Uranium metal is very reactive. It tarnishes in air, with the oxide film preventing further oxidation of massive metal at room temperature. However, finely divided uranium ignites spontaneously at room temperature, and massive uranium bums steadily at 700°C, forming U308.

Water attacks massive uranium slowly at room temperature and rapidly at higher temperatures. U02 and UH3 are formed, heat is evolved, and the metal swells and disintegrates. In water-cooled reactors uranium metal must be clad or canned in nonreacting metals such as aluminum, stainless steel, or zirconium. Nitric acid dissolves uranium readily.

Table 6.2 lists the most important isotopes of thorium, together with their properties of greatest significance in nuclear technology.

2200 m/s cross sections Neutron yield

Compounds of hafnium have physical and chemical properties very similar to the corresponding compounds of zirconium, except for the much higher density of hafnium compounds. The melting and subliming temperatures of some hafnium compounds are compared with corre­sponding zirconium compounds in Table 7.8. As the vapor pressures of corresponding hafnium and zirconium compounds are so nearly equal, separation by fractional distillation is imprac­tical. As corresponding compounds, such as HfCU and ZrCU, form solid solutions miscible over the entire composition range and have nearly the same melting point, separation by fractional crystallization is also difficult.

The free energies of formation of corresponding hafnium and zirconium compounds are also nearly equal, so that separation by preferential reaction of one species is difficult, too. Table 7.9 compares the free energies of formation of hafnium and zirconium tetrahalides at 1000 K.

Successful processes for separating hafnium from zirconium take advantage of rather rare occurrences of substantial differences in solubilities of corresponding hafnium and zirconium compounds in water, organic solvents, fused salts, or liquid metals.

Another radiation problem arises from fast neutrons produced in spontaneous fission of the even-mass plutonium isotopes. Half-lives and specific activities for spontaneous fission of the plutonium isotopes are listed in Table 8.17.

In determining the biological hazard from spontaneous fission, it is assumed that each fission produces an average of three neutrons, each with an energy of 2 MeV. Estimated surface fluxes and dose rates are given in Table 8.18.

The Purex process uses a mixture of tributyl phosphate (TBP) and a hydrocarbon diluent to extract uranyl nitrate and tetravalent plutonium nitrate from an aqueous solution containing nitric acid. The Purex process was suggested by the discovery of Warf [W2] in 1949 that tetravalent cerium nitrate could be separated from the nitrates of trivalent rare earths by solvent extraction with TBP. The Purex process was developed by the Knolls Atomic Power Laboratory of the General Electric Company and carried through the pilot-plant stage at Oak Ridge National Laboratory from 1950 to 1952. It was adopted by E. I. duPont de Nemours and Company for the Savannah River plutonium-production plant that company built for the U. S. AEC at Aiken, South Carolina, where the Purex process was put into operation in November 1954. Its success there led to replacement of the Redox process by the Purex process by the General Electric Company at Hanford in January 1956. The Purex process was used in a plant owned by Nuclear Fuel Services, Inc., which that company operated at West Valley, New York, from 1966 to 1972. The plant was designed to reprocess 1 MT/day of irradiated, slightly enriched uranium fuel. It also reprocessed irradiated thorium and irradiated plutonium, with appropriate flow-sheet modifications [R8]. This plant was noteworthy for being the only one to reprocess fuel from privately owned nuclear power plants in the United States.

Although the West Valley plant met all safety and environmental requirements in effect when it first went into operation, in the 1970s the plant was required to meet increasingly strict licensing requirements on permissible radioactive effluents and resistance to ground motion in an earthquake. It was also required to provide facilities for converting plutonium nitrate to oxide and for solidifying high-level acid wastes instead of neutralizing them and storing as liquid. Because of the high cost of retrofitting the plant to meet these later requirements and because its capacity was by then too small to permit it to compete with a larger plant under construction by Allied-General Nuclear Services (Sec. 4.14), the West Valley plant was permanently shut down in 1976 [N8].

A solvent extraction process similar to Purex using TBP was developed by the Commissariat a l’Energie Atomique [Gl] for use in the French plutonium separation plant at Marcoule. Since then, the Purex process has replaced the Butex process at Windscale [W3], has been used in the Soviet Union [SI 1], India [S7], and Germany [S3], and by now is the universal choice for separation of uranium and plutonium from fission products in irradiated slightly enriched uranium. Fuel from the liquid-metal fast-breeder reactor (LMFBR) is also reprocessed by the Purex process, with modifications to accommodate the higher concentrations of plutonium and fission products.

The Purex process has four significant advantages over the Redox process: (1) Waste volumes can be made much lower, as the nitric acid used as salting agent can be removed by evaporation. (2) The solvent, TBP, is less volatile and less flammable than hexone. (3) TBP is more stable against attack by nitric acid. (4) Operating costs are lower.

The Purex process will be described in considerable detail in Sec. 4.

Figure 10.28 shows the principal steps in reprocessing LMFBR fuel. Feed quantities are for a plant fed with 5 MT/day of irradiated heavy metal (uranium plus plutonium). Feed is combined core and blanket assemblies from LMFBRs operated under conditions nearly the same as those on which Fig. 3.34 and Tables 8.8 and 10.20 were based. The head-end steps 1 through 6 follow one alternative of several sketched in Report ORNL-4422 [05].

Fuel assemblies for the core and axial blanket consist of long bundles of stainless steel tubes, each about 0.6 cm in diameter, in which the spent fuel and fission products are sealed.

Table 10.20 Principal differences between irradiated fuel from LMFBR and LWR

Reactor

The lower end of each tube contains irradiated depleted U02, the middle portion irradiated mixed depleted U02 and Pu02, an upper portion irradiated depleted U02, and the top a plenum to accommodate buildup of fission-product gases. The rod bundles are surrounded by a square or hexagonal stainless steel sheath to the top and bottom of which are attached end fittings to direct sodium flow in the reactor and to facilitate handling outside. Fuel assemblies for the radial blanket are of the same length but contain rods of larger diameter charged initially with depleted U02.

In Fig. 10.28 it is assumed that assemblies from the core and radial blanket are reprocessed in the proportion in which they are discharged from the reactor. The average composition of feed to the reprocessing plant then is 10 w/o plutonium, 3.56 w/o fission products, and 86.44 w/o uranium. The 5000 kg of fuel processed per day is associated with 6858 kg of stainless steel and an indeterminate amount of metallic sodium that coats exterior surfaces of the assembly and possibly has penetrated imperfections in some of the fuel rods. Sodium is used as coolant in the LMFBR and is a likely candidate for removing decay heat in shipping irradiated fuel from the reactor to reprocessing.

The first step in Fig. 10.28 is deactivation of sodium coating the outside the fuel rods, either by dissolving it off or converting it to a less reactive sodium compound. In the second step, as much of the stainless steel as possible is removed without permitting fission products to escape. End fittings are removed and fuel rod bundles are extracted from the enclosing sheath, if possible. In the third step, the plenum is sheared from fuel rod bundles, thus releasing some of the fission product gases to a retention system. The portion of the rod bundle containing

fuel and blanket material is sheared into short lengths to facilitate subsequent processing. In step 4, voloxidation, the sheared fuel is heated to 550 to 600°C first in argon, to which is then added an increasing amount of air, to react with possible entrained sodium, convert U02 to U308, and release tritium. In step 5, the fuel is dissolved in 8 M nitric acid to which sufficient gadolinium nitrate, boric acid, or other soluble poison is added to control criticality. Undissolved residues rich in Pu02 are treated with special reagents. In step 6, feed adjustment, nitric acid concentration of solvent extraction feed is brought to З M and plutonium is made tetravalent by addition of N204. In the Purex process, step 7, solvent extraction with 30 v/o TBP is used to separate dissolver solution into high-level waste, decontaminated uranyl nitrate, and decontaminated plutonium.

85 Kr, a 10-year half-life krypton isotope, is currently released from reprocessing plants to the atmosphere. There will probably be no urgent need in terms of radiation dose to the local population to retain 83 Kr. However, in view of a worldwide accumulation of 8S Kr in the atmosphere, krypton recovery from reprocessing plants is required or will be required in the near future.

The krypton disposal problem is characterized by the fact that there is no easy way of converting it into a nongaseous form stable at ambient temperature. There are interesting experiments in progress to fix krypton in zeolites by adsorption under high pressure. In England a pilot plant for krypton implantation in metals is under construction. Never­theless, the containment technique presently envisaged for technical use is pressurization in steel bottles.

There are a number of problems in developing efficient krypton-removal processes. The great portion of xenon present in the noble gas fraction tends to solidify at the krypton condensation temperature and to block the equipment. Small fractions of krypton adsorbed in the pretreatment steps may be lost from the main krypton streams. A mechanical problem is presented by the need to exchange steel bottles for krypton collection without significant leakage.

The annual amount of krypton from a 1400 MT/year reprocessing plant is about 500 kg with a 85 Kr activity of about 2 X 107 Ci. This corresponds to 50 standard bottles at 175 atm pressure with a surface dose rate of 400 rem/h. The temperature may be as high as 150°C. The krypton bottles are to be stored in an engineered facility with dry cooling.

There is some consideration of ultimately disposing of these bottles into the sea. This may be well justified because of the relatively short half-life, the low radiotoxicity, and the chemical inertness of 85 Kr. It may even reduce the 85 Kr hazard in comparison with surface storage of high-pressure bottles. At present, however, the London Convention on sea disposal of radioactive waste permits only disposal of solid waste.

For more efficient fractional extraction of two or more extractable components, the extracting — scrubbing cascade of Fig. 4.4 is employed. Nomenclature for flow rates, concentration, and stage number is shown in Fig. 4.14. With the same assumptions and approach as in Sec. 6.1, a material balance for any one of the components in the portion of the cascade below stage n in the extracting section is

Eyf + (S + F)xf = Eyf_, + {S + F)xf

Л- — Уо = (X%-xf) (4.51)

The extracting-section operating line shown in the McCabe-Thiele diagram of Fig. 4.15 passes through the point (yf, xf) and has the slope (5 + F)jE.

A material balance around the portion of the plant above stage m of the scrubbing section is

5*o + Eysm = Sxsm _ j + £>?

or /„ — yf = ;£(*m_i -*o) (4.52)

E

In Fig. 4.15 this is represented by the operating line for the scrubbing section, which passes through (*o, Уі) and has the slope S/E.

As illustrated in Fig. 4.15, different equilibrium lines can exist for the extracting and scrubbing sections, as might occur if the scrub solution contains a different salting agent

concentration than the feed solution, or from the effect of one extractable component on the distribution coefficients of other extractable components (cf. Sec. 6.6).

In Fig. 4.15 a graphic solution is illustrated for a specified number of stages N and M in the extracting and scrubbing sections, respectively. Proceeding upward by vertical and hori­zontal steps from the point xf, yf for N vertical steps between the extracting-section operating and equilibrium lines, the concentration in the solvent leaving the extracting section is identical with that entering the scrubbing section; i. e.,

У%=А+і (4.53)

where Ух is found at (x%, yfj) on the equilibrium line for the extracting section. The horizontal projection of yf/ onto the scrubbing section operating line yields (xff, yff+1), and this point is then projected downward to the equilibrium line for the scrubbing section.

The number M of equilibrium stages in the scrubbing section specifies the number of vertical steps between the operating line and the equilibrium line, starting at y^+ and ending at У.

By thus determining the extract concentration for one of the extractable components in the feed, a similar graphic or numerical calculation is made for each of the other extractable components so that the composition of the organic product and aqueous raffinate can be determined. When two or more extractable components are each present in sufficient concentration to affect distribution coefficients of the other species, equilibrium lines must be calculated by an iterative procedure similar to that illustrated in Sec. 6.6 for TBP extraction in the Zr-Hf-HN03 system.

The operating lines for a given component in the extracting and scrubbing sections intersect at the feed composition x?. This can be demonstrated by defining xmn, ym„ as the intersection point, such that at the intersection

xmn ~ *n ~ xm -1 (4.54)

and ymn =yf_, =ysm (4.55)

Substituting Eqs. (4.54) and (4.55) into (4.51) and (4.52) and solving, we find that

Fxmn + EyE0 + Sxs0 =(S + F)xf + EyS (4.56)

However, an overall material balance for the extractable component written for the entire separation cascade of Fig. 4.14 is

Fx* + Eyf + Sxf = (S + F)xf + Eyf (4.57)

Comparison of Eqs. (4.56) and (4.57) shows that the operating lines intersect at

xmn = ^ (4-58)

In an extracting-scrubbing cascade with a finite number of stages, none of the aqueous streams entering or leaving a stage is at a concentration as high as xF. The intersection of the two operating lines represents only an extrapolated point that is useful in graphic construction of the operating lines.

An estimate of uranium resources in the United States more detailed and more recent than that of Table 5.17 provided by the U. S. Department of Energy in May 1978 [Ul] is summarized in Table 5.18. “Reserves” corresponds approximately with the “Reasonably assured resources” category of Table 5.17, and “Probable potential resources” corresponds with “Estimated additional resources.” The subtotal at <$30 of 1312 thousand MT in Table 5.18 may be considered an update of the 1361 for the United States in Table 5.17, and the subtotal at <$50 of 1758 thousand MT in Table 5.18, an update of the 1696 in Table 5.17.

“Possible potential resources” are defined [U2] as “those estimated to occur in undis­covered or partly defined deposits in formations.. . productive elsewhere in the same geologic province.” “Speculative potential resources” are defined [U2] as “those estimated to occur in undiscovered or partly defined deposits: 1. in formations … not previously productive within a productive geologic province, or 2. within a geologic province not previously productive.”

To relate these resource estimates to nuclear electric generation, it may be noted that a 1000-MWe pressurized-water reactor operating at 80 percent capacity factor without recycle, on uranium enriched to 3.3 w/o (weight percent) 235 U in an enrichment plant stripping natural uranium to 0.3 w/o 235 U, consumes around 200 MT of uranium per year. Thus the U. S. resource estimate of 1758 thousand MT available at less than $50/lb U308 would keep a 300,000-MWe nuclear power industry in fuel for

Inclusion of uranium in the “Possible” and “Speculative” resource categories would increase the total to 3357 thousand MT and extend the life of a 300,000-MWe nuclear power industry to 56 years.

1.9 Principal Thorium-containing Minerals

Heretofore, most of the world’s thorium has come from monazite in beach sands where coproduction of rare earths, titanium, and zirconium has defrayed much of the cost of extracting thorium. Recent increased interest in thorium as an alternative feed material for nuclear power systems has led to more extensive search for thorium deposits and to interest in other thorium minerals that could be produced if the demand for thorium (and its price) increased. Table 6.12 lists the principal thorium-containing minerals and gives their nominal composition and examples of where they have been found.

1.10 World Thorium Resources

Table 6.13 gives the thorium resources of the non-Communist world as estimated by the Organization for Economic Cooperation and Development [01] in December 1977. The definitions of the two resource categories are the same as for Table 5.17. The production cost of these thorium resources was not stated in [01], but was probably $15/lb Th02, from similar statistics cited by Nininger and Bowie [N3].

Since 1975, renewed interest in thorium as source material for production of 233U has led to extensive prospecting for thorium, discovery of numerous potentially commercial deposits and substantial increase in U. S. resource estimates over those listed in Table 6.13.