I
M. H. LIETZKE and R. W. STOUGHTON Oak Ridge National Laboratory, Oak Ridge, Tenn.
Feasible Chemical Forms for Thorium Breeder Blanket
Aluminum-iacketed metal slugs require elaborate processing and refabrication. Of the other blanket forms, thorium oxide and nitrate in deuterium oxide appear in long range most promising
IN
ORDER to utilize any appreciable fraction of the world supply of thorium, nuclear breeding must be feasible. There are only two known possibilities for such breeding which involves production of more fissionable material in a reactor than is consumed as fuelthe systems, uranium-plutonium-239 and thorium-uranium-233, In the former, fission neutrons from plutonium destruction are absorbed in uranium-238, giving plutonium-239 after two successive beta decays. In the latter, uranium-233 is destroyed and regenerated by a similar process, starting with neutron absorption by thorium-232. For breeding to be feasible the number of neutrons absorbed by all materials in a reactor must be less (by the amount of leakage) than those reproduced by the fissioning fuel; at least as many new fuel atoms must be made as are destroyed. These factors combined with the need for chemical stability under radiation greatly restrict the chemical forms which might be considered for the fuel or for the fertile material of a breeder. Also, chemical processing, and if necessary, refabrication methods must be neither excessively expensive nor involve excessive losses of fissionable material. Since processing methods will vary with the form of fuel or fertile material, the chemical form and processing method should be considered together in evaluating any particular system.
Nuclear Restrictions In principle, breeder reactors may be fast, thermal, or intermediate. That is, the average neutron energy may be within a factor of about 10 from the
202
average energy of fission neutrons (1.5 m.e.v.) ; it may correspond approximately to the thermal ( k T ) energy of the reactor operating temperature (probably in the range of hundredths of electron volts), or it may lie in some intermediate range. A thermal reactor requires a relatively high ratio of moderator to fuel atoms to degrade the energy of fission neutrons. A fast reactor, on the other hand, needs a sufficiently high ratio of coolant to fuel atom for heat removal without excessive moderationi.e., degradation of neutron energy. The neutron cross sections of uranium233 as a function of neutron energy have not been studied sufficiently to evaluate properly the feasibility of intermediate breeders. A convenient property of any nuclear fuel is 7, the number of neutrons reproduced by fission per neutron absorbed by fission plus capture. For uranium-233, the value of 7 for thermal neutrons (0.025 e.v.) is 2.31 =t0.03 (7) and since the ratio of capture to fission becomes low a t high neutron energies, that for fast neutrons is about 2.5 (7). The value is not known for intermediate regions. With the restriction of thermalized neutrons, there still remain questions of homogeneous us. heterogeneous reactors, one region us. two regions, and temperature of operation. A homogeneous reactor is one in which the fuel and moderator are intimately mixed, and usually this fuel-moderator mixture is pictured in fluid form circulating through the reactor and through a heat exchanger, In a heterogeneous reactor the fuel and moderator, or fuel, moderator, and coolant are geometrically separated. Similar definitions could hold for the blanket.
INDUSTRIAL A N D ENGINEERING CHEMISTRY
The blanket is defined as thorium plus all the supporting materials which are required for its mechanical and chemical stability such as the anion if it is a compound, structural materials holding it, moderator if any, and coolant. In a homogeneous breeder reactor, the fuel and fertile material may be intimately mixed in one region, or geometrically separated into two regions. I n the latter case the fuel would probably be in the inner region or core and the blanket in the outer region. In either a one- or two-region reactor, the blanket region has to contain a moderator, and perhaps be backed up by a reflector, for neutron leakage to be minimized with a minimum of thorium. I n either a one- or two-region reactor, the effective neutron losses may be summarized as: those in the core or in the fuel medium and blanket (blanket medium) lost through absorption by anion, moderator, and coolant atoms, and by other heavy isotopes; and neutron equivalent of chemical-processing fuel losses plus absorption by fission and corrosion products. In the core or fuel medium, some delayed neutrons may be lost if the fuel flows through the core, as in a homogeneous reactor. Also, there are leakage losses in the blanket; and for a heterogeneous or two-region homogeneous reactor, there are losses by core tank (boundary material) absorption. T o get an idea of the magnitude of allowable macroscopic neutron-absorption cross sections for materials other than thorium-232 and uranium-233, all these losses must be considered together and their combined equivalent losses keptless than 0.31---i.e.. ( 7 - 2) per
fuel atom destroyed in the core fuel compartment by neutron absorption. Since losses may vary greatly from one reactor design to another, it is not possible to give a single figure for allowable parasitic capture in the blanket. Nevertheless, by assigning reasonable ranges of values for all other losses, the range of probable allowable losses to moderator, anion, and other atoms in the blanket can be given. Then from known crosssection values, the allowable atomic ratio of any other element to thorium can be given. Of course, from an over-all economic point of view, it may be desirable to let some of these losses become larger and let the net breeding gain go to zero. However, here we are interested in the maximum allowable parasitic capture by other necessary materials and hence the losses resulting from other sources are pushed down to their lowest reasonable levels. Leakage losses can perhaps be kept to about 0.03 to 0.05 per fuel atom destroyed by making the reactor sufficiently large and having a reflector, although in some recent designs leakage losses as large as 0.2 have been considered. The total number of delayed neutrons per fuel atom destroyed amounts to only about 0.007 in uranium-233 fission; hence, not more than a loss of about 0.005 should be involved. The neutron equivalent of the chemicalprocessing fuel losses and the neutron losses from absorption by fission and corrosion products in the core or in a oneregion reactor, should be considered together, since the more frequent the processing for removal of neutron “poisons” the larger the processing losses of fuel atoms and the smaller the parasitic capture loss. Hence, the processing frequency should be optimized for the neutron flux-processing period product. Calculations by Halperin and Stoughton ( 5 ) using several simplifying assumptions, indicated that about 0.03 is a reasonably conservative estimate for the two types of losses combined for aqueous homogeneous reactors. In the blanket of a two-region breeder, these two losses combined should be lower because of the smaller amount of fissioning and the fact that the same uranium-233 atoms are processed only once during their experience in the blanket. Perhaps 0.01, or 0.02 at most, is a reasonable estimate. The losses from build-up and absorption by uranium-234, -235, and -236 have been considered by Visner (77) and by Halperin and Stoughton ( 5 ) . These losses vary with time and should always be less than 0.01 for both blanket and core combined. In the blanket there will also be losses from neutron capture by thorium-233 and protactinium233; absorption by uranium-233 may be assumed to give zero net losses, to a first approximation at least, in view of
the neutron reproduction of fissioning. The main sequence of reactions in the blanket is
p ThZ3’( n , ~ Th233 ) -pa233 27.4 23.3 m d
U233
Each member of the -233 chain will absorb neutrons a t a rate equal to the product of its concentration, cross section, and neutron flux. The capture by thorium-233 and protactinium-233 represents a double loss, since both a neutron and a potential uranium-233 atom are lost in the process; the products of both these reactions form uranium-234 by beta decay: Th234
P 24.1 d
1.1 m; 6.7 h
material in the two-region reactor might vary considerably, depending on its composition and thickness. The latter is determined by the mechanical stresses involved and the properties of the material used. The losses caused by anions, moderator, and coolant in the core and blanket of a two-region homogeneous breeder must be kept below 0.31 minus L, the total of all losses already covered. That is, to an adequate approximation, assuming that all species in each region see the same flux, Loss in fuel region S 0.31 L, or
-
+ loss in blanket region
U234
About 10% of the time, neutron absorption by uranium-233 results in uranium234 and about 90% of the time in fission products and extra neutrons. Hence, while there may be no net loss from absorption by uranium-233 in the blanket, there will be some additional uranium234 produced. Any uranium-234 in the uranium-233 product added to replenish the reactor fuel causes an additional small loss, less than 0.01. This includes the effects of the heavierisotope build-up from uranium-234 ( 5 ) . If the loss caused by thorium-233 or protactinium-233 capture is small, as it must be to breed, and if the concentration of each species is constant as in a continuously processed reactor, then the loss may be approximated as twice the ratio of neutron capture to beta decay-i.e., 24u/X where C$ is the neutron flux, u the capture cross section, and X the decay constant. Hyde, Bruehlman and Manning (8) reported 1350 barns for the effective neutron capture cross section of thorium-233 based on the thermal cross section of thorium-232. Halperin, and others (6) reported 140 barns for the protactinium233 cross section for reactor neutrons. For a fluidized blanket, the volume average flux, averaged over the entire blanket system, can be used in this expression. The losses from thorium-233 and protactinium-233 capture are only G X and, assuming protactinium is not removed by chemical processing, 0.10 at an average flux of IOi4 neutrons per sq. cm. per second. Actually, the average blanket flux in a two-region breeder is expected to be even lower than this. Thus, protactinium-233 capture losses can be lowered either by lowering the average blanket flux or by chemically removing the protactinium-233 from the blanket. For processing times of IO, 30, and 100 days, the total losses from protactinium233 capture would amount to about 0.013, 0.032, and 0.067, respectively, at a flux of The losses resulting from neutron absorption by the core tank or boundary
2
N23Qla(23)
+
50.31
-3 Nozaa(o2)
-L
(1)
Here Niand cr; refer to concentrations and cross sections of all species in the fuel region except uranium-233; N i and u i to concentrations and cross sections of all species in the blanket except 3 ) the conthorium-232; N23 and ~ ~ ( 2 to centration and absorption cross section of uranium-233; N O Zand ~ ~ ( 0 2to) the same quantities for thorium-232. In a one-region breeder reactor, the fuel destroyed is approximately equal to the absorption by thorium (the latter being slightly larger). The expression analogous to inequality (Equation 1) becomes
where subscript i refers to all species other than uranium-233 and thorium232. Using these approximate expressions and the data in Table I, an estimate can be made of the maximum tolerable concentrations of moderator and other substances relative to thorium which might be in the blanket. Reasonable values of L for aqueous homogeneous reactors may lie in the range of 0.1 to 0.3. Values of I for Aqueous Homogeneous Reactors Losses Range Absorption by higher uranium isotopes in core and blanket 21.0
330.0 0.56 43700.0 >5250.0 >105.0 >0.37 2.1 17.0 4.6
P
S
c1
CI3'
A Ca Zn Rb SI
Y Zr Nb
Sn Ba Ce Pb Bi
Max. Ratio 8.1
5.5 2.1 0.033 1.9 1.7 2.4 1.0 1.5 0.90 0.76 5.8 0.95 1.8 0.90 1.5 6.2 33.0
99.75% H2.
be greater than that of the lighter nuclides considered. For these reasons, any system under consideration should not be rejected for breeder purposes unless the amounts of other atoms required are about twice as great as those in Table 11. For example, it may be possible that a thermal breeder could be devised using thorium as thorium chloride-37. O n the other hand, any proposed system should not be considered definitely feasible from a neutroneconomy standpoint unless other atoms present, or their combined equivalents, are appreciably less than the maximum indicated by Table 11. I n a two-region breeder, losses due to anions, moderator, and coolant in each region must be added together. Because uranium-233 has a much higher absorption cross section (about 585 barns) than thorium, anion atoms associated with it should be relatively unimportant, for the same anions at least. O n the other hand, because the concentration of uranium-233 in the moderator or coolant should be lower than the corresponding concentration of thorium, the losses from absorption by moderator and coolant may be roughly the same in the two regions. If we assume that losses due to capture by anions, moderator, and coolant in the core are about one half those in the blanket, then the maximum ratios given in Table I1 should be reduced by a factor of about 2/3 for a two-region breeder on a comparable basis. For fast reactors, absorption cross sections will depend greatly on the exact neutron energy spectrum, generally increasing rather rapidly with decreasing neutron energy. The fast-neutron cross sections are roughly equal to the geo-
INDUSTRIAL AND ENGINEERING CHEMISTRY
metric cross section, increasing roughly with increasing atomic number (3, 7). Lead and bismuth are striking exceptions in a plot of absorption cross sections us. atomic number, because they deviate on the low side. They have been seriously considered as fastreactor coolants because they have low cross sections, relatively low melting points, and poor moderating properties. Other proposed coolants for fast reactors include sodium and a low melting sodiumpotassium alloy which has actually been used in the experimental breeder reactor developed a t the Argonne National Laboratories. At a few tenths of an m.e.v.-neutron energy, sodium has a cross section of a few tenths of a millibarn; potassium, lead, and bismuth a few millibarns, and thorium a few tenths of a barn. At fast-neutron energies, the 7 value of uranium-233 is about 2.5 rather than 2.31. Assuming equal allowable losses in the fuel and blanket and a loss of 0.1 in the structural boundary materials, the maximum allowable sodium-thorium atomic ratio would be about 200, and that for potassium, lead, or bismuth to thorium about 20, considering only the parasitic capture problem. Actually the maximum allowable sodium-thorium ratio may depend more on the desired average energy, because sodium will moderate neutrons appreciably. Chemical and Engineering Restrictions
Besides having low neutron cross sections, the materials going into a breeder reactor must meet a number of other requirements. They must be stable at reactor temperatures and under the existing levels of nuclear radiations. Fluid material should not corrode or erode any parts of the system at an appreciable rate, because in large-scale power reactors, the cost of replacement of parts under high levels of radioactivity is very high. Care must be taken that separated isotopes, used because of their nuclear properties, are not contaminated with natural elements. If flowing solutions are used, the solute must have a suitably high solubility and the viscosity must not be excessive for pumping. Slurries must have suitably low viscosities and not settle out into cakes incapable of redispersal. One-region reactors impose more restrictions on chemical compatibility of fuel and blanket materials than tworegion breeders. For example, it is not feasible, even in principle, to have in the same region fuel as a fused salt and the blanket in aqueous solution or suspension. I n fact, according to McBride (12), it is not possible to keep uranium-233 as uranyl sulfate or uranyl fluoride in aqueous solution a t 250' C. in the presence of a thorium oxide
slurry; the uranium is soon incorporated into the slurry particles. Of course, even in two-region reactors, there are practical restrictions on divergence of properties exhiljited by the fuel and blanket. If the two regions exhibit widely different corrosion properties a "double" core tank or boundary material may be necessary-Le., made of two sheets of different metals closely connected, one resistant to the fuel medium and the other to the blanket medium. If operating temperatures of the fuel and blanket are appreciably different, such a core tank may be required with a space or insulator between the two metal sheets. AI1 these factors must then be taken into account in designing a suitable blanket system.
Systems Considered Aqueous Systems. From an engineering design standpoint, one of the more attractive systems for a breeder blanket is a circulating aqueous solution of a thorium salt. Of the systems investigated, however, only two have the necessary thermal stability, low viscosity, and sufficiently low neutron cross sections of anion atoms-Th(NOa)d excess " 0 3 (using separated N16) and Th~(P04)4 H3P04 with a P04-Th ratio of 10. The phosphorus-thorium atomic ratio is higher than the limiting value of Table 11, but because the resonance absorption of thorium is appreciably larger than that of phosphorus, a ratio of 10 may be acceptable. Thorium halides, except fluoride, are soluble a t room temperature and are probably thermally stable a t elevated temperatures; however, the anions in these salts show excessive neutron capture. I t is just possible that chloride using separated chlorine-37 could be used if all other neutron losses could be reduced to a few per cent. Other mixed anion system have been considered but none has shown promise. Thorium nitrate is soluble and its solutions, except perhaps a t low concentrations, are stable at room temperature. As the temperature is raised, however, there is a tendency toward hydrolysis and precipitation a t lower concentrations, while nitrogen dioxide fumes can be seen above saturated solutions a t about 184' C. (74). By adding excess nitric acid in a closed system, hydrolysis can be suppressed and precipitation prevented, even a t the higher temperatures, while nitric acid comes to equilibrium with a small amount of decomposition products in the vapor phase. Solutions containing 400 grams of thorium per liter with NOS-Th ratios of 5.47 and 6.65 are stable against hydrolysis u p to temperatures between 300" and 340' C. (75). Crystalline solids which appear in con-
+
+
centrated solutions containing excess acid are reversible in solubility as contrasted to the apparent irreversibility in precipitation of hydrolysis products from thorium solutions. Composition of the crystalline solids may correspond to acid salts. Basic thorium nitrates appear to be quite soluble a t room temperature but precipitate presumably as the hydroxide or basic salts, a t less than 100" C. If the nitrate system is employed, separated nitrogen-15 is needed and incorporation of special facilities into the breeder system may be required to reconvert nitrogen to nitric acid. Although apparently stable phosphate solutions containing up to 1100 grams of thorium per liter with phosphate ratios of 5 to 7 can be prepared, the high viscosities of these solutions leave considerable doubt as to their applicability. However, solutions having a P04-Th ratio of 10 with a total thorium concentration of 400 grams per liter are stable a t 250' to 300' C. and have a viscosity little higher than that of concentrated phosphoric acid a t room temperature. Efforts to lower the viscosity of thorium phosphate solutions containing POd-Th ratios of 5 to 7 without precipitating thorium have been unsuccessful. Methods tried involved the addition of hydrofluoric acid, sulfuric acid, and nitric acid. In all cases precipitation occurs before any significant improvement in properties becomes evident. The phosphate system attacks both titanium and zirconium a t relatively low temperatures and except for the possible use of a thin platinum liner, no satisfactory container is yet known for this system as far as corrosion stability is concerned. Other Aqueous Solutions. Other aqueous systems investigated include thorium solutions involving sulfate and fluoride; phosphate and fluoride; sulfate, phosphate, and fluoride; and others involving sulfate, selenate, lithium, and magnesium. I n all these, the solutions are either unstable a t higher temperatures or it is impossible to dissolve sufficient thorium to produce a suitable solution. The thorium-lithium carbonate system under carbon dioxide pressure has also been suggested (78) but has not been investigated. Because of difficulty in finding a suitable thorium solution for an aqueous thermal breeder blanket, much development work has been focused on aqueous slurries of thorium compounds, mostly with slurries of thorium oxide. Attention has been directed toward producing a reasonably nonabrasive and noncorrosive thorium oxide slurry containing 1000 to 1350 grams of thorium per kg. of water that has good handling properties a t 250" C. A product, probably adequate from an engineering and chemical standpoint, may be prepared
by a three-stage calcination of thorium oxalate (75). Much engineering development work has been done at the Oak Ridge National Laboratory on pumping slurries at 250' C. While there is no reason to believe that a satisfactory slurry cannot be devised, certain problems must be solved before a satisfactory slurry is a certainty. For example, some slurries containing thorium oxide produced by caIcination a t a temperature as low as 650" C. became creamy on pumping. These slurries, containing 1000 grams of thorium per kg. of water tended to become pseudoplastic upon standing a t room temperature. The increased viscosity after pumping was believed to be caused by degradation of slurry particles and subsequent flocculation of fragments (72). Localized caking, and in one case caking throughout an entire circulating loop, have also been observed and present the most serious problems so far encountered. Thorium-uranium systems where thorium is in a slurry and uranium in solution would offer economic advantages for use as a blanket or a fuel (72). I n principle, thorium could be readily separated from the uranium and the supernatant could be processed continuously for the uranium-233 leached from the slurry solids. Hence the bred fuel could be used sooner and inventory costs for the initial change of fissionable material in the core could be reduced. However, there are serious disadvantages to thorium-uranium one-region systems. Corrosion-erosion effects might be more serious. Moreover, heating a thorium oxide slurry at 250' C. with a few per cent of uranium as the sulfate, oxide, carbonate, or fluoride has been found (75) to cause uranium to go into the slurry particles and not be removable with a nitric acid leach. I n addition, because of protactinium-233 build-up, the entire chemical system may need processing more frequently than would the blanket in a two-region reactor. The proposed processing cycle for the blanket in a two-region reactor considered a t ORNL is once in 130 days a t an average blanket flux of 5 X The possibility of using dispersed thorium oxide in water was considered by Kraus and Phillips (9). The material, prepared by precipitating hydrous thorium oxide from thorium chloride solutions, was filtered with little washing and heated overnight a t 300" C. Similar materials were prepared from thorium nitrate solutions with somewhat lower drying temperatures. The resulting dispersions in water were clear to the eye a t room temperature, but not stable upon prolonged heating a t 250" C. Work on thorium oxide sols is being continued a t ORNL (79). The potential advantage of such a sol is that the particles are small and comVOL. 49, NO. 2
FEBRUARY 1957
205
pletely dispersed; therefore, they should neither settle out in stagnant regions of the blanket system nor cause erosion when pumped. However, a few experiments have failed to show stability of such sols at the elevated temperatures required by a homogeneous reactor system. A blanket of thorium fluoride particles slurried in deuterium oxide has also been considered. It has been found (70)) however, that anhydrous crystalline thorium fluoride abrades 347 stainless steel a t 12 times the rate of thorium oxide prepared by two-stage calcination. Blankets composed of thorium oxide pellets cooled by circulating water have been proposed because in this way it is possible to achieve very high concentrations of thorium. Thorium oxide pellets fired a t 1700" C. are stable for as long as 150 hours in water a t 250' C. (70). The pellets will dissolve in nitric acid solutions containing fluoride ions giving a dissolver solution suitable as a feed for a solvent extraction process. Pellet beds are not being seriously considered at the present time because of engineering difficulties associated with their use. I t is felt that vibrations associated with operating a reactor will cause the pellets to disintegrate slowly. Moreover, there is lack of information on stability of the pellets to radiation.
pumping equipment and heat exchangers to remove heat from the core and blanket. The use of a lithium salt in the reactor requires isotopically pure lithium-7 because lithium-6 has a large thermal neutron cross section. However, 99.95Y0 lithium-7 has the same thermal neutron cross section as sodium or 0.49. Further purification is achieved by burning the lithium-6 in the reactor, although this may be a slow process. Table 111 shows the freezing point of a 507, salt mixture of lithium and beryllium fluorides containing various mole percentages of thorium fluoride, while Table IV shows the kinematic viscosity in centipoises and the freezing point of two other such mixtures.
Table 111. Freezing Point of 50% Salt Mixtures of Lithium and Beryllium Fluorides Containing Thorium Fluoride
Fluidized Solid Blankets
50% LiF50% BeF2,
ThFa,
%
%
100 75 50 25
0
F.P., O
c.
360 740 870
25 50 75
1000
Tuble IV. Viscosity and Freezing Point of Two Lithium-Beryllium Salt Mixtures F. P.,
Fused Salt System A fused salt blanket would have both advantages and disadvantages when compared with other types of blankets. The low vapor pressure of fused salts allows the use of near atmospheric pressures in the blanket even at high temperatures. A fused salt blanket permits continuous chemical processing. However, the melting points of fused salts are high and heat transfer properties are poor compared to liquid metals. Three anionic types of fused salts have been considered for use in a breeder blanket-hydroxides, chlorides, and fluorides. Although moderating properties of the hydroxide system are good, solubility of thorium in hydroxides is low. Moderating properties in the chloride system are poor; also it is necessary to use separated chlorine-37 and the fused chlorides are very corrosive. Nevertheless, work at the Argonne National Laboratories has shown that thorium and protactinium can be separated by fractional distillation of the chlorides (73). The fluoride system containing lithium-7 and beryllium fluoride has good moderating properties and its componental elements have low -neutron absorption cross sections. Hence, of these systems the fused fluoride system appears most favorable. A fused-salt breeder reactor, like the aqueous breeder, can be either one region or two region. The tworegion reactor requires two sets of
206
perature at 50" to 60" C. higher than the freezing point of the mixture. Thermal analysis of the sodiumthorium fluoride system has given data difficult to interpret. The tentative phase diagram of the system indicates that a mixture of 75 and 2570 respectively melts a t about 610" C. while a 50-507, mixture melts at about 735" C. Further studies on this system have been discontinued. I t w-as thought that systems of thorium fluoride with potassium or sodium hydrofluoride which exhibit low melting points at high thorium concentrations, might be suitable for use in a breeder blanket (20). An advantage of this type of fluid breeder blanket is that it might be processed continuously by bubbling fluoride through it. However, upon investigation it was found that decomposition occurred which made the systems appear unsuitable (20).
?I
50% LiF-50% BeFz 69y0 LiF-31% BeFz
22.2 ' 7.11
c.
O
350 505
Increasing the proportion of lithium fluoride and decreasing beryllium fluoride decreases the viscosity by a factor of 3, while the freezing point is increased 155". When lithium fluoride is increased beyond 6770, the viscosity of the mixture decreases little compared to the decrease when the per cent is increased to 677,. Also surface tensions of the mixtures decrease as the per cent of lithium fluoride is increased. Table V shows the freezing points of tw-o different lithium-beryllium fluoride salt mixtures, each containing 10 mole yo of thorium fluoride ( 7 ) which in these mixtures corresponds to nearly 1000 grams per liter of thorium.
Table V. Dependence of Freezing Point of Lithium-Beryllium-Thorium Fluorides Salt Mixtures on Composition F. P., Composition 45% LiF, 45% BeFz, 10% ThF4 65% LiF, 25'% BeF2, 10% ThFa
O
c.
630 500
Three compounds of thorium seem promising in a fluidized blanket systemthe oxide, carbide, and fluoride. One type of blanket proposed consists of the oxide mixed with some moderatore.g.) beryllium oxide or graphitefluidized with an inert gas such as argon or helium ( 4 ) . I t seems preferable to include in the blanket more moderation than is obtained from thorium oxide alone. This becomes more important as greater fractions of the total number of fissions occur in the blanket. Thorium oxide or carbide can be incorporated in the graphite by impregnation or a properly prepared physical admixture of the thorium compound and graphite may be used (4). i\nother type of fluidized-solid blanket has been proposed (76) which consists of thorium fluoride particles fluidized in helium and continuously processed for uranium-233 and protactinium-233. The process consists of separating uranyl hexafluoride and protactinium fluoride as the blanket stream passes through a hydrogen-fluoride flame. However, development work on the thorium fluoride fluidized blanket has been temporarily suspended ; the time and money needed to prepare a large bath of spherical, torch-product thorium fluoride appeared too great in view of the relative importance of blanket development. Liquids such as water and deuterium oxide as well as inert gases such as helium and argon, may be used to fluidize the solids in a blanket of this type. Liquids may have certain advantages and should be further evaluated in a more complete examination of a fluidized-solids blanket.
Slurries i n Liquid Metal T o avoid freezing the salt in the blanket, it is necessary to maintain the tem-
INDUSTRIAL A N D ENGINEERING CHEMISTRY
Another proposed type of breeder blanket (27) consists of thorium bis-
muthide slurried in molten bismuth metal or in molten bismuth-lead alloy. Two blanket designs have been suggested-one uses thorium dispersion both as a coolant and as a mobile blanket; the other uses dispersion in a stationary external blanket. Thorium dispersions may be prepared in four ways: direct preparation by heating the metals in a graphite crucible to 1200’ C. for an hour and then cooling; exfoliation where chips of thorium are added to the molten low-melting bismuth or alloy at a low temperature; decomposition of thorium hydride in the molten metal or alloy; and electrolysis of thorium into the molten melt. For processing, the blanket material may be diluted with bismuth a t high temperature so that all the thorium bismuthide goes into solution. Two methods have been suggested for extracting uranium and protactinium from the molten solution. In the more promising method, they are removed by a modified fused-salt extraction (27). Also they can be extracted from thorium-bismuth alloys by molten aluminum. However, although protactinium concentrates in the aluminum, it is not possible to make this the basis of a practical separation, since removal of protactinium from the aluminum is too difficult (22). As a structural material, zirconium is preferable to titanium in liquid bismuth loops, because it shows less mass transfer (71). As in mercury systems, magnesium can be used to reduce the oxygen content of liquid-metal slurry systems.
Beds of Thorium Fluoride A blanket of thorium fluoride processed in situ for removal of protactinium fluoride and/or uranyl hexafluoride in a stream of hydrofluoric acid and fluorine is no longer seriously considered (76). In view of the low thermal conductivity of thorium fluoride powder, heat removal from the blanket would be very difficult. Also, low rates of diffusion appear to make continuous removal of volatile protactinium fluoride and uranyl hexafluoride from the crystallites in the blanket unfeasible. In order to overcome low diffusion, frequent reconversion of the system, thorium oxide and fluoride, by hydrolysis-hydrofluorination cycles a t high temperatures has been attempted, but presented serious corrosion problems.
Thorium Metal The only thorium successfully irradiated and processed for uranium-233 has been thorium metal canned in aluminum. Processing for uranium-233 involves dissolution of the slugs in aqueous solutions; however, recasting of the metal and refabrication of slugs is
costly. If the metal is used in a breeder blanket system, a cheaper method of repurification may have to be developed. One possible method of repurification is zone melting, which has been tried in the case of uranium. In zone melting, solutes can be concentrated a t the ends of a bar of material by repeated unidirectional passage of a small molten zone. This process is based upon the fact that many solutes are preferentially distributed into either solid or liquid phase. Another possible method of repurifying thorium is by slagging ( 2 ) . This method is being studied in connection with the purification of the plutoniumuranium alloy to be employed in the power breeder reactor (PBR). In slagging, the metal or alloy is melted in a refractory oxide crucible. The crucible materials acting as solid oxidants, concentrate certain elements (fission elements in the case of PBR) by incorporating them into an oxide slag layer. Volatile impurities are lost by diffusion. A breeder blanket bed composed of 2-cm. rods of thorium on a 10-cm. pitch was proposed for an early homogeneous reactor to be built at the Clinton Laboratory. The rods could be surrounded by a moderator such as graphite or heavy water. A blanket of this type would be ideal from the standpoint of having no neutron absorption by anion atoms.
(2) Foltz, J. R., Rosen, F. D., Gardner, W. J., Preprint 145, 311, Nuclear Engineering and Science Congress, Cleveland, Ohio, December 1955. (3) Haines, G., Way, K., Oak Ridge National Laboratory, Oak Ridge, Tenn., AECD-2368, October 1948. (4) Halik, R. R., Beckberger, L. H., Haibeck, J. M., Nealia, J. E., Northrup, W. A., Rees, D. R., Robba, W. A., Oak Ridge National Laboratory, Oak Ridge, Tenn., ORNL-CF-541-81, January 1954. ( 5 ) Halperin, J., Stoughton, R. W., Oak Ridge National Laboratory, Oak Ridge, Tenn., ORNL-1368, September 1952. ( 6 ) Halperin, J., Stoughton, R. W., Ellison, C. V., Ferguson, D. E., Nuclear Science and Engtneertng 1 , 1 (1956). ( 7 ) Hughes, D. J., Harvey, J. A., “Neutron Cross Sections,” BNL-325 (Supt. Documents, Washington 25, D. C.), July 1955. (8) Hyde, E. K., Bruehlman, R. J., Manning, W. M., Argonne National Laboratory, Lemont, Ill., ANL4165 (1948). (9) Kraus, K. A., Phillips, H. O., Oak Ridge National Laboratory, Oak Ridge, Tenn., private communication, January 1956. (10) Lyon, R. N., Oak Ridge National Laboratory, Oak Ridge, Tenn., ORNL-1554 (July 1953). (11) Lyon, R. N., Zmola, P. C., Oak Ridge National Laboratory, Oak Ridge, Tenn., ORNL-CF-54-6-29 (June 1954). (12) McBride, J. P., Oak Ridge National Laboratory, Oak Ridge, Tenn., ORNL-1605, September 1953, ORNL-1943, September 1955. (13) Malm, J. G., Fried, S. M., Argonne National Laboratory, Lemont, Ill.,
Summary
(14) Marshall, W. L., Gill, J. S., and Secoy, C. H., J . Am. Chem. Sac. 73 4991 (1951); Reactor Handbook, vol. 2, Chap. 4.3; AECD-3646, ISuDt. Documents, Washington - 25, D. b.),May 1955: (15) Marshall, W. L., Secoy, C. H., Oak Ridge National Laboratory, Oak Ridge. Tenn., ORNL-1658, February’l9’54. . (16) Miles, F. T., Wiswall, R. H., Heus, R. J., Hatch, L. P., Nucleonics 12, No. 7, 26 (1954); Chem. Eng. Prog. Symposium Ser., 50, No. 12, 173 (June 1954). (17) Reactor Handbook, vol. 2, Chap. 4.2, AECD-3646, May 1955. (Supt. Documents, Washington 25, D. C.). (18) Secoy, C. H., Oak Ridge National Laboratory, Oak Ridge, Tenn., private communication, January 1956. (19) Sweeton, F. H., Oak Ridge National Laboratory, Oak Ridge, Tenn., private communication, January 1956. (20) Vogel, R. C., Stein, L., Linzer, F. R., Argonne National Laboratory, Lemont, Ill., ANL-5422, May 1955. (21) Williams, C., Miles, F. T., Nucleonics 12, No. 7, 11 (1954); Teitel, R. J., Gurinsky, D. H., Bryner, J. S., Zbid., 12, No. 7, 14 (1954); Chem. Eng. Progr. Symposium Ser., 50, No. 11, 245; No. 13, p. 11 (June 1954). (22) Wiswall, R. H., Brookhaven National Laboratory, Upton, L. I., New York, BNL-336, February 1955.
ANL-4411, 4490, 4545, 4593, 4667.
The thorium blanket systems which have received serious consideration have been briefly described. The only blanket form demonstrated to work adequately consists of aluminum-jacketed metal slugs, although the current processing and refabrication methods used for such slugs are undesirably elaborate, Of the other systems, thorium oxide slurry in deuterium oxide and thorium nitrate (using nitrogen-1 5 ) dissolved in deuterium oxide now appear in long range to be the most promising. Fused salts offer attractive possibilities, although the corrosion problem presents a serious deterrent. A slurry of thorium bismuthide in liquid bismuth and fluidized solids in gas streams both offer promise, although they have not received sufficient experimental attention to make possible a rigorous comparison with the aqueous systems. Since no completely satisfactory blanket has yet been devised and developed, more work on devising new systems and more development work on currently suggested systems should be encouraged.
literature Cited (1) Barton, C. J., Oak Ridge Nztional Laboratory, Oak Ridge, Tenn., private communication, January 1956.
RECEIVED for review May 31, 1956 ACCEPTED September 9, 1956 VOL. 49, NO. 2
FEBRUARY 1957
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