Process thorium fuel cycle

FIPS [Fission Product Solidification] A process for immobilizing the radioactive waste products from the thorium fuel cycle in a borosilicate glass for long-term storage. Developed at Kemforschungsanlage Jiilich, Germany, from 1968, until abandoned in favor of PAMELA in 1977. 

The Acid-Thorex process has been used in recent years to recover 233U from neutron irradiated thoria targets. (] M This process uses n-tributyl-phosphate (TBP) in normal paraffin hydrocarbon (NPH) as the extractant and the relative uranium and thorium solubilities in each phase are adjusted by control of the nitric acid concentration. The Acid-Thorex process is the primary candidate for use in proposed aqueous thorium fuel cycles. In this process, uranium is separated from thorium through exploitation of the difference in equilibrium distributions since no usable valence change is available to aid in this separation. 

Interest now is centered on the thorium cycle (23) and laboratory studies have continued to investigate both an adaptation of the Thorex process to CANDU fuel and the application of the amine process to recovering uranium-233 from irradiated thorium. The program to develop and fully demonstrate the thorium fuel cycle has been outlined, and would require about 25 years to complete. However, the current research level will not be expanded until a decision can be taken by the Canadian Government when the information from the current International Nuclear Fuel Cycle Evaluation has been assessed. 

In the thorium fuel cycle the recycled uranium-233 inevitably is contaminated with uranium-232 and its decay products. The first of these, thorium-228, will be contained in any recycled thorium. Thallium-208 in this decay chain emits a very-high-energy gamma ray and for this reason fabrication of recycle fuels in the thorium fuel cycle will have to be done remotely in heavily shielded cells. Conventional fuel febri-cation processes may not be the most economical under these conditions. 

A thorium fuel cycle is also possible for use in nuclear power reactors. This involves using thorium-232 to generate uranium-233, which is capable of undergoing fission processes to generate energy in the form of heat. 

EPA (1984) estimated that about 0.2 Ci of thorium-230 is annually emitted into the air from uranium mill facilities, coal-fired utilities and industrial boilers, phosphate rock processing and wet- process fertilizer production facilities, and other mineral extraction and processing facilities. About 0.084 Ci of thorium-234 from uranium fuel cycle facilities and 0.0003 Ci of thorium-232 from underground uranium mines are emitted into the atmosphere annually (EPA 1984). 

In the first level of the hierarchy, radioactive waste that arises from operations of the nuclear fuel cycle (i.e., from processing of uranium or thorium ores and production of nuclear fuel, any uses of nuclear reactors, and subsequent utilization of radioactive material used or produced in reactors) is distinguished from radioactive waste that arises from any other source or practice. The latter type of waste is referred to as NARM (naturally occurring and accelerator-produced radioactive material), which includes any radioactive material produced in an accelerator and NORM [naturally occurring radioactive material not subject to regulation under the Atomic Energy Act (AEA)]. 

At the end of irradiation in such reactors, fuel consists of a mixture of thorium, uranium containing fissile isotopes, and fission products. Figure 3.33 showed a fuel-cycle flow sheet for an HTGR. The Thorex process has been developed for recovering the uranium and thorium from such fuel cycles, freeing them from fission products and separating them from each other. The Thorex process will be described in this section. When the fuel being irradiated contains appreciable the plutonium thus formed requires that a combination of the Thorex and Purex processes be used. 

There are two breeder reactor fuel cycles. One involves the irradiation of U/ Pu oxide fuel with fast neutrons and is at the prototype stage of development. The other involves the irradiation of Th/ U oxide fuel with thermal neutrons and is at the experimental stage. Fuel from the U/ Pu cycle may be reprocessed using Purex technology adapted to accommodate the significant proportion of plutonium present in the fuel. Increased americium and neptunium levels will also arise compared with thermal reactor fuel. The Th/ U fuel may also be reprocessed using solvent extraction with TBP in the Thorex (Thorium Recovery by Extraction) process. In this case the extraction chemistry must also take account of the presence of Pa arising as shown in Scheme 2. 

In setting up a fuel cycle with transmutation and seeking proper environmental protection, it is appropriate to modify the uranium mining process so that the associate highly radioactive radium and thorium could be co-extracted together with natural uranium for subsequent transmutation in fast reactors. 

For breeder MSR versions, on-site continuous processing is typically proposed. In the early work of ORNL, as rapidly as the entire fuel salt on a 10-day cycle. In more recent proposed designs with harder spectrums, this time can be extended to several months and still allow breeding. With the thorium- U cycle, a factor that greatly complicates processing is that thorium behaves chemically very similar to the lanthanide fission products. 

The nuclear fuel cycle starts with the ore being extracted from the earth and follows it through processing and use until a final waste form is placed back for permanent disposal. Both uranium and thorium exist in nature as minerals that can be mined however, uranium will be used in this discussion because the use of thorium has not been extensively... 

The FUJI concept was proposed in connection with the philosophy of the thorium molten salt nuclear energy synergetic system (THORIMS-NES) [XXX-4 to XXX-6], explained in more detail in Section XXX-1.5. Different from the MSBR, the FUJI is a concept of a simplified molten salt reactor without continuous chemical processing and periodic core graphite replacement, aimed at attaining near-breeder characteristics in a Th-U closed fuel cycle. 

As far as reprocessing in the U/Pu fuel cycle is concerned, several chemical separation techniques have been proposed and developed in the past few decades. The most efficient process to date remains the PUREX process (Plutonium and Uranium Recovery by Extraction). This process uses nitric acid HNO3 and organic solvents to dissolve and extract selectively U and Pu, resulting in two separate product streams (U on one side and Pu on the other side of the process chain). As far as reprocessing in the Th/ U fuel cycle is concerned, THOREX (Thorium Oxide Recovery by Extraction) technology must be used, also based on dissolution in nitric acid and solvent extraction (however, with special care for the extraction of Pa, for the separa-tion of U and U, and for the dissolution of thorium dioxide in pure nitric acid). 

Assuming this once-used fuel can be safely processed (see below), the 6 Mt Pu/year would be processed into 5% enriched plutonium—thorium fuel to kick start the Pu —Th —cycle, which can be reused as fuel and burnt to about 40,000 MW(t) d/t using present fuel technology. 

Fust, the amount and type of information that states will have to provide to the IAEA is greatly expanded. In addition to the former requirement for data about nuclear fuel and fuel cycle activities, states will now have to provide an expanded declaration on a broad array of nuclear-related activities, such as nuclear fuel cycle-related research and development activities—not involving nuclear materials and the location, operational status and the estimated annual production of uranium mines and thorium concentration plants. (Thorium can be processed to produce fissile material, the key ingredient for nuclear weapons.) All trade in items on the NSG trigger list will have to be reported to the IAEA as well. 

Nuclear reactors can be designed on the basis of their fuel cycle such that they breed more fissile nuclides than what they use. Breeder reactors can utilize uranium, thorium, and plutonium resources more efficiently. There are two types of breeder reactors (1) fast neutron spectmm breeder and (2) thermal neutron spectmm breeder reactors, which are designed based on (99.2% natural abundance) and Th (100% natural abundance), respectively. Fertile nuclides and Th capture neutrons and trans-form, respectively, to fissile nuclides Pu and U. Through this process, which is known as breeding, the reactor produces more fissile nuclides than what it consumes. Fast-breeder reactors (FBRs) can also be used in order to transmute the long-lived... 

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