)}; M. M, Yarosh, Evaluatlon of the Performance of Liquid
fiétal and Molten Salt Heat Exchangers, Nuc. Sci. and Eng. 8, 32-43 (1960).

 

W. B. McDonald and C. K. McGlothlan, Remote Maintenance Development for
Molten Salt Reactors, ORNL-2981 (in process).
1-5.

7-16.
17.
18.
19.
20.
21.
22,
23,
2k,

2.

57
58.
59.
60.
61.
62.

Distribution
W. Ritzmann, AEC, Washington 63.
F. Cope, AEC, ORO 64.
G. MacPherson 65.
G. Alexander 66.
E. Beall 67.
E. Bettis 68.
S. Bettis 69.
F. Blankenship TO.
Blumberg T1.
L. Boch 72.
. E. Bolt T3.
J. Borkowski Th.
F. Boudreau 5.
E. Boyd 76 .
J. Breeding e
. B. Briggs 78.
E. Browning 9.
0. Campbell 80.
L. Carlsen 81.
H. Carr 82.
. L. Carter 83.
I. Cathers 8.,
R. Chambers 85.
H. Chapman 86.
A, Charpie 87.
G. Cobb 88.
A. Cristy 89.
L., Culler 0.
A. Douglas 91.
K. Ergen g2.
P. Fraas 93,
H. Frye . ok,
H. Gabbard 95.
R. Gall 9%6.
B. Gallaher a7.
. R. Grimes 98.
G. Grindell Q9.
S. Harrill 100.
C. Hise 101.
. W. Hoffman 102.
. N. Howell 103.
Tnouye 104,
H. Jordan 105.
R. Kasten 106.
Jd. Kedl 107 .
W. Keilholtz 108.
T. Kelley 109.
. W. Kinyon 110.
. A. Lane 111-112.

32

M. I. Lundin
R. N. Lyon

W. D. Manly

E. R, Mann

L. A. Mann

W. B. McDonald
H. F. Mcbuffie
H. J. Metz

R. P. Milford
A, J. Miller
J. W, Miller
R. L. Moore

J. C. Moyers
D. J. Murphy
C. W. Nestor
T. E. Northup
W. R. Osborn
L. F. Parsly
P. Patriarca
H. R. Payne

A. M. Perry

R. M. Pierce
M. Richardson
Jd. T. Roberts
R. C. Robertson
D. Scott

M. J. Skinner
A. N. Smith

P. G. Smith

I. Spiewak

C. D. SBusano
J. A. Swartout
A. Taboada
R. E. Thoma

D. B. Trauger
W. C. Ulrich
F. C. Vonderlage
D. C. Watkin
G. M. Watson
A. M. Weinberg
g« He Westsik
J. C. White

G. D. Whitman
L. V. Wilson
C. E. Winters
C. H. Wodtke
J. Zasler
Laboratory Records, R. C.
Laboratory Records