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An International Spent Nuclear Fuel Storage Facility: Exploring a Russian Site as a Prototype - Proceedings of an International Workshop UTILIZATION OF HIGH-LEVEL WASTE
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An International Spent Nuclear Fuel Storage Facility: Exploring a Russian Site as a Prototype - Proceedings of an International Workshop Types of High-Level Radioactive Wastes Formed as a Result of Dry Methods of Spent Fuel Regeneration and Technologies for Their Management Valentin B. Ivanov Institute of Geology of Ore Deposits, Petrography, Mineralogy, and Geochemistry Russian Academy of Sciences At present for the nuclear industry to adopt new nuclear fuel cycle technologies and for public opinion to be prepared to accept their implementation, it is absolutely essential that all fuel cycle-related problems without exception be resolved, especially those problems involving the management of radioactive wastes. In proposing new approaches to spent fuel regeneration, suppliers of such technologies must ensure that all aspects of the separations process are comprehensively addressed. This sort of approach was taken in the development and testing of dry technologies for spent fuel regeneration. The pyroelectrochemical technology, which is based on the use of salt melts, is the most ready for industrial application, and has been developed and tested on a semi-industrial scale at the Scientific Research Institute of Atomic Reactors (NIIAR). A top priority in any technology used in the nuclear fuel cycle is its safety, and this means not only nuclear and radiation safety but also general industry safety as well. A safety analysis has shown the fundamental advantages of the pyroelectrochemical regeneration technology in comparison with currently used water-based methods for reprocessing spent nuclear fuel. The basic safety aspects and physical-chemical properties of the process on which it is based are presented in Box 1. This technological process is also notable for the fact that it is self-protected against the unauthorized removal of fissile materials, as it is characterized by the presence of powerful gamma radiation in all its stages. Therefore, an additional requirement for this type of regeneration process is that it should facilitate the separation of the greatest possible quantity of fissile materials with the least possible removal of fission products. Such an approach also minimizes the amount of radioactive materials, and especially radioactive wastes, that are removed from the fuel cycle.
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An International Spent Nuclear Fuel Storage Facility: Exploring a Russian Site as a Prototype - Proceedings of an International Workshop BOX 1 Analysis of the Safety of Nonwater-Based Processes for Spent Fuel Regeneration Key safety aspects: does not use water or organic solvents; all chemical operations conducted in one closed apparatus; reprocessed product is practically ready for use after removal from apparatus; high-level wastes are in solid compact form Radiation safety: crystalline product base; operational environment is a molten salt mixture Nuclear safety: no moderators or neutron reflectors; processes take place discretely; no possibility of self-sustaining chain reaction in the event of serious accidents Technical and chemical safety: no radiolysis of the chemical environment; no pieces of equipment vulnerable to fire; explosive gases and substances not used Protection against dispersion of radiation: placement in protective chambers; automated and remote controls; chemical stability of regenerated product BOX 2 Composition of Fuel Consisting of Plutonium Extracted During Reprocessing of VVER-440 Spent Fuel and Combined Uranium Plutonium oxide with the following isotopic composition: Pu-238 1.0% Pu-239 67.4% Pu-240 21.5% Pu-241 6.9% Pu-242 3.1% Am-241 in an amount equaling 2.8 kg per metric ton of plutonium Remainder of fuel load: combined U-238 in dioxide form and 5–10 percent in metal form, U-235 less than 0.2 percent of the total mass of uranium To illustrate, I present the following data on the composition of plutonium-based fuel separated from spent fuel from a water-moderated water-cooled power reactor (VVER-440), as well as experimental data obtained as a result of the regeneration of spent fuel removed from a BOR-60 fast reactor, with this material being analyzed for its potential fuel content to be reloaded into a reactor (see Boxes 2, 3, and 4). It is obvious that after regeneration, the fuel contains not only fission products but also minor actinides. These elements do not hinder the phys-
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An International Spent Nuclear Fuel Storage Facility: Exploring a Russian Site as a Prototype - Proceedings of an International Workshop BOX 3 Composition of Fuel Obtained Through Pyroelectrochemical Regeneration of BOR-60 Spent Fuel (Approximately 28 Percent Burn-Up Rate) Plutonium oxide and uranium oxide with the following isotopic composition: Pu-238 1.06% U-234 0.65% Pu-239 63.97% U-235 48.71% Pu-240 26.04% U-236 3.84% Pu-241 4.72% U-238 46.80% Pu-242 4.21% (For BN-600/800 reactors only U-238) Am-241 in an amount equaling 5.4 kg per metric ton of plutonium and 0.376 kg per metric ton of uranium; the regenerated uranium output also includes Cm-242 (1.2 g per metric ton) and Cm-244 (14 g per metric ton) Fission products (Curies per gram of PuO2): Ru(Rh)-106 1.3 · 10−1 Sb-125 2.1 · 10−3 Cs-134 1.9 · 10−5 Cs-137 2.0 · 10−4 Ce(Pr)-144 5.3 · 10−2 Eu-152 1.4 · 10−3 Eu-154 5.8 · 10−4 Eu-155 2.0 · 10−3 Corrosion elements: Mn-54 at 5.9 · 10−5 and Co-60 at 3.2 · 10−6 Curies per gram of PuO2 Cation additives 8.86 mass percent of the mass of PuO2 and by nomenclature Si, Fe, Mg, Cr, Zr, Mo, Gd, Ni, Pb, Ag, Y, La, Tb, Dy, B, Ga, Be, Ca, Ce, Pr, Eu, Ti, Cu, Na, Pd, Nd, Sm, Sc BOX 4 Composition of Fuel Obtained Through Pyroelectrochemical Regeneration of Spent Fuel from VVER-1000 or RBMK Reactors Left after regeneration: U 99.9% Pu 99.96% Am and Cm in approximately the same quantity per kg as in fuel prepared on the basis of spent fuel from BOR-60 reactors Standard purification coefficients for fission products: Ru-106 13 Ce-144 19 Sb-125 120 Eu-154 and -155 33 Cs-137 30,000 Obviously a substantial part of the fission products and minor actinides remains in the fuel to be reloaded into the active zone of the reactor.
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An International Spent Nuclear Fuel Storage Facility: Exploring a Russian Site as a Prototype - Proceedings of an International Workshop ics of the reactor’s operation, that is, they may pass through the cycle repeatedly without increasing the volumes and radioactivity of the materials removed. This is a great advantage of this technology, inasmuch as with the use of remote-controlled transport procedures this fulfills the conditions for self-protection of the fuel and reduces the quantity and activity of the radioactive substances removed from the cycle. Dry regeneration also results in the formation of radioactive wastes. The sources and forms of these wastes are presented in Box 5. These types of wastes are subjected to special processing so that they may be placed either immediately or after a certain holding period into deep geological formations for permanent storage. Thus, enterprises manufacturing regenerated fuel for the new generation of fast neutron reactors will not accumulate radioactive wastes. The technologies and procedures for primary processing of radioactive wastes formed during the pyrochemical regeneration of spent fuel are shown in Box 6. BOX 5 Sources of Radioactive Wastes Created During Pyroelectrochemical Regeneration of Spent Nuclear Fuel Liquid wastes: soda solution (gamma and alpha nuclides suitable for underground burial in liquid form); water solutions after flushing of precipitates (after steaming, the salts are ready to be returned to the start of the process) Gaseous wastes: technical gases from the chlorinator-electrolyzer (radioactivity 95 percent due to Sb-125 aerosols; gases are purified with a filter, absorption column, and two stages of treatment); air from protective chambers Solid radioactive wastes and products: nontechnical (equipment, pyrographite items, filters, fuel rod casings); technical (spent electrolyte, steamed salts, phosphate precipitate) BOX 6 Primary Processing of High-Level Radioactive Wastes Concentration of fission products and additives in a phosphate precipitate and a spent salt electrolyte Vitrification of the phosphate precipitate in aluminofluorophosphate glass Vitrification of the phosphate precipitate in aluminofluorophosphate glass along with the spent electrolyte Conversion of the phosphate precipitate into a monazite-like structure and conversion of the spent electrolyte into a ceramic
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An International Spent Nuclear Fuel Storage Facility: Exploring a Russian Site as a Prototype - Proceedings of an International Workshop The specific characteristics of high-level radioactive wastes formed during the regeneration of spent fuel with a burn-up rate of 21 percent of the heavy atoms are shown in Table 1. It should be noted that the output volume of contaminated electrolyte, 2.5 kg/kg, is valid only for this particular experiment. In real-life application of the technology the electrolyte undergoes regeneration and the output volume that must be disposed of is on the order of 10 times less. One method of disposing of the wastes that are created is vitrification. This procedure is carried out directly with the separated product without any intermediate stages. The characteristics of the glass matrices that are produced are presented in Table 2. The high radiation resistance of the matrix and the low rate of leaching are particularly noteworthy. When wastes are placed in a ceramic form, the leaching rate is slightly accelerated, but this process provides a substantially greater degree of thermal stability, which is especially important in the regeneration of spent fuel that has been stored for a relatively short period of time (see Table 3). The technology for radioactive waste reprocessing is illustrated in Figure 1. As the schematic clearly shows, the products separated out by the process undergo only two or three operations, and the wastes are then ready for burial. A full listing of the various types of wastes and their isotopic composition is provided in Table 4. One can see that the bulk of the radioactivity is concentrated in the phosphate precipitate (Ce(Pr)-144), the spent electrolyte, and the steamed salts (Cs-134 and -137). All of these substances are contained in the same chlorinator-electrolyzer device both during the regeneration process and after its completion. They are not transported and consequently do not contaminate any other equipment or hot chambers. All of this is very significant for reducing the volume of radioactive wastes created during the decontamination process. The fluoride gas technology for regenerating spent nuclear fuel is less ready for industrial application. Nevertheless, a certain amount of research has been completed with regard to the management of radioactive wastes created as a result of this process. In regenerating only uranium irradiated fuel, an average of 86 TABLE 1 Characteristics of Solid High-Level Wastes (Spent Fuel from BOR-60 Reactors with a Burn-up Rate of 21 Percent of Heavy Atoms, Stored for Two Years) Type Fuel Output, kg/kg Specific Heat, Wt/kg Temperature of Spontaneous Combustion, °C Phosphate precipitate 0.14 14.8 40 (180 g) Electrolyte 2.5 0.95 30 (2 kg)
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An International Spent Nuclear Fuel Storage Facility: Exploring a Russian Site as a Prototype - Proceedings of an International Workshop TABLE 2 Vitrification of High-Level Wastes (HLW) Using the Pyroelectrochemical Process Type of HLW Type of Glass Matrix Means of Incorporation Phosphate precipitate Pb(PO3)2, NaPO3 Vitrification, T-950°C Spent salt electrolyte NaPO3, AlF3, Al2O3 Vitrification without conversion of chlorides, T-950°C Phosphate precipitate plus spent salt electrolyte NaPO3, AlF3, Al2O3 Joint vitrification without conversion of chlorides, T-950°C TABLE 3 Ceramization of High-Level Wastes (HLW) Using the Pyroelectrochemical Process Type of HLW Type of Ceramic Means of Incorporation Phosphate precipitate Monazite Pressing, kilning, T-850°C Spent salt electrolyte Kosnarite (sodium zirconium phosphate—NZP) Conversion into NZP from melt or water solution, pressing, kilning, T-950°C TABLE 4 Solid Technical Wastes and Products Waste Product Mass, kg Mn-54 Co-60 UO2-1 product after electrolysis of restored melt 0.489 < 4.4 < 1.6 UO2-2 product after electrolysis of acidified melt 1.510 0.3 0.5 PuO2 product after precipitation crystallization 0.504 2.2 0.1 Phosphate precipitate 0.442 24.8 1.3 Spent electrolyte 8.114 — 0.1 Steamed salts 0.968 2.1 0.4 Sublimates 0.495 traces traces Pyrographite materials, filters, fuel rod casings 20 0.8 —
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An International Spent Nuclear Fuel Storage Facility: Exploring a Russian Site as a Prototype - Proceedings of an International Workshop Amount of Wastes Incorporated (percent) Cs-137 Leaching Rate Over 7 Days, g/cm2/day Thermal Stability, °C Radiation Resistance 28 7 · 10−6 400 20 7 · 10−6 400 1 · 107 Gr (γ, β) 1 · 1018 α emitters/g 36 4 · 10−6 400 Amount of Wastes Incorporated (percent) Cs-137 Leaching Rate Over 7 Days, g/cm2/day Thermal Stability, °C Radiation Stability 100 1 · 10−6 850 5 · 108 grays 30–40 3 · 10−6 1,000 (γ, β) 1 · 1019 α emitters/g Specific Activity of Nuclides, GBq/kg on June 1, 1995 Ru(Rh)-106 Sb-125 Cs-134 Cs-137 Ce(Pr)-144 Eu-154 Eu-155 40,700 362.6 0.4 4.1 196.1 2.9 4.8 814 35.2 0.3 3 70.3 51.8 136.9 4,810 77.7 0.7 7.4 1,961 21.5 74 151.7 229.4 16.7 170.2 96,200 888 4,440 2 1.1 207.2 2,072 244.2 6.3 — 777 74 170 1,369 2,664 21.8 144 48.1 484.7 26.3 263 29.6 traces — 21 10.5 4.5 44.5 15.4 0.2 —
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An International Spent Nuclear Fuel Storage Facility: Exploring a Russian Site as a Prototype - Proceedings of an International Workshop FIGURE 1 Vitrification of wastes and preparation of glass ceramics.
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An International Spent Nuclear Fuel Storage Facility: Exploring a Russian Site as a Prototype - Proceedings of an International Workshop BOX 7 Preparation of Solid Wastes with High Specific Activity Spent fuel and sorbent materials are first subjected to a dosing procedure and then smelted in a crucible at a temperature of up to 800°C. Then they are poured into stainless steel containers, which are hermetically sealed and inspected. The outside surfaces of the containers are decontaminated. After this process, the wastes are characterized by adequate radiation stability (no gas separation observed in a container of wastes up to a dose of 4.4 · 105 grays) thermal stability (depending on the proportions of components loaded, the melt temperature varies from 760 to 800°C) chemical stability (leaching rate for Cs-137 from melts of various compositions with high specific activity equals (1.6–8.8) · 10−2 g/cm2 per day) percent of the activity of the fission products is concentrated in the fluorination residues, whereas in regenerating mixed fuel this figure is only 53 percent, which is explained by the characteristics of the technical process of fluorination (increased temperature, large surplus of fluorine). The remainder of the activity is distributed among the chemical and other sorbents and the various tubes and pipes in the processing equipment. The output of fluorination residues totals about 0.2 kg/kg of spent fuel processed, with these residues showing concentrations of mainly long-lived radionuclides (cerium-144, praseodymium-144, cesium-134 and -137, and strontium-90) that form relatively nonvolatile fluorides. The stability of the fluorination residues depends on the activity of the fuel being regenerated and increases over time. All of the solid wastes are highly active powders (fluorination residues) or granules (sorbents), and they require special storage technology. All of these materials can be safely stored without any special preparation in stainless steel containers; however, for long-term controlled storage a technology has been developed for the preliminary smelting of solid wastes. Box 7 describes the process for preparing solid wastes with high specific activity and presents information on their fundamental properties. The release of gas-phase Cs-137 from the wastes increases along with increases in temperature; however, the rate of its volatilization is lower than that for the sorbents and spent fuel wastes separately (that is, the figures go from 1.3 · 10−4 to 1.8 · 10−10 g/cm2 per day). The heat conductivity of the smelted wastes in the temperature range from 100 to 400°C totaled 0.51 to 0.79 Wt/m °C. The temperature of the walls of the stainless steel waste containers must not exceed 300°C in order to avoid notable corrosion of the material of the container casing.
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An International Spent Nuclear Fuel Storage Facility: Exploring a Russian Site as a Prototype - Proceedings of an International Workshop CONCLUSION Research studies and the technologies that have been developed for managing high-level radioactive wastes produced in the regeneration of spent nuclear fuel using dry methods (pyroelectrochemical and gas fluoride technologies) indicate the feasibility of controlled storage of such wastes in simple hermetically sealed vessels for a prolonged period (decades). Radioactive wastes created as a result of pyroelectrochemical regeneration may be placed fairly easily in glass or ceramic forms suitable for permanent burial. A program of research work must be completed to determine the optimal form (for example, chemical composition, type of matrix) for solid high-level radioactive wastes from the standpoint of ensuring the safety of their burial in geological formations. This research is a vital element of the regeneration technologies being developed to facilitate the new closed fuel cycles.
Representative terms from entire chapter: