- Email: [email protected]
Molten-salt converter reactors
Annals of Nuclear Energy. Vol. 2. pp. 809 to 818. Pergamon Press 1975. Printr,d in Northern Ireland
MOLTEN-SALT C O N V E R T E R REACTORS ALFRED M. PERRY Institute for Energy Analysis, Oak Ridge Associated Universities, Oak Ridge, Tennessee 37830, U.S.A.
Abstract--Molten-salt reactors appear to have substantial promise as advanced converters. Conversion ratios of 0.85-0.9 should be attainable with favorable fuel cycle costs, with ~asU valued at $12/g. An increase in "6U value by a factor of two or three ($10--$30/1b. U3Os, $75/SWU) would be expected to increase the optimum conversion ratio, but this has not been analyzed in detail. The processing necessary to recover uranium from the fuel salt has been partially demonstrated in the MSRE. The equipment for doing this would be located at the reactor, and there would be no reliance on an established recycle industry. Processing costs are expected to be quite low, and fuel cycle optimization depends primarily on inventory and bumup or replacement costs for the fuel and for the carrier salt. Significant development problems remain to be resolved for molten-salt reactors, notably the control of tritium and the elimination of intergranular cracking of HasteUoy-N in contact with tellurium. However, these problems appear to be amenable to solution. I think that it is appropriate to consider separating the development schedule for molten-salt reactors from that for the processing technology required for breeding. The Molten-Salt Converter Reactor should be a useful reactor in its own right and would carry us far towards the achievement of true breeding in thermal reactors. I. INTRODUCTION Molten-salt reactor technology had its beginning in the Aircraft Nuclear Propulsion program at .Oak Ridge. In the latter fifties attention shifted to civilian power applications, with emphasis on breeder reactors. The motivation for this was the classical one for fluid-fuel reactors, that is, the achievement o f good neutron economy by fast and (it was hoped) inexpensive on-stream processing of the liquid fuel to keep fission-product poisoning small. Over the years relatively little attention has been given to converter reactors, since we believed that the molten-salt reactor concept was uniquely suited for thermal b r ~ d i n g (Perry and Weinberg, 1972), and that has remained the principal motivation for its development. It was recognized, however, that development of the complex processing technology did not necessarily have to proceed on the same schedule as that of the reactor; and some consideration was given to the performance of reactors designed for breeding but lacking a facility for continuous on-stream processing. In this case, the fuel would be batch-processed at intervals to be chosen on the basis of an economic evaluation. A satisfactory procedure for this batch processing was successfully demonstrated in conjunction with the MoltenSalt Reactor Experiment (Haubenreich and Engle, 1970), and will be described later. N o special effort has been made to optimize a molten-salt converter reactor as such, but the breeder-without-processing 809
designs have nonetheless proved to have the potential for quite interesting performance. It is these reactors that are discussed in this paper. II. GENERAL DESCRIPTION In a molten-salt reactor the fuel consists of U F 4 or PuFa dissolved in fluorides of lithium, beryllium, sodium, potassium, zirconium, thorium, or some combination of these (Grimes, 1970) (Rosenthal et al., 1972). Fluorides of the metals other than U, Pu, or Th are used as diluents and to keep the melting point low enough for practical use (MacPherson, 1968). As a class, these salts are stable, have low vapor pressures at the temperature necessary to drive a m o d e m steam cycle, and have adequate solubilities for thorium, uranium, and plutonium; they are chemically compatible with a satisfactory moderating material, graphite and with a number of commercial nickel-based alloys. Other anions and other cations have been considered (Grimes, 1970); Rosenthal et al., 1972), but most of the development effort has centered on those mentioned. In the A N P program a mixture of N a F , Z r F 4 and UF4 was used, but in the subsequent Molten-Salt Breeder Reactor program, for reasons of neutron economy, the preferred salts have been L i F and BeFo., with the lithium enriched to 99-995 ?/o ~ in the 7Li isotope. (At this level, in a typical MSBR configuration, the remaining SLi absorbs 40 ~.{,as many neutrons as the 7Li.) Composition of the salt specified for the
A. M. l~gg¥
810
MSBR is LiF (72 mole %), BeF z (16~), ThF 4 12 9/o). UF4 would be dissolved in this salt with a concentration of about 0.3 mole ~ ; at equilibrium the uranium would contain "-aaU, z~U, za*U and zasU approximately in the proportions 65[22[6[7. Some pertinent properties of this salt are shown in Table 1, along with those of the salt that was used as fuel in the MSRE (Molten-Salt Reactor Experiment) (Haubenreich and Engle, 1970). Also shown are properties of some coolant salts for use in intermediate heat transfer loops between the primary fuel circuit and the steam generators (salt-to-air heat exchanger in the MSRE). Graphite is a suitable moderator for molten-salt reactors. Its moderating power is about twice that of the salt, while its effective macroscopic absorption cross section is about one-fifth that of the salt, exclusive of the actinides. Fuel salt will typically /o to 15 9/o of the core volume, the balance occupy 10 °/ being graphite. No metals are used in the core proper. It has been established that the salt does not wet the graphite, that if the graphite pores are small enough the salt will not enter them, owing to surface tension, and that a satisfactory distribution of pore sizes is readily attainable in commercial graphites (Rosenthal et aL, 1972; Scott and Eatherly, 1970). Fission-product gases, on the other hand, notably laSXe, can diffuse into the graphite (Scott and Eatherly, 1970), with a non-negligible effect on the neutron balance; more of this later. Graphite is, of course, subject to radiation damage (Rosenthal et aL, 1972; Scott and Eatherly, 1970). In particular, the isotropic graphites found suitable for the MSBR undergo significant volume changes, at first shrinking and then expanding. For
fast neutron fluences greater than about 3 × 1022 n/cm z (E n > 50 keV) the rate of expansion becomes quite rapid, and it appears that this presently represents an approximate upper limit to the acceptable exposure of the graphite. An approximate inverse relationship thus exists between the useful life of the graphite and the core power density. L x Pm"~200 where L is the moderator life in full-power years, and P,n is the maximum core power density in W/ c m s.
In the reference M s B R design, (Haubenreich and Engle, 1970; Robertson, 1971) the maximum power density is about 70 W/cm s, and the useful graphite life would be about three years at full power, or four years at 0.75 plant factor. The design must, therefore, allow for fairly frequent replacement of the moderator. Much larger cores have also been studied, with the maximum power density limited to about 8 W/ cm z, so that a useful core life of about 24 full power years might be achieved (based on 30 yr at 0.8 plant factor). The studies of converter reactors with batch processing cycles have been based mainly on these larger cores (Bauman, 1972). These cores are visualized essentially as larger versions of the MSRE in which the moderator consisted of 2-inch square bars of graphite arranged vertically in a regular rectangular array. Fuel salt flowed in passages between the bars, and spacing was maintained by machined ridges on the outer surfaces of the bars. In a larger reactor, it is likely that larger graphite assemblies would be used,
Table 1. Composition and properties of molten-salts Fuel salt MSRE Composition, mole ~
Liquidus °C °F Properties at °C °F Density (g/crn 3) Heat capacity (cal/(g°C)} or BTU/(lb°F) Viscosity, cP Vapor pressure, Tort Thermal conductivity W/(°C cm)
LiF BeF~ ZrF4 UF4
65 29. I 5 0.9 434 813 600 1112 2.27 0.47 9 Negligible 0.014
Coolant salt MSBR
LiF BeF~ ThF~ UF4
MSRE
71.7 LiF 16 BeF2 12 0"3 500 932 600 1112 3"35 0'33 12 Negligible 0-011
MSBR 66 34
455* 851
NaF NaBF4
8 92 385 725 454 850 1'94 0"36 2'5 200t 0.005
* Temperature inconveniently high for power reactor application, except in a three loop-system. t Pressure of BF3 in equilibrium with this melt composition at 607°C, the highest normal operating temperature of the coolant.
Molten-salt converter reactors possibly consisting of rectangular slabs in a square graphite box. The fuel is, of course, also the heat transport medium and must be piped to heat exchangers where its heat is transferred to an intermediate coolant salt which in turn flows to the steam generators. The metal components in contact with the fuel salt (reactor vessel, core supports, pipes, pumps and heat exchangers) would probably be fabricated from a modification of the Hastelloy-N (Ni-17Mo-7Cr5Fe) that was used in the MSRE. In the Molten-Salt Breeder Reactor, the fuel would be processed continuously for removal of fission products and ~aaPa. The solubility of xenon and krypton in the molten-salt is very low. A major fraction oftbese gases is continuously removed from the salt by sparging with helium and collected on activated charcoal absorption beds. Removal of most other fission products and of ~aaPa is accomplisbed in a closely coupled processing plant through which the fuel salt is passed on a 10-day cycle. The concentration of fissile material in the fuel is regulated by controlling the composition of the return stream from the processing plant to the reactor. Certain fission products, notably the metals Nb, Mo, To, Ru, Te and some others of lesser importance do not form stable fluorides and appear in elemental form. Tests indicate that they quickly disappear from the melt and deposit on surfaces of the primary system, chiefly the metal surfaces of the heat exchangers. A small, but not precisely determined, fraction of these metals deposits on the graphite.moderator, and for the MSBR we estimated that their adverse effect on the conversion ratio probably would not exceed 0-004, when the graphite is replaced every 4 yr. The build-up is not linear, and for a 30-yr graphite life the effect would probably not exceed 0.01-0.02, although this remains somewhat uncertain. In the converter reactors, the noble gases would still be removed by helium sparging and the noble metals would still leave the salt and deposit on primary-circuit surfaces. Most of the other fission products, however, would remain in the salt. At intervals of 6-10 yr, the uranium would be removed from the salt by the fluoride volatility process, as demonstrated in the MSRE, and the salt itself would then be discarded, i.e. treated and stored as waste. (There is a possibility of recovering some of the lithium and beryllium fluorides, e.g. by fractional distillation, but that possibility has not been fully explored and will not be considered further here.) in the MSRE, a mixture of helium and fluorine was bubbled through the salt which had been trans-
811
ferred to a fuel storage tank, oxidizing the UF4, to gaseous U F e. The effluent gas stream passed through a bed of sodium fluoride pellets at 750°F to remove volatile impurities and then into a series of cannisters filled with N a F pallets at 200-300°F, where the UFe was absorbed. In 47 hr offinorination over a six-day period, more than 9 9 . 9 ~ of the uranium, 218 kg, was removed from 4730 kg of salt. The U F e on the sodium fluoride traps was dccomaminated of gamma radiation by a factor of 4 × 109 and was sufficiently free of fission products to allow the traps to be handled without shielding. The UF6 would subsequently be recovered by heating the N a F beds and reducing the released UF~ to U F 4 with hydrogen. Addition of uranium to the carrier salt is easily accomplished by dissolving in the melt pellets of a UF4-LiF eutectic salt containing 61 ~ uranium by weight. During the initial fueling of the MSRE, most of the fuel was added to the carrier salt in the fuel drain tank, with the mixture then transferred to the reactor to test the approach to criticality. The final additions to achieve criticality were made directly to thecirculating fuel by means of a sampling access to the free liquid surface in the fuel pump bowl. Subsequent small additions to maintain reactivity throughout the cycle were easily accomplished in the same manner with the reactor at power. This procedure was demonstrated many times in the MSRE. We are accustomed to saying that on-line fueling of a molten salt reactor is as easy as taking an aspirin tablet, and since the mole fraction of uranium in the circulating salt is so small (---0.2-0.3~), the addition of tiny amounts of uranium concentrate has negligible effects on the volume and properties of the circulating salt. Thus, the long-term reactivity changes associated with fuel burn-up and build-up of fission products are accommodated by continuous adjustment of the fuel concentration. It is also worth noting that plutonium as well as uranium can be used as the initial fuel charge and as makeup feed for molten-salt reactors, and the majority of our converter reactor studies have in fact been based on the use of plutonium, as I shall discuss below. III. PERFORMANCE The key question with respect to the attainment of very high conversion ratios in thermal reactors operating on the thorium cycle is essentially the same as for breeding: The question is not whether these goals are physically possible but whether they are economically feasible. D o conversion ratios as high as 0.9-0.95 actually result in the lowest-cost
812
A. M. PERRY
power, either at present or at prospective future prices of uranium and of separative work? Our impression in the past has been that they do not, that systems capable in principle of providing such high conversion ratios would do so in general only at the expense of a substantial fuel cycle cost premium, possibly running to several mills per kwh. At this conference, it is our task in part to put the economics of high-conversion systems in clear perspective, and to see so far as possible how high the optimum conversion ratio may be driven, for each reactor type, by increases in uranium price or by potential changes in other important economic variables. Our studies of molten-salt converter reactors have not been sufficiently extensive to permit a comprehensive analysis of fuel-cycle costs as a function of the key variables. I can only present the data we do have, which provide a particular slice through the parameter-space. Tho molten-salt converter reactors we are discussing consist essentially of a graphite core with fuel salt passages occupying 10-15% of the core volume. The salt is a mixture of LiF, BeF2 and ThF4, with a very small amount (up to about 0.3 mole %) of UF4 or PuFa. The principal variables available for optimizing the performance of the reactor are the volume fraction of salt in the core, the concentration of thorium in the salt, and the duration of the fuel cycle. Additional degrees of freedom include the reflector thickness, the possibility of zoning the core (varying the salt volume fraction) to flatten the power distribution, the choice of feed material, and the possibility of changing the feed material during the cycle. With respect to the last point, it should be noted that plutonium is not removed from the salt by the fluoride volatility process. Thus, any plutonium left at the end of a cycle would be discarded along with the fission products and the carrier salt. While the amount of plutonium in the salt is not large, even for cycles in which it is used as initial fuel and makeup feed, the end-of-cycle loss can be reduced by switching to enriched uranium feed during the cycle, thus allowing most of the plutonium present to burn out. In studying the effect of the above variables individually (though not comprehensively over the whole parameter space), we have found it desirable to adopt a salt volume fraction of 0.12 and a fuel cycle duration of 6 equivalent full-power years (though significantly shorter and longer cycles appear to yield costs only slightly higher). The principal effect of varying the thorium concentration is to vary the competition between the fissile and fertile materials on the one hand and the fixed neutron
poisons (moderator, carrier salt and fission-products) on the other, so that increased conversion ratio is achieved (by increasing thorium concentration) at the expense of increased fissile and fertile inventory, while the energy generated per cycle is held fixed. This is made possible because the fissile fuel and its diluent can be mixed in widely varying proportions-a characteristic that is shared by the H T G R , but not by most other solid-fueled heterogeneous reactors. The relationship between conversion ratio and thorium concentration is shown in Fig. 1. It should be noted that uranium feed and plutonium feed (switching to uranium after 4 yr) give quite similar results. This is true largely because zzzU rather quickly becomes the dominant fuel isotope in either case. This may be seen from Fig. 2, which shows the principal nuclide inventories for the case with 8 mole ~ thorium as a function of time during the 4 cycles. It may also be noted in Fig. 2 that a reduced thorium concentration was used in the first cycle. This was done to avoid quite high inventories of plutonium associated with self-shielding of the resonances of ~Z°Pu and 24°Pu during the early years of the first cycle while bred 2aaU is building up. (The thorium concentrations, in mole per cent, that were used in the first and in subsequent cycles for each of six cases shown in Fig. 1 are indicated by the following pairs of numbers: 10/14, 8/12, 6/10, 6/8, 6[6 and 4/4.) Finally, it should be noted that the plutonium feed specified for these cases had a nominal composition chosen to represent plutonium discharged
.~
,00
~..C
0"95
~.5
0-90
/
"~
/"
/ 1o
3.0
.> "
T
~
!
0-80
L)
2'0
0-75
j~" .~ )/
/
// 1
0-70
6
/
,~
o~
f~N~ switch t'oZ35U o ~ ¢ 4 efpy C~Ol~fl~te h f e 24 efpy (35'core clio ) eU feed r.~ XPu feed no sw~tch ~o U 28 a~o zoned core AU feed, 27'd10 zoned core Fissile inventory oapmx,mote overa~ p for equriibrium cycle [ I'0 8 liO 1'2 14 16
Thorium
concentrotion-rnole %
Fig. 1. Molten-salt convener reactors. Conversion ratio and fissile inventory vs thorium concentration.
Molten-salt conve~er reactors ~o(, I-------~,-- u(,)---4
i2000,
L
"~0,000
I~ Th
b
I" __.~t~O,O00 k~Th
_l
}---- Pu(~)---+~-uie) --4
~I~ u(~l-'4
~ I I 0 , 0 0 0 kg Th (8 mole % ThF~)
(8 mOW~% "i'nF~)
~6oo
Po,)
813
,. I i -] [~
~,o,ooo wgTh ' (Bmote% "[~Fa) - - -
---i
"1
I
/
0
Z
i
2
4
i
i
6
8 Operot"ing
tO -hme
I~ 12 (eqmvolent
14 full-power
16
18 18
i- ......
20
22
2"
years)
Fig. 2. Principal nuclide inventories for MSCR with 8 mole ,% ThF,. (LWR plutonium and enriched uranium feed.) for all constituents of the fuel. The fuel-cycle costs (Bauman, 1972) in the lower solid curve of Fig. 3 were calculated by present-value accounting (using a discount factor of 7 % annum); this allows for the displacement in time of expenses and revenues, and includes both the initial core loading and the final fuel recovery at the end of reactor life. The top curve in Fig. 3 displays my own estimates from the same data, not present-valued, based on average inventories, and average feed rates. While the results of the two computations are somewhat different, the trends are similar, both showing some preference for the lower thorium concentrations I"a I b (4-8 mole ~°,o). The relationship between fuel cycle cost and conversion ratio for these data is shown in Fig. 4 and suggests an economic optimum conver- 0 ~:'3e sion ratio in the range, 0.8-0.85, although somewhat higher conversion ratios would appear to involve 0"~ e h .e fairly small cost penalties. The partial fuel cycle costs shown in Figs. 3 and 4 do not include the cost of the fluoride volatility process for recovering the uranium or the cost of treating and storing the wastes. On the basis of the i cost of the equipment used in the MSRE, we have previously (MacPherson, 1968) estimated the cost of SoVt ( my es~-imo~e) uranium recovery to add less than 0-1 mill/kWh to L_ - o . . . . • - - _ _ - o . . . . . . • . . . . . . • . . . . . . • the fuel cycle cost. While this may now be insufficient I to cover all costs of processing and waste disposal, t1 Includes ~plocemen'r ond inventory this cost component seems likely to remain rela~l i ; i tively minor. Furthermore, since the fluoride volatility equipment would be at the" reactor site Thorium concenl'rotion-mole % and used primarily during the infrequent processing Fig. 3. MSCR fuel cycle cost. Basis: 11"9 $/g ~35U, 13.8 $/g zs3U, inventory charge rate = 0-132/yr. Ex- campaigns, the processing cost in mills/kwh should be relatively insensitive to fuel cycle duration, wherecludes processing--see text.)
from light-water reactors after one cycle, i.e., 60% zaaPu, 24% ~4°Pu, 12% ~lpu and 4% ~2pu. Fuel cycle costs associated with the upper curve in Fig. 1 (one-zone core, Pu feed switched to ~ U after 4 yr) are shown in Fig. 3. These costs were based on unit costs of $11.90/g = s u , $13.80/g ~ U , $9.90/g fissile plutonium, $120/kg of ~Li (99.995 %), $16.50/kg BeF 2 ($86/kg of contained beryllium), and $14.30/kg ThF 4 ($19/kg of contained thorium). An inventory charge rate of 13.2 % per year was used
,
L
814
A.M.
understood and has been demonstrated. The longterm behavior of the fission-product metals, such as Nb, Mo, Tc, Ru, Te and a few others is not fully established. No difficulties associated with these metals as a group were experienced in the MSRE, but their behavior over longer times and in greater amounts remains to be investigated in a larger reactor. Post-test examination of metal components of the MSRE showed some intergranular attack, and this is discussed below under "structural metals."
I-4
I.C
o-e "
0.~
0-~
~o ~
a
n
a
i nventOrY
c-. . . . . . . .~. . . _
: 0-7
PERRY
0-75
_ .c-
'
I
0.8
0-85
Conversion
replaeernen t )
ratio
~.,iU0 I 0.9
1 0-95
~.0
(net)
Fig. 4. Fuel cycle cost vs conv=rsion ratio.
(Excludes
processing.) as the salt replacement cost would be almost inversely proportional to cycle length. Because the cost tradeoffs illustrated in Fig. 4 involve primarily the burnup and inventory of the fissile fuel, the cost optimization--in this particular slice through the parameter-space--is not sensitive to the value of the fuel. A major increase in the cost of 2asU, assuming a proportional increase in the value of other fissile isotopes, would raise the curve by a nearly constant factor, but would not appreciably shift the minimum. However, such a major increase in fuel value, without a proportional increase in the value of the carrier salt, would shorten the optimum cycle time, permitting a higher conversion ratio without increasing the fuel inventory. IV. THE STATUS OF MOLTEN-SALT REACTOR
TECHNOLOGY The technology necessary for the construction of molten-salt reactors has been developed over more than two decades, and major elements were demonstrated successfully by the operation of the MSRE. In the following paragraphs I will summarize very briefly the present status of the technology. A very comprehensive discussion may be found in Ref. 4, which in most respects is still current. 4.1. Fuel (Rosenthal et al., 1972; Grimes 1970) The fuel is stable and compatible with graphite and with Hastelloy-N. While the fission process is slightly oxidizing in the molten-salt fuel, the control of the oxidation-reduction potential in the salt is
4.2. Graphite (Rosenthal, 1972; Scott and Fatherly, 1970) Graphites are now available that would be satisfactory for use in the converter reactors discussed above. It will be desirable, especially for the breeder, to develop graphites capable of irradiations significantly greater than 3 x 10t~ nvt, and there is reason to hope that this will be possible (Rosenthal, 1972). The present limitation, however, is acceptable. In addition, for the breeder, it is desirable to hold x~Xe poisoning to the lowest possible level, and we believe that a poison fraction (neutron absorptions in z~Xe divided by absorptions in fissile fuel) of 0.005 can be achieved. To do this, however, will require graphites of very low permeability, e.g., I0-s cm2/sec (helium STP) or less, and at present it appears that this can only be accomplished by coating the graphite surface, with pyrolytic carbon. The procedures for applying the coatings and the durability and long-term integrity of the coatings are not yet fully established. However, uncoated graphites could be used in the converter reactors with a poison fraction probably not exceeding 0.01-0.02. (A detailed computation has not been made for the particular conditions qf the MSCR's discussed here.) 4.3. Structural metals (Rosenthal, 1972; McCoy, 1970) The Hastelloy°N used in the construction of the MSRE (Ni-17Mo-7Cr-5Fe), like a number of other alloys, has been found to be very sensitive at elevated temperatures, e.g., 600--700°C, to the presence of the helium bubbles which form along grain boundaries during irradiation, causing the alloy to lose most of its ductility. The helium is produced by (n, =) reactions, for example, by slow neutrons in boron and by fast and slow neutrons in nickel. Since very small amounts of helium are harmful, it has not seemed feasible to attack the problem by preventing the helium production (e.g. by removing boron from the alloy). However, a satisfactory solution to this problem has bee.n found in the
Molten-salt converter reactors addition of about 2 ~ titanium to the alloy. This promotes the formation of finely dispersed carbides which trap the helium and do not allow it to migrate to the grain boundaries. More recently, another problem has appeared in the form of intergranular cracking of Hastelloy-N in surveillance specimens and in several components of the MSRE (essentially all metal components in contact with the fuel salt). Both out-of-pile and inpile tests have now established that these cracks are caused by tellurium. Results of in-pile and out-ofpile tests are similar and have indicated that irradiation p e r se is not a factor. The depth of the cracks is very small and is believed to be proportional to the fourth root of the exposure time, a characteristic of grain boundary diffusion as compared with the square foot for bulk diffusion. It is estimated (Rosenthal et al., t972) that the penetration in the MSBR operating at 650°C for 30 yr would be about 4 mils. This should be tolerable. However, this estimate is only tentative, and a goal of the MSR project is to develop an alloy that does not exhibit this intergranular cracking. Several materials have in fact been found to be very resistant to this type of attack, and some of these might be suitable for use in an MSR (Rosenthal et al., 1972). These include 304 stainless steel and nickel--high chromium alloys, such as Inconel 600. A potential problem with 304 stainless steel is corrosive attack by the fuel salt and by the coolant salt. Corrosion by the fuel salt might be controlled by reducing the outlet fuel temperature in the reactor (from 704°C to about 650°C), but a transition to Hastelloy-N would probably have to be made in the intermediate heat exchanger. Inconel 600 is embrittled by neutron irradiation, and the solution of this problem by controlling composition, grain size and heat treatment (which has been found effective for Hastelloy-N and types 304 and 316 stainless steel) has not been demonstrated for the nickel-high chromium alloys. The presently preferred solution to the cracking problem is to continue with Hastelloy-N, with some additional small modifications of composition. In the material screening tests with tellurium, a modified Hastelloy-N alloy containing 2y/~ niobium was found to be entirely free from cracks. The addition of niobium or a small increase in the amount of titanium (presently at 2 Y/o to control radiation embrittlement) presently seem to be the best prospects for solution of this problem. There remains some uncertainty concerning the compatibility of Hastelloy-N with steam, and this will require further investigation.
815
4.4. Tritinm While tritium is produced to some degree in all reactors, the ZH production in an MSBR is unusually large owing primarily to the reactions 7Li(n, n, ~)ZH (threshold about 3"8 MeV) and eLi(n, ct)3H (all neutron energies). Total production in a 2250 MW(t)MSBR is calculated to be 2420 Ci/day, of which only 30 Ci/day come directly from fission. The tritium tends to escape rapidly from the salt, partly by diffusion through the large heat-exchange surfaces, and it has been estimated that about a third of the tritium produced would reach the steam svstern. Tritium in the steam system would be difficult to control; hence methods must be found to collect and store the tritium and to prevent it from reaching the steam system. Several methods offer promise of adequately controlling the distribution of tritium in the MSBR plant, and investigations are under way to determine their relative merit and effectiveness. These methods include adding hydrogen to the fuel salt, reducing the permeability of the metal walls, altering the noble-gas stripping system, exchanging tritium for hydrogen in hydrogenous compounds or reacting it with oxide in the coolant salt, and using other fluids to couple the primary system to the steam generators. Several possibilities are discussed in the following paragraphs. The sodium fluoroborate coolant salt used in the intermediate heat transfer loop is probably an effective getter for tritium. Evidence from the Coolant Salt Test Facility, which still contains some tritium in the salt after 2500 hr of operation, suggests that the tritium may be tied up in a compound and is not just dissoved in the metal walls of the lo0p. If increased resistance to tritium diffusion in the steam generators proves still to be needed, a possibility is to use duplex tubing having Incoloy 800 on the steam side and nickel on the salt side. Upon reaction with the steam, an oxide film is formed, on the surface of the Incoloy, which is expected to reduce the permeability of the tube wall for tritium by two orders of magnitude. Such duplex tubing has been manufactured by the International Nickel Company and has been tested in a corrosion test facility at TVA's Bull Run (supercritical) Steam Plant. A third possibility, which I judge to be economically much less attractive than the others, is to introduce a third heat transfer loop. The secondary loop might then contain the alternate coolant salt 23LiF, 41NaF, 36BeFe or the salt 66LiF, 34BeF z, while the tertiary loop might contain helium (containing small amounts of oxygen and water vapor) or Hitec heat transfer salt (a mixture of NaNO2, NaNOa and KNO3 in which the tritium could be
A. M. PrJ~Y
816
oxidized to T20, collected and disposed of). These possibilities have not been thoroughly evaluated. Other possible approaches have been considered and are discussed in (Rosenthal, et al, 1972) (pages 412-414). It may be that two or more methods in conjunction may be needed to reduce tritium escape to acceptable levels (e.g., less than 2 Ci/day). With a multiplicity of approaches available, we are confident that the release of tritium can be adequately controlled. Since tritium equivalent to about 8 1. of water would be produced by a 1000 MWe plant over its lifetime, the dilution of tritium by ordinary hydrogen must be limited in order to avoid storage of excessive volumes of tritiated water. A dilution factor of 104 would yield 80m a of tritiated water, which is probably manageable, but much larger dilutions could present significant problems.
Scott, D. and Eatherly, W. P. (1970) Graphite and xenon behavior and their influence on molten-salt reactor design. NucL Applic. Technol. 8, 179. Easy-reference curves f o r some quantities of interest in discussing advanced converter reactors (Submitted 21 July 1975) 4; 4C~
3-." F--" 3(
/ /°,
2.'.
z~ U
2<
..~ 239puNj
i'\ r ~ j,
/
/" //
24 Pu
/
v
I ,1
i!
4.5. Plant components While engineering components in the MSRE performed well, much larger components will be needed for full-scale power plants; and no steam generator, with heat supplied by a molten-salt, has been tested. Scale-up of components, development and testing of a steam generator, and further development of the off-gas system will be major development requirements of the program. Similarly, tools and techniques for maintenance or replacement of large activated or contaminated components will require substantial further development effort, notwithstanding the considerable progress already made in connection with the MSRE.
14: o.'
1
I0-z
t
!
i
[
I
io z
to 3
io"
io ~
to 6
T
I(~I
I0 0
id
Neutron
energy,
eV
Fig. 1A.
REFERENCES
Bauman, H. F. (1972) Fuel for molten-salt converter reactors. MSR Program Semiannu. ProM. Rep. August 31, 1972, ORNL-4832, pp. 16-25. Grimes, W. R. (1970) Molten-salt reactor chemistry. NucL Apfllic. Technol. 8, 137. Haubenreich, P. N. and Engel, J. R. (1970) Experience with the molten-salt reactor experiment. NucL Applic. Technol. 8, 118. MacPherson, H. G. (1968) Molten-salt reactors. Proceedinffs of the International Conference on Constructice Uses of Atomic El~rgy, Washington, D.C., November 15-18. McCoy, H. E. et al. (1970) New developments in materials for molten-salt reactors. Nucl. Applic. Technol. 8, 156. Perry, A. M. and Weinberg, A. M. (1972) Thermal breeder reactors. Ann. Rev. Nucl. Sci. 22, 317. Robertson, R. C. (ed) (1971) Conceptual Design Study of a Sinffle.Fluid Molten-Salt Breeder Reactor, USAEC Report ORNL-4541. Rosenthal, M. W. et al (1972) The Development Status of Molten-Salt Breeder Reactors, USAEC Report ORNL-4812.
-~
I
~_
/~3~
23%//
C~ ~-'"'~
u
'''~
(f~-i)
13
z33, f
5
7
~
~
900"K
5
..-2
T
0 9--
0
z3
7
o.~ i 0 z.
~
239pu
r
r ; [llrrl
2.
~
t tO3
2
N¢/N,,
Fig. IB.
BRm (900"K i r ~Tilr!
[ i 04
i07
Molten-salt converter reactors
817
I00,
9C
8C
02:Z L~: glss~ons ,' b r c m u m
C'E
O:E --
0.i2--
-._~----f
|
j X = I 0 0 MWd / kg
mlned :
atom
i -L ( ( l * o ) (I-Ca: L : G D D feed/Utah,urn mined L : z'b ctoms in pro(~uct/U atoms it, feed
70
= pcsf frss~ons factor ~ ] © !.e ~v_~ absorption , fission ratro ~-~ 12-J ~5 fo~ (t~erma ''ehOrlUm reocfors C~ = Comversrot" ra~la mc!uOln~ losses
60
~
}
5C
MWd / k g
X:50
MWO/kg J
L ~ 0 ~ ,c~epenolng on ore 0 9a: value used here
for 9~15 % 23~U = 0 0 0 6 5 7 fc, Ol % "rods =000512 for 0 2 % i*ods : 0 00412 for 0 Z % t0;Is
X-15 MWO/kg I
0 rOI
t1____.._..._.
30--
X=iO MWd/kg h
I
Toils Assoy=OI %
20--
ToJIs Assay: 0 2% Tods Assoy:O 3 %
a = 0"02} a = recycle
10--
"~ JJ'~
= =0-03~
O'04L 65
0 02~-
i 0"70
i 0"75
losses ,
(processing + f a b r i c a f i o n ) I 0`80
I 0"85
! 0-90
I
I'O0
0-95
Fig. IE. 0 70
i
•
0 75
0"80
0"B5
I
0-90
095
~OC
CR
Fig.
1C.
Uromum
:f co"
Nominal of gx l=O,
0-05
0.10
0.15
enriched
1"o 93 15 %
$ 3 / k g U for
,/
235U
/ // /'~ /
converslor~
U~OB ~?o UF6
• is included in cost of Z~U • iS -no't *ncluded m c o s t of U~Q B
0.2C
Fer*He/ fissile rol"~o= ,~0 g : Annual fractional growth !=SaecJfLc InvenToryI kg/MW(e) ( Eleci"r~co I ef f icletlc y= ~ I/~)
/
/
//
/
/
// /
"/
50
s ~5,sw2 L/;/~
aO
~ Ot9
~
I'-C
i
Breeaing
raf,o
Fig.
1D.
~.Jl
!
40/SWU
I-2 02°/°
r ' o $ 30/SWU o $ 26/SWU
------03% Note
J
I0
2C
30
toils assay tails assay
scale change ! 50
Fig. 1F.
l 70
I 90
[ iOD ~13
818
A . M . Pr.~'~"
%.
2
o
c c I-¢
I0
20
50
50
Fig. IG.
70
~0
100 11(3
MOLTEN-SALT C O N V E R T E R REACTORS ALFRED M. PERRY Institute for Energy Analysis, Oak Ridge Associated Universities, Oak Ridge, Tennessee 37830, U.S.A.
Abstract--Molten-salt reactors appear to have substantial promise as advanced converters. Conversion ratios of 0.85-0.9 should be attainable with favorable fuel cycle costs, with ~asU valued at $12/g. An increase in "6U value by a factor of two or three ($10--$30/1b. U3Os, $75/SWU) would be expected to increase the optimum conversion ratio, but this has not been analyzed in detail. The processing necessary to recover uranium from the fuel salt has been partially demonstrated in the MSRE. The equipment for doing this would be located at the reactor, and there would be no reliance on an established recycle industry. Processing costs are expected to be quite low, and fuel cycle optimization depends primarily on inventory and bumup or replacement costs for the fuel and for the carrier salt. Significant development problems remain to be resolved for molten-salt reactors, notably the control of tritium and the elimination of intergranular cracking of HasteUoy-N in contact with tellurium. However, these problems appear to be amenable to solution. I think that it is appropriate to consider separating the development schedule for molten-salt reactors from that for the processing technology required for breeding. The Molten-Salt Converter Reactor should be a useful reactor in its own right and would carry us far towards the achievement of true breeding in thermal reactors. I. INTRODUCTION Molten-salt reactor technology had its beginning in the Aircraft Nuclear Propulsion program at .Oak Ridge. In the latter fifties attention shifted to civilian power applications, with emphasis on breeder reactors. The motivation for this was the classical one for fluid-fuel reactors, that is, the achievement o f good neutron economy by fast and (it was hoped) inexpensive on-stream processing of the liquid fuel to keep fission-product poisoning small. Over the years relatively little attention has been given to converter reactors, since we believed that the molten-salt reactor concept was uniquely suited for thermal b r ~ d i n g (Perry and Weinberg, 1972), and that has remained the principal motivation for its development. It was recognized, however, that development of the complex processing technology did not necessarily have to proceed on the same schedule as that of the reactor; and some consideration was given to the performance of reactors designed for breeding but lacking a facility for continuous on-stream processing. In this case, the fuel would be batch-processed at intervals to be chosen on the basis of an economic evaluation. A satisfactory procedure for this batch processing was successfully demonstrated in conjunction with the MoltenSalt Reactor Experiment (Haubenreich and Engle, 1970), and will be described later. N o special effort has been made to optimize a molten-salt converter reactor as such, but the breeder-without-processing 809
designs have nonetheless proved to have the potential for quite interesting performance. It is these reactors that are discussed in this paper. II. GENERAL DESCRIPTION In a molten-salt reactor the fuel consists of U F 4 or PuFa dissolved in fluorides of lithium, beryllium, sodium, potassium, zirconium, thorium, or some combination of these (Grimes, 1970) (Rosenthal et al., 1972). Fluorides of the metals other than U, Pu, or Th are used as diluents and to keep the melting point low enough for practical use (MacPherson, 1968). As a class, these salts are stable, have low vapor pressures at the temperature necessary to drive a m o d e m steam cycle, and have adequate solubilities for thorium, uranium, and plutonium; they are chemically compatible with a satisfactory moderating material, graphite and with a number of commercial nickel-based alloys. Other anions and other cations have been considered (Grimes, 1970); Rosenthal et al., 1972), but most of the development effort has centered on those mentioned. In the A N P program a mixture of N a F , Z r F 4 and UF4 was used, but in the subsequent Molten-Salt Breeder Reactor program, for reasons of neutron economy, the preferred salts have been L i F and BeFo., with the lithium enriched to 99-995 ?/o ~ in the 7Li isotope. (At this level, in a typical MSBR configuration, the remaining SLi absorbs 40 ~.{,as many neutrons as the 7Li.) Composition of the salt specified for the
A. M. l~gg¥
810
MSBR is LiF (72 mole %), BeF z (16~), ThF 4 12 9/o). UF4 would be dissolved in this salt with a concentration of about 0.3 mole ~ ; at equilibrium the uranium would contain "-aaU, z~U, za*U and zasU approximately in the proportions 65[22[6[7. Some pertinent properties of this salt are shown in Table 1, along with those of the salt that was used as fuel in the MSRE (Molten-Salt Reactor Experiment) (Haubenreich and Engle, 1970). Also shown are properties of some coolant salts for use in intermediate heat transfer loops between the primary fuel circuit and the steam generators (salt-to-air heat exchanger in the MSRE). Graphite is a suitable moderator for molten-salt reactors. Its moderating power is about twice that of the salt, while its effective macroscopic absorption cross section is about one-fifth that of the salt, exclusive of the actinides. Fuel salt will typically /o to 15 9/o of the core volume, the balance occupy 10 °/ being graphite. No metals are used in the core proper. It has been established that the salt does not wet the graphite, that if the graphite pores are small enough the salt will not enter them, owing to surface tension, and that a satisfactory distribution of pore sizes is readily attainable in commercial graphites (Rosenthal et aL, 1972; Scott and Eatherly, 1970). Fission-product gases, on the other hand, notably laSXe, can diffuse into the graphite (Scott and Eatherly, 1970), with a non-negligible effect on the neutron balance; more of this later. Graphite is, of course, subject to radiation damage (Rosenthal et aL, 1972; Scott and Eatherly, 1970). In particular, the isotropic graphites found suitable for the MSBR undergo significant volume changes, at first shrinking and then expanding. For
fast neutron fluences greater than about 3 × 1022 n/cm z (E n > 50 keV) the rate of expansion becomes quite rapid, and it appears that this presently represents an approximate upper limit to the acceptable exposure of the graphite. An approximate inverse relationship thus exists between the useful life of the graphite and the core power density. L x Pm"~200 where L is the moderator life in full-power years, and P,n is the maximum core power density in W/ c m s.
In the reference M s B R design, (Haubenreich and Engle, 1970; Robertson, 1971) the maximum power density is about 70 W/cm s, and the useful graphite life would be about three years at full power, or four years at 0.75 plant factor. The design must, therefore, allow for fairly frequent replacement of the moderator. Much larger cores have also been studied, with the maximum power density limited to about 8 W/ cm z, so that a useful core life of about 24 full power years might be achieved (based on 30 yr at 0.8 plant factor). The studies of converter reactors with batch processing cycles have been based mainly on these larger cores (Bauman, 1972). These cores are visualized essentially as larger versions of the MSRE in which the moderator consisted of 2-inch square bars of graphite arranged vertically in a regular rectangular array. Fuel salt flowed in passages between the bars, and spacing was maintained by machined ridges on the outer surfaces of the bars. In a larger reactor, it is likely that larger graphite assemblies would be used,
Table 1. Composition and properties of molten-salts Fuel salt MSRE Composition, mole ~
Liquidus °C °F Properties at °C °F Density (g/crn 3) Heat capacity (cal/(g°C)} or BTU/(lb°F) Viscosity, cP Vapor pressure, Tort Thermal conductivity W/(°C cm)
LiF BeF~ ZrF4 UF4
65 29. I 5 0.9 434 813 600 1112 2.27 0.47 9 Negligible 0.014
Coolant salt MSBR
LiF BeF~ ThF~ UF4
MSRE
71.7 LiF 16 BeF2 12 0"3 500 932 600 1112 3"35 0'33 12 Negligible 0-011
MSBR 66 34
455* 851
NaF NaBF4
8 92 385 725 454 850 1'94 0"36 2'5 200t 0.005
* Temperature inconveniently high for power reactor application, except in a three loop-system. t Pressure of BF3 in equilibrium with this melt composition at 607°C, the highest normal operating temperature of the coolant.
Molten-salt converter reactors possibly consisting of rectangular slabs in a square graphite box. The fuel is, of course, also the heat transport medium and must be piped to heat exchangers where its heat is transferred to an intermediate coolant salt which in turn flows to the steam generators. The metal components in contact with the fuel salt (reactor vessel, core supports, pipes, pumps and heat exchangers) would probably be fabricated from a modification of the Hastelloy-N (Ni-17Mo-7Cr5Fe) that was used in the MSRE. In the Molten-Salt Breeder Reactor, the fuel would be processed continuously for removal of fission products and ~aaPa. The solubility of xenon and krypton in the molten-salt is very low. A major fraction oftbese gases is continuously removed from the salt by sparging with helium and collected on activated charcoal absorption beds. Removal of most other fission products and of ~aaPa is accomplisbed in a closely coupled processing plant through which the fuel salt is passed on a 10-day cycle. The concentration of fissile material in the fuel is regulated by controlling the composition of the return stream from the processing plant to the reactor. Certain fission products, notably the metals Nb, Mo, To, Ru, Te and some others of lesser importance do not form stable fluorides and appear in elemental form. Tests indicate that they quickly disappear from the melt and deposit on surfaces of the primary system, chiefly the metal surfaces of the heat exchangers. A small, but not precisely determined, fraction of these metals deposits on the graphite.moderator, and for the MSBR we estimated that their adverse effect on the conversion ratio probably would not exceed 0-004, when the graphite is replaced every 4 yr. The build-up is not linear, and for a 30-yr graphite life the effect would probably not exceed 0.01-0.02, although this remains somewhat uncertain. In the converter reactors, the noble gases would still be removed by helium sparging and the noble metals would still leave the salt and deposit on primary-circuit surfaces. Most of the other fission products, however, would remain in the salt. At intervals of 6-10 yr, the uranium would be removed from the salt by the fluoride volatility process, as demonstrated in the MSRE, and the salt itself would then be discarded, i.e. treated and stored as waste. (There is a possibility of recovering some of the lithium and beryllium fluorides, e.g. by fractional distillation, but that possibility has not been fully explored and will not be considered further here.) in the MSRE, a mixture of helium and fluorine was bubbled through the salt which had been trans-
811
ferred to a fuel storage tank, oxidizing the UF4, to gaseous U F e. The effluent gas stream passed through a bed of sodium fluoride pellets at 750°F to remove volatile impurities and then into a series of cannisters filled with N a F pallets at 200-300°F, where the UFe was absorbed. In 47 hr offinorination over a six-day period, more than 9 9 . 9 ~ of the uranium, 218 kg, was removed from 4730 kg of salt. The U F e on the sodium fluoride traps was dccomaminated of gamma radiation by a factor of 4 × 109 and was sufficiently free of fission products to allow the traps to be handled without shielding. The UF6 would subsequently be recovered by heating the N a F beds and reducing the released UF~ to U F 4 with hydrogen. Addition of uranium to the carrier salt is easily accomplished by dissolving in the melt pellets of a UF4-LiF eutectic salt containing 61 ~ uranium by weight. During the initial fueling of the MSRE, most of the fuel was added to the carrier salt in the fuel drain tank, with the mixture then transferred to the reactor to test the approach to criticality. The final additions to achieve criticality were made directly to thecirculating fuel by means of a sampling access to the free liquid surface in the fuel pump bowl. Subsequent small additions to maintain reactivity throughout the cycle were easily accomplished in the same manner with the reactor at power. This procedure was demonstrated many times in the MSRE. We are accustomed to saying that on-line fueling of a molten salt reactor is as easy as taking an aspirin tablet, and since the mole fraction of uranium in the circulating salt is so small (---0.2-0.3~), the addition of tiny amounts of uranium concentrate has negligible effects on the volume and properties of the circulating salt. Thus, the long-term reactivity changes associated with fuel burn-up and build-up of fission products are accommodated by continuous adjustment of the fuel concentration. It is also worth noting that plutonium as well as uranium can be used as the initial fuel charge and as makeup feed for molten-salt reactors, and the majority of our converter reactor studies have in fact been based on the use of plutonium, as I shall discuss below. III. PERFORMANCE The key question with respect to the attainment of very high conversion ratios in thermal reactors operating on the thorium cycle is essentially the same as for breeding: The question is not whether these goals are physically possible but whether they are economically feasible. D o conversion ratios as high as 0.9-0.95 actually result in the lowest-cost
812
A. M. PERRY
power, either at present or at prospective future prices of uranium and of separative work? Our impression in the past has been that they do not, that systems capable in principle of providing such high conversion ratios would do so in general only at the expense of a substantial fuel cycle cost premium, possibly running to several mills per kwh. At this conference, it is our task in part to put the economics of high-conversion systems in clear perspective, and to see so far as possible how high the optimum conversion ratio may be driven, for each reactor type, by increases in uranium price or by potential changes in other important economic variables. Our studies of molten-salt converter reactors have not been sufficiently extensive to permit a comprehensive analysis of fuel-cycle costs as a function of the key variables. I can only present the data we do have, which provide a particular slice through the parameter-space. Tho molten-salt converter reactors we are discussing consist essentially of a graphite core with fuel salt passages occupying 10-15% of the core volume. The salt is a mixture of LiF, BeF2 and ThF4, with a very small amount (up to about 0.3 mole %) of UF4 or PuFa. The principal variables available for optimizing the performance of the reactor are the volume fraction of salt in the core, the concentration of thorium in the salt, and the duration of the fuel cycle. Additional degrees of freedom include the reflector thickness, the possibility of zoning the core (varying the salt volume fraction) to flatten the power distribution, the choice of feed material, and the possibility of changing the feed material during the cycle. With respect to the last point, it should be noted that plutonium is not removed from the salt by the fluoride volatility process. Thus, any plutonium left at the end of a cycle would be discarded along with the fission products and the carrier salt. While the amount of plutonium in the salt is not large, even for cycles in which it is used as initial fuel and makeup feed, the end-of-cycle loss can be reduced by switching to enriched uranium feed during the cycle, thus allowing most of the plutonium present to burn out. In studying the effect of the above variables individually (though not comprehensively over the whole parameter space), we have found it desirable to adopt a salt volume fraction of 0.12 and a fuel cycle duration of 6 equivalent full-power years (though significantly shorter and longer cycles appear to yield costs only slightly higher). The principal effect of varying the thorium concentration is to vary the competition between the fissile and fertile materials on the one hand and the fixed neutron
poisons (moderator, carrier salt and fission-products) on the other, so that increased conversion ratio is achieved (by increasing thorium concentration) at the expense of increased fissile and fertile inventory, while the energy generated per cycle is held fixed. This is made possible because the fissile fuel and its diluent can be mixed in widely varying proportions-a characteristic that is shared by the H T G R , but not by most other solid-fueled heterogeneous reactors. The relationship between conversion ratio and thorium concentration is shown in Fig. 1. It should be noted that uranium feed and plutonium feed (switching to uranium after 4 yr) give quite similar results. This is true largely because zzzU rather quickly becomes the dominant fuel isotope in either case. This may be seen from Fig. 2, which shows the principal nuclide inventories for the case with 8 mole ~ thorium as a function of time during the 4 cycles. It may also be noted in Fig. 2 that a reduced thorium concentration was used in the first cycle. This was done to avoid quite high inventories of plutonium associated with self-shielding of the resonances of ~Z°Pu and 24°Pu during the early years of the first cycle while bred 2aaU is building up. (The thorium concentrations, in mole per cent, that were used in the first and in subsequent cycles for each of six cases shown in Fig. 1 are indicated by the following pairs of numbers: 10/14, 8/12, 6/10, 6/8, 6[6 and 4/4.) Finally, it should be noted that the plutonium feed specified for these cases had a nominal composition chosen to represent plutonium discharged
.~
,00
~..C
0"95
~.5
0-90
/
"~
/"
/ 1o
3.0
.> "
T
~
!
0-80
L)
2'0
0-75
j~" .~ )/
/
// 1
0-70
6
/
,~
o~
f~N~ switch t'oZ35U o ~ ¢ 4 efpy C~Ol~fl~te h f e 24 efpy (35'core clio ) eU feed r.~ XPu feed no sw~tch ~o U 28 a~o zoned core AU feed, 27'd10 zoned core Fissile inventory oapmx,mote overa~ p for equriibrium cycle [ I'0 8 liO 1'2 14 16
Thorium
concentrotion-rnole %
Fig. 1. Molten-salt convener reactors. Conversion ratio and fissile inventory vs thorium concentration.
Molten-salt conve~er reactors ~o(, I-------~,-- u(,)---4
i2000,
L
"~0,000
I~ Th
b
I" __.~t~O,O00 k~Th
_l
}---- Pu(~)---+~-uie) --4
~I~ u(~l-'4
~ I I 0 , 0 0 0 kg Th (8 mole % ThF~)
(8 mOW~% "i'nF~)
~6oo
Po,)
813
,. I i -] [~
~,o,ooo wgTh ' (Bmote% "[~Fa) - - -
---i
"1
I
/
0
Z
i
2
4
i
i
6
8 Operot"ing
tO -hme
I~ 12 (eqmvolent
14 full-power
16
18 18
i- ......
20
22
2"
years)
Fig. 2. Principal nuclide inventories for MSCR with 8 mole ,% ThF,. (LWR plutonium and enriched uranium feed.) for all constituents of the fuel. The fuel-cycle costs (Bauman, 1972) in the lower solid curve of Fig. 3 were calculated by present-value accounting (using a discount factor of 7 % annum); this allows for the displacement in time of expenses and revenues, and includes both the initial core loading and the final fuel recovery at the end of reactor life. The top curve in Fig. 3 displays my own estimates from the same data, not present-valued, based on average inventories, and average feed rates. While the results of the two computations are somewhat different, the trends are similar, both showing some preference for the lower thorium concentrations I"a I b (4-8 mole ~°,o). The relationship between fuel cycle cost and conversion ratio for these data is shown in Fig. 4 and suggests an economic optimum conver- 0 ~:'3e sion ratio in the range, 0.8-0.85, although somewhat higher conversion ratios would appear to involve 0"~ e h .e fairly small cost penalties. The partial fuel cycle costs shown in Figs. 3 and 4 do not include the cost of the fluoride volatility process for recovering the uranium or the cost of treating and storing the wastes. On the basis of the i cost of the equipment used in the MSRE, we have previously (MacPherson, 1968) estimated the cost of SoVt ( my es~-imo~e) uranium recovery to add less than 0-1 mill/kWh to L_ - o . . . . • - - _ _ - o . . . . . . • . . . . . . • . . . . . . • the fuel cycle cost. While this may now be insufficient I to cover all costs of processing and waste disposal, t1 Includes ~plocemen'r ond inventory this cost component seems likely to remain rela~l i ; i tively minor. Furthermore, since the fluoride volatility equipment would be at the" reactor site Thorium concenl'rotion-mole % and used primarily during the infrequent processing Fig. 3. MSCR fuel cycle cost. Basis: 11"9 $/g ~35U, 13.8 $/g zs3U, inventory charge rate = 0-132/yr. Ex- campaigns, the processing cost in mills/kwh should be relatively insensitive to fuel cycle duration, wherecludes processing--see text.)
from light-water reactors after one cycle, i.e., 60% zaaPu, 24% ~4°Pu, 12% ~lpu and 4% ~2pu. Fuel cycle costs associated with the upper curve in Fig. 1 (one-zone core, Pu feed switched to ~ U after 4 yr) are shown in Fig. 3. These costs were based on unit costs of $11.90/g = s u , $13.80/g ~ U , $9.90/g fissile plutonium, $120/kg of ~Li (99.995 %), $16.50/kg BeF 2 ($86/kg of contained beryllium), and $14.30/kg ThF 4 ($19/kg of contained thorium). An inventory charge rate of 13.2 % per year was used
,
L
814
A.M.
understood and has been demonstrated. The longterm behavior of the fission-product metals, such as Nb, Mo, Tc, Ru, Te and a few others is not fully established. No difficulties associated with these metals as a group were experienced in the MSRE, but their behavior over longer times and in greater amounts remains to be investigated in a larger reactor. Post-test examination of metal components of the MSRE showed some intergranular attack, and this is discussed below under "structural metals."
I-4
I.C
o-e "
0.~
0-~
~o ~
a
n
a
i nventOrY
c-. . . . . . . .~. . . _
: 0-7
PERRY
0-75
_ .c-
'
I
0.8
0-85
Conversion
replaeernen t )
ratio
~.,iU0 I 0.9
1 0-95
~.0
(net)
Fig. 4. Fuel cycle cost vs conv=rsion ratio.
(Excludes
processing.) as the salt replacement cost would be almost inversely proportional to cycle length. Because the cost tradeoffs illustrated in Fig. 4 involve primarily the burnup and inventory of the fissile fuel, the cost optimization--in this particular slice through the parameter-space--is not sensitive to the value of the fuel. A major increase in the cost of 2asU, assuming a proportional increase in the value of other fissile isotopes, would raise the curve by a nearly constant factor, but would not appreciably shift the minimum. However, such a major increase in fuel value, without a proportional increase in the value of the carrier salt, would shorten the optimum cycle time, permitting a higher conversion ratio without increasing the fuel inventory. IV. THE STATUS OF MOLTEN-SALT REACTOR
TECHNOLOGY The technology necessary for the construction of molten-salt reactors has been developed over more than two decades, and major elements were demonstrated successfully by the operation of the MSRE. In the following paragraphs I will summarize very briefly the present status of the technology. A very comprehensive discussion may be found in Ref. 4, which in most respects is still current. 4.1. Fuel (Rosenthal et al., 1972; Grimes 1970) The fuel is stable and compatible with graphite and with Hastelloy-N. While the fission process is slightly oxidizing in the molten-salt fuel, the control of the oxidation-reduction potential in the salt is
4.2. Graphite (Rosenthal, 1972; Scott and Fatherly, 1970) Graphites are now available that would be satisfactory for use in the converter reactors discussed above. It will be desirable, especially for the breeder, to develop graphites capable of irradiations significantly greater than 3 x 10t~ nvt, and there is reason to hope that this will be possible (Rosenthal, 1972). The present limitation, however, is acceptable. In addition, for the breeder, it is desirable to hold x~Xe poisoning to the lowest possible level, and we believe that a poison fraction (neutron absorptions in z~Xe divided by absorptions in fissile fuel) of 0.005 can be achieved. To do this, however, will require graphites of very low permeability, e.g., I0-s cm2/sec (helium STP) or less, and at present it appears that this can only be accomplished by coating the graphite surface, with pyrolytic carbon. The procedures for applying the coatings and the durability and long-term integrity of the coatings are not yet fully established. However, uncoated graphites could be used in the converter reactors with a poison fraction probably not exceeding 0.01-0.02. (A detailed computation has not been made for the particular conditions qf the MSCR's discussed here.) 4.3. Structural metals (Rosenthal, 1972; McCoy, 1970) The Hastelloy°N used in the construction of the MSRE (Ni-17Mo-7Cr-5Fe), like a number of other alloys, has been found to be very sensitive at elevated temperatures, e.g., 600--700°C, to the presence of the helium bubbles which form along grain boundaries during irradiation, causing the alloy to lose most of its ductility. The helium is produced by (n, =) reactions, for example, by slow neutrons in boron and by fast and slow neutrons in nickel. Since very small amounts of helium are harmful, it has not seemed feasible to attack the problem by preventing the helium production (e.g. by removing boron from the alloy). However, a satisfactory solution to this problem has bee.n found in the
Molten-salt converter reactors addition of about 2 ~ titanium to the alloy. This promotes the formation of finely dispersed carbides which trap the helium and do not allow it to migrate to the grain boundaries. More recently, another problem has appeared in the form of intergranular cracking of Hastelloy-N in surveillance specimens and in several components of the MSRE (essentially all metal components in contact with the fuel salt). Both out-of-pile and inpile tests have now established that these cracks are caused by tellurium. Results of in-pile and out-ofpile tests are similar and have indicated that irradiation p e r se is not a factor. The depth of the cracks is very small and is believed to be proportional to the fourth root of the exposure time, a characteristic of grain boundary diffusion as compared with the square foot for bulk diffusion. It is estimated (Rosenthal et al., t972) that the penetration in the MSBR operating at 650°C for 30 yr would be about 4 mils. This should be tolerable. However, this estimate is only tentative, and a goal of the MSR project is to develop an alloy that does not exhibit this intergranular cracking. Several materials have in fact been found to be very resistant to this type of attack, and some of these might be suitable for use in an MSR (Rosenthal et al., 1972). These include 304 stainless steel and nickel--high chromium alloys, such as Inconel 600. A potential problem with 304 stainless steel is corrosive attack by the fuel salt and by the coolant salt. Corrosion by the fuel salt might be controlled by reducing the outlet fuel temperature in the reactor (from 704°C to about 650°C), but a transition to Hastelloy-N would probably have to be made in the intermediate heat exchanger. Inconel 600 is embrittled by neutron irradiation, and the solution of this problem by controlling composition, grain size and heat treatment (which has been found effective for Hastelloy-N and types 304 and 316 stainless steel) has not been demonstrated for the nickel-high chromium alloys. The presently preferred solution to the cracking problem is to continue with Hastelloy-N, with some additional small modifications of composition. In the material screening tests with tellurium, a modified Hastelloy-N alloy containing 2y/~ niobium was found to be entirely free from cracks. The addition of niobium or a small increase in the amount of titanium (presently at 2 Y/o to control radiation embrittlement) presently seem to be the best prospects for solution of this problem. There remains some uncertainty concerning the compatibility of Hastelloy-N with steam, and this will require further investigation.
815
4.4. Tritinm While tritium is produced to some degree in all reactors, the ZH production in an MSBR is unusually large owing primarily to the reactions 7Li(n, n, ~)ZH (threshold about 3"8 MeV) and eLi(n, ct)3H (all neutron energies). Total production in a 2250 MW(t)MSBR is calculated to be 2420 Ci/day, of which only 30 Ci/day come directly from fission. The tritium tends to escape rapidly from the salt, partly by diffusion through the large heat-exchange surfaces, and it has been estimated that about a third of the tritium produced would reach the steam svstern. Tritium in the steam system would be difficult to control; hence methods must be found to collect and store the tritium and to prevent it from reaching the steam system. Several methods offer promise of adequately controlling the distribution of tritium in the MSBR plant, and investigations are under way to determine their relative merit and effectiveness. These methods include adding hydrogen to the fuel salt, reducing the permeability of the metal walls, altering the noble-gas stripping system, exchanging tritium for hydrogen in hydrogenous compounds or reacting it with oxide in the coolant salt, and using other fluids to couple the primary system to the steam generators. Several possibilities are discussed in the following paragraphs. The sodium fluoroborate coolant salt used in the intermediate heat transfer loop is probably an effective getter for tritium. Evidence from the Coolant Salt Test Facility, which still contains some tritium in the salt after 2500 hr of operation, suggests that the tritium may be tied up in a compound and is not just dissoved in the metal walls of the lo0p. If increased resistance to tritium diffusion in the steam generators proves still to be needed, a possibility is to use duplex tubing having Incoloy 800 on the steam side and nickel on the salt side. Upon reaction with the steam, an oxide film is formed, on the surface of the Incoloy, which is expected to reduce the permeability of the tube wall for tritium by two orders of magnitude. Such duplex tubing has been manufactured by the International Nickel Company and has been tested in a corrosion test facility at TVA's Bull Run (supercritical) Steam Plant. A third possibility, which I judge to be economically much less attractive than the others, is to introduce a third heat transfer loop. The secondary loop might then contain the alternate coolant salt 23LiF, 41NaF, 36BeFe or the salt 66LiF, 34BeF z, while the tertiary loop might contain helium (containing small amounts of oxygen and water vapor) or Hitec heat transfer salt (a mixture of NaNO2, NaNOa and KNO3 in which the tritium could be
A. M. PrJ~Y
816
oxidized to T20, collected and disposed of). These possibilities have not been thoroughly evaluated. Other possible approaches have been considered and are discussed in (Rosenthal, et al, 1972) (pages 412-414). It may be that two or more methods in conjunction may be needed to reduce tritium escape to acceptable levels (e.g., less than 2 Ci/day). With a multiplicity of approaches available, we are confident that the release of tritium can be adequately controlled. Since tritium equivalent to about 8 1. of water would be produced by a 1000 MWe plant over its lifetime, the dilution of tritium by ordinary hydrogen must be limited in order to avoid storage of excessive volumes of tritiated water. A dilution factor of 104 would yield 80m a of tritiated water, which is probably manageable, but much larger dilutions could present significant problems.
Scott, D. and Eatherly, W. P. (1970) Graphite and xenon behavior and their influence on molten-salt reactor design. NucL Applic. Technol. 8, 179. Easy-reference curves f o r some quantities of interest in discussing advanced converter reactors (Submitted 21 July 1975) 4; 4C~
3-." F--" 3(
/ /°,
2.'.
z~ U
2<
..~ 239puNj
i'\ r ~ j,
/
/" //
24 Pu
/
v
I ,1
i!
4.5. Plant components While engineering components in the MSRE performed well, much larger components will be needed for full-scale power plants; and no steam generator, with heat supplied by a molten-salt, has been tested. Scale-up of components, development and testing of a steam generator, and further development of the off-gas system will be major development requirements of the program. Similarly, tools and techniques for maintenance or replacement of large activated or contaminated components will require substantial further development effort, notwithstanding the considerable progress already made in connection with the MSRE.
14: o.'
1
I0-z
t
!
i
[
I
io z
to 3
io"
io ~
to 6
T
I(~I
I0 0
id
Neutron
energy,
eV
Fig. 1A.
REFERENCES
Bauman, H. F. (1972) Fuel for molten-salt converter reactors. MSR Program Semiannu. ProM. Rep. August 31, 1972, ORNL-4832, pp. 16-25. Grimes, W. R. (1970) Molten-salt reactor chemistry. NucL Apfllic. Technol. 8, 137. Haubenreich, P. N. and Engel, J. R. (1970) Experience with the molten-salt reactor experiment. NucL Applic. Technol. 8, 118. MacPherson, H. G. (1968) Molten-salt reactors. Proceedinffs of the International Conference on Constructice Uses of Atomic El~rgy, Washington, D.C., November 15-18. McCoy, H. E. et al. (1970) New developments in materials for molten-salt reactors. Nucl. Applic. Technol. 8, 156. Perry, A. M. and Weinberg, A. M. (1972) Thermal breeder reactors. Ann. Rev. Nucl. Sci. 22, 317. Robertson, R. C. (ed) (1971) Conceptual Design Study of a Sinffle.Fluid Molten-Salt Breeder Reactor, USAEC Report ORNL-4541. Rosenthal, M. W. et al (1972) The Development Status of Molten-Salt Breeder Reactors, USAEC Report ORNL-4812.
-~
I
~_
/~3~
23%//
C~ ~-'"'~
u
'''~
(f~-i)
13
z33, f
5
7
~
~
900"K
5
..-2
T
0 9--
0
z3
7
o.~ i 0 z.
~
239pu
r
r ; [llrrl
2.
~
t tO3
2
N¢/N,,
Fig. IB.
BRm (900"K i r ~Tilr!
[ i 04
i07
Molten-salt converter reactors
817
I00,
9C
8C
02:Z L~: glss~ons ,' b r c m u m
C'E
O:E --
0.i2--
-._~----f
|
j X = I 0 0 MWd / kg
mlned :
atom
i -L ( ( l * o ) (I-Ca: L : G D D feed/Utah,urn mined L : z'b ctoms in pro(~uct/U atoms it, feed
70
= pcsf frss~ons factor ~ ] © !.e ~v_~ absorption , fission ratro ~-~ 12-J ~5 fo~ (t~erma ''ehOrlUm reocfors C~ = Comversrot" ra~la mc!uOln~ losses
60
~
}
5C
MWd / k g
X:50
MWO/kg J
L ~ 0 ~ ,c~epenolng on ore 0 9a: value used here
for 9~15 % 23~U = 0 0 0 6 5 7 fc, Ol % "rods =000512 for 0 2 % i*ods : 0 00412 for 0 Z % t0;Is
X-15 MWO/kg I
0 rOI
t1____.._..._.
30--
X=iO MWd/kg h
I
Toils Assoy=OI %
20--
ToJIs Assay: 0 2% Tods Assoy:O 3 %
a = 0"02} a = recycle
10--
"~ JJ'~
= =0-03~
O'04L 65
0 02~-
i 0"70
i 0"75
losses ,
(processing + f a b r i c a f i o n ) I 0`80
I 0"85
! 0-90
I
I'O0
0-95
Fig. IE. 0 70
i
•
0 75
0"80
0"B5
I
0-90
095
~OC
CR
Fig.
1C.
Uromum
:f co"
Nominal of gx l=O,
0-05
0.10
0.15
enriched
1"o 93 15 %
$ 3 / k g U for
,/
235U
/ // /'~ /
converslor~
U~OB ~?o UF6
• is included in cost of Z~U • iS -no't *ncluded m c o s t of U~Q B
0.2C
Fer*He/ fissile rol"~o= ,~0 g : Annual fractional growth !=SaecJfLc InvenToryI kg/MW(e) ( Eleci"r~co I ef f icletlc y= ~ I/~)
/
/
//
/
/
// /
"/
50
s ~5,sw2 L/;/~
aO
~ Ot9
~
I'-C
i
Breeaing
raf,o
Fig.
1D.
~.Jl
!
40/SWU
I-2 02°/°
r ' o $ 30/SWU o $ 26/SWU
------03% Note
J
I0
2C
30
toils assay tails assay
scale change ! 50
Fig. 1F.
l 70
I 90
[ iOD ~13
818
A . M . Pr.~'~"
%.
2
o
c c I-¢
I0
20
50
50
Fig. IG.
70
~0
100 11(3