Thorium Fuel Reprocessing

In order to make use of thorium as a nuclear resource for power generation, development of efficient separation processes are necessary to recover 233U from irradiated thorium and fission products. The THORium uranium Extraction (THOREX) process has not been commercially used as much as the PUREX process due to lack of exploitation of thorium as an energy resource (157,180). Extensive work carried out at ORNL during the fifties and sixties led to the development of various versions of the THOREX process given in Table 2.6. The stable nature of thorium dioxide poses difficulties in its dissolution in nitric acid. A small amount of fluoride addition to nitric acid is required for the dissolution of more inert thorium (181). 

formed by neutron capture of 232Th, decays to 233U with a half-life of 27.4 days. This necessitates a longer cooling period for the complete recovery of 233U in one step. 

The contamination due to 232U in the recovered 233U product leads to intense gamma radiation, which requires specially designed shielded facilities during fuel reprocessing and fuel fabrication. 

The use of fluoride ion, however, enhances the corrosion of stainless steel equipment. This problem is mitigated by the addition of appropriate amounts of aluminium 

Hexone - U Extraction of 233U Acid deficient Hexone3 (182) 

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Np, and fission products. The Thorex solvent extraction process is generally used to reprocess spent Th-based fuels. As in the Purex process, the solvent is TBP diluted in an appropriate mixture of aliphatic hydrocarbons. Figure 12.9 shows the Thorex process flow sheet used by Kuchler et al. [41] for reprocessing high-burn-up thorium fuel. 

AECL has evaluated some of the basic information and development requirements in some detail (24, 25) and has outlined the type of fuel recycle development program which would be required. It would involve research and development of thorium fuels and fuel fabrication methods, reprocessing, demonstration of fuel management techniques and physics characteristics in existing CANDU reactors and demonstration of technology in health, safety, environmental, security and economics aspects of fuel recycle. 

J. R. Allen and co-workers, eds., Gmelin Handbook of Inorganic Chemistry, Thorium, Suppl. Vol. A3, Technology, Uses, Irradiated Fuel, Reprocessing, 8th ed., Springer-Verlag, Berlin, 1988. 

Having now determined to total amount of nuclear electricity required, the thorium fuel input to the energy amplifiers can be calculated from the design data of Rubbia and Rubio (1996). The thermal output from the prototype design reactor is 1500 MW, with a fuel amount of 27.6 t in the reactor (Fig. 5.42). The fuel will sit in the reactor heat-generating unit for 5 years, after which the "spent" fuel will be reprocessed to allow for manufacture of a new fuel load with only 2.9 t of fresh thorium oxide supply. This means that 2.6/5 t y of thorium fuel is required for delivery of 5 x 1500 MWy of thermal power over 5 years, or 675 MWy of electric power, of which the 75 MWy is used for powering the accelerator and other in-plant loads. The bottom line is that 1 kg of thorium fuel produces very close to 1 MWy of electric power, and 1 kt thorium produces close to 1 TWh. ... 

Liquid-liquid extraction (LLE) systems using neutral phosphorus-based organic compounds have been the subject of extensive study since Warf (1) first reported the use of tributyl phosphate, TBP, as a useful extractant for cerium(IV), uranyl and thorium nitrates. After more than twenty years, liquid-liquid extraction systems (such as the Purex and Thorex processes) employing TBP dissolved in a suitable diluent versus an aqueous HNO3 phase remain the most widely accepted systems for reactor fuel reprocessing. 

Another feature of the thorium fuel cycle which affects reprocessing is the buildup of 232U in the irradiated fuel. 

Thorium makeup requirements for one reactor system, the HTGR (high-temperature gas-cooled reactor), may be estimated from Fig. 3.33. A 1000-MWe HTGR requires 7.4 MT of thorium as feed pet year. Reprocessing recovers 6.8 MT, which can be recycled after storage for 20 to 30 years to permit excess Th to decay. The net thorium consumption of a 1000-MWe reactor then is 0.6 MT/year. Thus, the 441,000 MT of U.S. ThOa thorium reserves listed in Table 6.14 would provide thorium fuel for... 

The calculated elemental composition, radioactivity, and decay-heat rate for discharge fuel are shown in Table 8.7 for the uranium-fueled PWR (cf. Fig. 3.31), in Table 8.8 for the liquid-metal fast-breeder reactor (LMFBR) (cf. Fig. 3.34), and in Table 8.9 for the uranium-thorium-fueled HTGR (cf. Fig. 3.33). These quantities, expressed per unit mass of discharge fuel, are useful in the design of reprocessing operations. For the purpose of comparison, all quantities are calculated for 150 days of postirradiation cooling. 

Figure 10.29 shows the principal steps in applying the Purex process to irradiated LMFBR fuel, step 7 of Fig. 10.28. The flow scheme and the compositions and locations of solvent, scrubbing, and stripping streams have been taken from the process flow sheet of a 1978 Oak Ridge report [Oil] describing a planned experimental reprocessing facility designed for 0.5 MT of uranium-plutonium fuel or 0.2 MT of uranium-plutonium-thoiium fuel per day. As that report gave process flow rates only for the uranium-plutonium-thorium fuel. Fig. 10.29 does not give flow rates for the uranium-plutonium fuel of present interest. This flow sheet shows the codecontamination step, in which flssion products are separated from uranium and plutonium the partitioning step, which produces an aqueous stream of partially decontaminated...

Nuclear Fuel Reprocessing and Related Discharges. Single Compound Radiocarbon Measurements. Uranium-Thorium Decay Series in the Oceans Overview. 

There are many examples of the studies on SLM for nuclear applications in the literature. SLMs were tested for high-level radioactive waste treatment combined with removal of actinides and other fission products from the effluents from nuclear fuel reprocessing plants. The recovery of the species, such as uranium, plutonium, thorium, americium, cerium, europium, strontium, and cesium, was investigated in vari-ons extracting-stripping systems. Selective permeation... 

Chemistry used in the recovery of plutonium from irradiated fuel must provide a separation from all these elements, other fission and activation products, and the actinides (including a large amount of unburned uranium), and still provide a complete recovery of plutonium. The same issues apply to the recovery of uranium from spent thorium fuel. Most of the processes must be performed remotely due to the intense radiation field associated with the spent fuel. As in the enrichment of uranium, the batch size in the later steps of the reprocessing procedure, where the fissile product has become more concentrated, is limited by the constraints of criticality safety. There is a balance between maximizing the yield of the precious fissile product and minimizing the concentrations of contaminant species left in the final product These residual contaminants, which can be detected at very small concentrations using standard radiochemical techniques, provide a fingerprint of the industrial process used to recover the material. 

The MASLWR design benefits from the efforts for PWR/BWR fuel diversification, and can exploit all the same possibilities being considered for use in PWRs. This includes the utilization of mixed oxide (MOX) and thorium fuels and advances in reprocessing as the technologies become available and find acceptance. MASLWR also takes advantage of the activities on PWR spent nuclear fuel cask design, transportation methods, and disposal technologies. 

Adopting a thorium fuel cycle is an intrinsic measure that could hinder the possibility of misuse of nuclear materials for nuclear weapons. Within such cycle, U is produced with a noticeable admixture of a highly radioactive which essentially complicates reprocessing and assembly operations for nuclear weapons. The mixing of thorium with low enriched... 

ThN is the compound with the lowest N Th ratio. In addition to its (former) nuclear interest due to its thermal and radiation stability, it has many very interesting physicochemical properties. Thorium nitrate, the other well-investigated compound, is of importance because it is (in the form of an adduct with tri-n-butylphosphate) the extracted compound when burnt-up thorium fuels are reprocessed. 

An option to use thorium fuel (GT-MHR) and an option to operate in a closed fuel cycle with the reprocessing of the TRISO coated particle fuel (GT-MHR, GTHTR300, HTR-PM) for the GTHTR300 it is indicated that the feasibility of TRISO fuel reprocessing has already been investigated. 

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