The thorium molten salt reactor: Moving on from the MSBR
L. MathieuD. HeuerR. BrissotC. GarzenneChristian Le BrunDavid LecarpentierE. LiatardJ.M. LoiseauxO. MéplanE. MerleA. NuttinE. van WalleJ. N. Wilson
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Molten salt reactor
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Topics discussed concerning the molten salt breeder reactor program include: resource utilization; historical development of molten salt reactors; molten salt breeder reactor concept description; status of molten salt breeder reactor technology; and industrial participation in the molten salt breeder reactor program. (17 references) (DCC)
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Molten-Salt Reactors (MSRs) use molten-salt fluid-fuel, which contains thorium (Th) and uranium-233 (233U) as fertile and fissile, respectively. Also, the MSRs can accept plutonium (Pu) as fissile. Besides the above benefit on nuclear fuel cycle, the MSR has other many advantages such as the power-size flexibility, no necessity of refueling, effective incineration of minor actinides (MA), excellent safety, and good economy. The cycle produces very small amount of MA compared with the conventional U fuel cycle. Also, 233U accompanies high gamma-ray emissions, and this fact contributes to the resistance to nuclear proliferation issues. The MSRs has been studied since early 1950's in Oak Ridge National Laboratory (ORNL), and the Molten-Salt test Reactor Experiment (MSRE) has shown excellent operation in 1965-69. The conceptual design study was issued for Molten-Salt Breeder Reactor (MSBR) as large as 1GWe plant. In Japan, small MSRs of 150MWe to 200MWe(FUJI-, FUJI-12 and FUJI-U3) have been proposed. FUJI-12 is a simple one-region core, and it attained a high conversion ratio of 0.92 by the batch chemical processing at every 7.5 years. However, FUJI-12 has to replace the graphite moderator at every 15 years, and this may cause additional maintenance work and cost. Then, an improved small MSR (FUJI-U3) has been proposed. The improvement is to introduce a 3-region core design concept in order to eliminate graphite replacement by reducing the maximum neutron flux. It is concluded that there is no need to replace graphite moderator for 30 years operation of FUJI-U3. In this paper, we investigated the other possibility of a two region core for the simplicity. Using one energy group neutron diffusion theory and assuming bare reactor, the optimum selection of region wise neutron multiplication factors can be theoretically and easily obtained. The validity of the selection was also proved by a numerical calculation. In MSRs, there is no burnup distribution of the fuel. The region wise neutron multiplications can be obtained by adjusting the volume fraction of fuel in a cell with a given composition of fuel salt. Thus, it is easy to apply the theoretical result directly in selecting the boundary of the region and the multiplication factor in each region and also to maintain the characteristics during the core burnup. The optimization of the actual core configuration was searched by a nuclear analysis code SRAC95 with the nuclear data library of JENDL3.2. In such case the region wise multiplication factor can be adjusted to compensate the neutron loss by leakage in order to attain criticality. Taking the theoretical result as the initial condition, the optimization was quite easy to accomplish. It is concluded that there is no need to replace graphite moderator for 33 years operation of a two-region core design concept.
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Molten salt reactors (MSRs) represent a class of reactors that use liquid salt, usually fluoride based or chloride based, as either a coolant with a solid fuel (such as fluoride salt–cooled high-temperature reactors) or as a combined coolant and fuel with the fuel dissolved in a carrier salt. For liquid-fueled MSRs, the salt can be processed online or in a batch mode to allow for removal of fission products as well as for introduction of fissile fuel and fertile materials during reactor operation. The MSR is most commonly associated with the 233U/thorium fuel cycle, as the nuclear properties of 233U combined with the online removal of parasitic absorbers enable the design of a thermal-spectrum breeder reactor. However, MSR concepts have been developed using all neutron energy spectra (thermal, intermediate, fast, and mixed-spectrum zoned concepts) and with a variety of fuels including uranium, thorium, plutonium, and minor actinides. Early MSR work was supported by a significant research and development (R&D) program that resulted in two experimental systems operating at Oak Ridge National Laboratory in the 1950s and 1960s: the Aircraft Reactor Experiment and the Molten Salt Reactor Experiment. Subsequent design studies in the 1970s focusing on thermal-spectrum thorium-fueled systems established reference concepts for two major design variants: (1) a molten salt breeder reactor (MSBR) with multiple configurations that could breed additional fissile material or maintain self-sustaining operation and (2) a denatured molten salt reactor (DMSR) with enhanced proliferation resistance. MSRs have been selected as one of the Generation IV systems, and development activity has been seen in fast-spectrum MSRs, waste-burning MSRs, and MSRs fueled with low-enriched uranium as well as in more traditional thorium fuel cycle–based MSRs. This paper provides a historical background of MSR R&D efforts, surveys and summarizes many of the recent developments, and provides analysis comparing thorium-based MSRs by way of example.
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Molten salt reactor (MSR), the only one using liquid fuel in the six types “Generation IV” reactors, is very different from reactors in operation now and has initiated very extensive interests all over the world. This paper is primarily aimed at investigating the breeding characteristics of high-power thorium molten salt reactor (TMSR) based on the two-fluid molten salt breeder reactor (MSBR) with a superior breeding performance. We explored the optimized structure to be a thorium-based molten salt breeder reactor with different core conditions and different postprocessing programs, and finally got the breeding ratio of 1.065 in our TMSR model. At last we analyzed the transient security of our optimized model with results show that the temperature coefficient of core is −3 pcm/K and a 2000 pcm reactivity insertion can be successfully absorbed by the core if the insertion time is more than or equal to 5 s and the core behaves safely.
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One of the pending questions concerning Molten Salt Reactors based on the {sup 232}Th/{sup 233}U fuel cycle is the supply of the fissile matter, and as a consequence the deployment possibilities of a fleet of Molten Salt Reactors, since {sup 233}U does not exist on earth and is not yet produced in the current operating reactors. A solution may consist in producing {sup 233}U in special devices containing Thorium, in Pressurized Water or Fast Neutrons Reactors. Two alternatives to produce {sup 233}U are examined here: directly in standard Molten Salt Reactors started with Plutonium as fissile matter and then operated in the Th/{sup 233}U cycle; or in dedicated Molten Salt Reactors started and fed with Plutonium as fissile matter and Thorium as fertile matter. The idea is to design a critical reactor able to burn the Plutonium and the minor actinides presently produced in PWRs, and consequently to convert this Plutonium into {sup 233}U. A particular reactor configuration is used, called 'unique channel' configuration in which there is no moderator in the core, leading to a quasi fast neutron spectrum, allowing Plutonium to be used as fissile matter. The conversion capacities of such Molten Salt Reactors are excellent. For Moltenmore » Salt Reactors only started with Plutonium, the assets of the Thorium fuel cycle turn out to be quickly recovered and the reactor's characteristics turn out to be equivalent to Molten Salt Reactors operated with {sup 233}U only. Using a combination of Molten Salt Reactors started or operated with Plutonium and of Molten Salt Reactors started with {sup 233}U, the deployment capabilities of these reactors fully satisfy the condition of sustainability. (authors)« less
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The Molten Salt Reactor (MSR), the only one using liquid fuel in the six types ‘Generation IV’ reactors, is very different from reactors in operation now and has initiated very extensive interests all over the world. This paper is primarily aimed at investigating the breeding characteristics of high power (1000 MWe) Thorium Molten Salt Reactor (TMSR) based on the two-fluid Molten Salt Breeder Reactor (MSBR) with superior breeding performance. We explored the optimized structure to be a thorium based molten salt breeder reactor with different core conditions and different post-processing programs, and finally got the breeding ratio of 1.065 in our TMSR model. At last we analyzed the transient security of our optimized model with results show that the temperature coefficient of core is −3 pcm/K and a 2000 pcm reactivity insertion can be successfully absorbed by the core if the insertion time is more than or equal to 5 seconds and the core behaves safely.
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Molten Salt Reactors based on the thorium cycle were studied in the 1950 to 1960s to lead to the Molten Salt Breeder Reactor concept, which was finally discontinued prior to any industrial development. In the past few years, this concept has once again been studied in order to generalize it and seek configurations ensuring a high intrinsic safety level, an initial inventory compatible with intensive deployment on a worldwide scale, and a not-too-demanding salt chemical reprocessing scheme.The Thorium Molten Salt Reactor (TMSR) thus defined is studied in the Th-233U cycle in various configurations obtained by modulating the amount of graphite in core to obtain a thermal, an epithermal, or a fast spectrum. In particular, configurations of a fast spectrum TMSR have been identified with outstanding safety characteristics and minimal fuel-reprocessing requirements.
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