Enhanced Proliferation Resistance Research Articles

Defining the Challenges—Identifying the Key Poisoning Elements to Be Separated in a Future Integrated Molten Salt Fast Reactor Clean-Up System for iMAGINE

Nuclear fission technologies have the potential to play a significant role in the energy mix of a net-zero and sustainable society. However, to achieve the sustainability goal two significant challenges remain: efficient and sustainable fuel usage and the minimization of long-term nuclear waste. Civil nuclear molten salt systems and technologies offer the opportunity to address both, delivering future reactors at scale for efficient and effective power production and nuclear waste burnup. Potentially, both objectives could be fulfilled in one reactor system, which could significantly improve sustainability indices. The key to this innovation is demand driven development of a significantly reduced fuel cycle with enhanced proliferation resistance which offers further potential for improvement. To achieve these goals, a transformative approach for salt clean-up during molten salt reactor operation is proposed, by concentrating on the detection and removal of key neutron poisoning elements which prevent the reactor from long-term operation. To enable this highly innovative development work, a novel analysis of the evolving elementary fuel composition, their concentrations, and their criticality influence is now provided in this work. This, combined with consideration of the oxidation states of each of these elements then provides the basis for the selection of these key poisons and the development of advanced separation processes and process monitoring. This work also discusses the importance of the effective integration of physics and chemistry when systems modelling in achieving these system development goals.

Liquid Fuel Molten Salt Reactors for Thorium Utilization

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.

ICONE23-1659 AN INVESTIGATION INTO THE FEASIBILITY OF UTILIZING THORIUM-BASED FUEL CYCLES IN A SMALL FLUORIDE-SALT-COOLED HIGH-TEMPERATURE REACTOR

The objective of this study is to investigate the potential use of thorium-based fuels, ThO2-UO2 and ThO2-PuO2 in particular, in a small (125 MWth) fluoride-salt-cooled high-temperature reactor (FHR) with lithium-beryllium fluoride salt (Li2Be4F) coolant. This study investigates thorium as an alternative fuel because it offers several potential benefits including enhanced proliferation resistance, lower waste radiotoxicity, and a higher abundance in nature compared to uranium. The deterministic lattice physics code WIMS was used to find the k∞ values and reactivity (fuel Doppler and coolant temperature) coefficients and to carry out fuel depletion. The results were verified using the Monte Carlo code SERPENT. For the analysis of homogeneous ThO2-UO2 fuels, calculations are performed over a range of thorium volume fractions and uranium enrichments (up to 20 wt.%) that can achieve the same total burnup of 5.07 MWd/kg (with acceptable reactivity coefficients) as a 100% UO2 4.69 wt.% enriched fuel reference case. The maximum thorium volume fraction that can be utilised to obtain same burnup is 0.668 at 20 wt.%. In ThO2-PuO2 analysis, reactor grade Pu with an isotopic vector taken from typical light water reactor (LWR) discharge fuel with initial 4.5 wt.% enrichment, 50 MWd/kg burnup is used. The volumetric proportions of ThO2 and PuO2 are varied and the impact on total burnup, plutonium incineration rates and reactivity coefficients investigated. To achieve the same total burnup, the thorium volume fraction is estimated to be ∼0.86.

Linkage options between unit process to enhance proliferation resistance of pyroprocessing

Pyroprocessing technology has been actively developed at KAERI as one of the options to address the national spent fuel management issue. As much as the unit process development itself, the linkage method between each unit process is also of importance. There may be different linkage options between a set of two process units, or different options caused by different technologies for one of two unit processes connected to each other. The linkage method may affect the loss of nuclear materials and distribution of fission products, which are essential from the safeguards and proliferation resistance points of view. In this study, different technology options for some of the major process steps and possible linkage options were identified, and aspects of their proliferation resistance including the impact on the safeguards are discussed. Several options for decladding and feeding to electro-reduction, dross and the residual salt treatment of reduced materials as well as feeding to electro-refining, and salt transfer from electro-refining to electro-winning were identified and their technical and proliferation resistance aspects were analyzed. This study does not conclude any ranking of each option or suggest which option is the best. It is important to note that the objectives to review the process and linkage options are to give the basic characteristics and technical information related to safeguards measures and proliferation resistance and to suggest a conceptual idea for certain options to investigate further and enhance those aspects.

Study on FBR core concepts to increase proliferation resistance of plutonium in LWR–FBR transition period

The possibility to enhance proliferation resistance of discharged plutonium in fast breeder reactor (FBR) has been investigated in terms of reactor core-design aspects. The provisional target for proliferation resistance measures based on Saito's attractiveness (ATTR) is defined. It is found that a few percent of plutonium loading and/or Am/Cm loading, which come from various types of spent fuel, might satisfy the provisional target with a minimum impact on the core neutronic performances. On Am/Cm loading core, decay heat constraints for fuel-handling aspects are found to be important and should be considered in design. There is not significant change on the current developing scenarios for light water reactor–FBR transition period by applying the measures based on Saito's ATTR. It is found that applying Kimura's proposal, 15% 238Pu content requires about 7% MA loading fraction both in the core and the blanket, and it only can be applied at limited case. The period to consume minor actinides is shorter than to consume plutonium.

INPRO Studies of Proliferation Resistance for DUPIC Fuel Cycle

The evaluation of proliferation resistance on the DUPIC fuel cycle was performed using the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) user manual. The user manual was finally published from INPRO as a tool for proliferation resistance evaluation. Five user requirements were well organized under one basic principle, and each evaluation parameter for each user requirement has criteria for qualitative and quantitative evaluation for an innovative nuclear system. The DUPIC fuel cycle is to fabricate CANDU fuel from spent pressurized water reactor fuel by use of a dry thermal process without separating the stable fission products. The DUPIC process and fabricated fuel have very intense radiation background and a low amount of fissile plutonium and uranium. The DUPIC fuel cycle has a number of intrinsic features that enhance proliferation resistance. The number of assemblies in the DUPIC process to get 1 significant quantity is very large ([approximately]48 assemblies). The assessment results using the user manual show that the DUPIC fuel cycle is very strong against nuclear proliferation by the material property itself and the facility condition. Additionally, several suggestions and conditions were made to increase the proliferation resistance for innovative future nuclear energy system.

Feasibility study on shipping cask for JSFR fresh fuel contained minor actinide

Japan Atomic Energy Agency (JAEA) initiated “Fast Reactor Cycle Technology Development (FaCT)” project since 2005 to accomplish necessary R&Ds goals for Japan sodium-cooled fast reactor (JSFR) cycle (the loop type sodium-cooled fast reactor with MOX fuel and its related fuel cycle system). Since the concept of multi-transuranic (TRU) fuel recycling has been adopted to reduce environmental burden and to enhance proliferation resistance, decay heat and radiation of MA, etc. became major consideration in designing fresh fuel shipping cask. The use of dry cask and helium gas filling for its high thermal conductivity has been selected, since water as a coolant is not preferable in order to avoid corrosion of fuel steel by sodium–water reaction products. Hence the helium gas cask with five fuel subassemblies storage was designed. To assure the thermal integrity, the cladding temperature of the fuel assembly has to be kept below the creep temperature to prevent any creep damage before fuel loaded into the core. In this study, a whole cask analysis using three-dimensional FEM model has been carried out to confirm absence of creep damage on fuel cladding and thermal integrity complying with the IAEA safety standards. Furthermore, detailed fuel subassembly analysis using two-dimensional FEM model estimated the maximum cladding tube temperature in detail to confirm the heat transport performance. In addition, the cask filled with water is assessed by Monte Carlo method for its integrity at criticality.

Enhancement of proliferation resistance properties of commercial FBRs by material barriers

The possibility to enhance proliferation resistance of discharged plutonium in the blanket of fast breeder reactor (FBR) has been investigated in terms of reactor core design aspects. It can be achieved by increasing the fraction of even-mass-number plutonium isotopes in the discharged plutonium. Although it can be achieved by a radial blanket-free core, it also decreases the breeding ratio. By loading the blanket with plutonium (Pu) or minor actinides (MAs) or Pu-MA combination, it is possible to reduce the attractiveness of discharged plutonium. However, the loading material in the blanket should be kept in balance with achievable core performances, fuel fabrication, and fuel handling, accordingly and hence the criteria needed to comply with. It is found that applying the attractiveness or FOM criteria that using the combination of even-mass-number plutonium isotopes need more material to be loaded than the one needed when applying Pellaud’s or Kessler’s proposal that using only solely isotope fraction of plutonium.

A Conceptual Process Study for Recovery of Uranium Alone from Spent Nuclear Fuel by Using High-Alkaline Carbonate Media

This work studied a conceptual process to recover uranium alone from spent nuclear fuel using high-alkaline carbonate media with hydrogen peroxide for the purposes of reducing the volume of high-level active waste and recycling of uranium from the spent fuel with greatly enhanced proliferation resistance, environmental friendliness, and operational safety. The transuranium (TRU) elements were evaluated to be undissolved and precipitated together with other fission products during the oxidative leaching of uranium from the spent fuel. The leaching ratio of uranium dioxide to TRU dioxide from spent fuel in the carbonate solution with H2O2 was estimated to be more than about 108. Only Cs, Tc, Mo, and Te among the major fission products in the spent fuel were dissolved together in the carbonate solution. In the carbonate solution with H2O2, UO2 was dissolved in the form of uranyl peroxo-carbonato complex ions, which could be recovered in the form of uranium peroxide precipitate with a very low solubility by acidification of the solution in a succeeding step. All the inorganic salts of Na2CO3, NaOH, and HNO3 used in the process suggested could be almost completely recovered and recycled into the process again without any generation of secondary wastes.

Opportunities, challenges and strategies for innovative SMRs

The designers of innovative Small- and Medium-sized Reactors (SMRs) are pursuing new design and deployment strategies, making use of certain advantages provided by a smaller reactor size and capacity to achieve reduced design complexity and simplified operation and maintenance requirements and to explore the advantages of an incremental capacity increase through plant clustering or a multimodule approach to plant design. The competitiveness of SMRs depends on the incorporated strategies to overcome the loss of economies of scale, but equally it depends on finding the appropriate market niches for such reactors. For many less developed countries, these are the features of enhanced proliferation resistance and the increased robustness of the barriers for sabotage protection that may ensure the progress of nuclear power. Based on the outputs of the ongoing International Atomic Energy Agency (IAEA) activities for innovative SMRs, the paper provides a summary of the state-of-the-art approaches to improve SMR competitiveness and incorporate enhanced proliferation resistance and energy security.

Options for small and medium sized reactors (SMRs) to overcome loss of economies of scale and incorporate increased proliferation resistance and energy security

The designers of innovative small and medium sized reactors pursue new design and deployment strategies making use of certain advantages provided by smaller reactor size and capacity to achieve reduced design complexity and simplified operation and maintenance requirements, and to provide for incremental capacity increase through multi-module plant clustering. Competitiveness of SMRs depends on the incorporated strategies to overcome loss of economies of scale but equally it depends on finding appropriate market niches for such reactors. For many less developed countries, these are the features of enhanced proliferation resistance and increased robustness of barriers for sabotage protection that may ensure the progress of nuclear power. For such countries, small reactors without on-site refuelling, designed for infrequent replacement of well-contained fuel cassette(s) in a manner that impedes clandestine diversion of nuclear fuel material, may provide a solution. Based on the outputs of recent IAEA activities for innovative SMRs, the paper provides a summary of the state-of-the-art in approaches to improve SMR competitiveness and incorporate enhanced proliferation resistance and energy security.

Application of Fusion Neutron Source for Denaturing of Plutonium

Potential of DT fusion neutron source to enhance proliferation resistance properties of plutonium by means of its isotopic denaturing is addressed. The approach is exemplified by denaturing of pure 239Pu and plutonium of typical LWR spent fuel through transmutation of neptunium. The essential feature of a fusion driven system proposed in the study is a zero mass balance of plutonium: total plutonium inventory is constant during irradiation. The system is capable to convert pure 239Pu into plutonium composition with more than 20% fraction of key 238Pu isotope during 1,000 d of irradiation under initial neutron loading of 1 MW.m−2. Denaturing of LWR spent fuel plutonium under the same conditions would increase its 238Pu content up to 10-12%.

ICONE15-10617 COMMERICALIZATION OF THE GLOBAL NUCLEAR ENERGY PARTNERSHIP (GNEP)

In February 2006 President Bush announced the Advanced Energy Initiative, which included the Department of Energy's (DOE) Global Nuclear Energy Partnership (GNEP). GNEP has seven broad goals, one of the major elements being to develop and deploy advanced nuclear fuel recycling technology. DOE is contemplating accelerating the deployment of these technologies to achieve the construction of a commercial scale application of these technologies. DOE now defines this approach as "…two simultaneous tracks: (1) deployment of commercial scale facilities for which advanced technologies are available now or in the near future, and (2) further research and development of transmutation fuels technologies." GE believes an integrated technical solution, using existing reactor and fuel reprocessing technologies, is achievable in the near term to accelerate the commercial demonstration of GNEP infrastructure. The concept involves a single, integrated, commercial scale, recycling facility consisting of the Consolidated Fuel Treatment Center (CFTC), capable of processing LWR and fast reactor Spent Nuclear Fuel (SNF) and fabricating Advanced Recycling Reactor (ARR) actinide fuel. The integrated facility would include a fast reactor that uses actinide-bearing fuel to produce electricity. For optimal performance, GE believes this integrated facility should be co-located to eliminate transportation between the CFTC and ARR, and enhance proliferation resistance. This Advanced Recycling Center takes advantage of previous investments by government and industry in fast reactor technology research and development. To allow for commercial acceptance, a prototypical demonstration reactor and associated fuel cycle facility will be constructed, tested, and licensed. Taking advantage of GE's NRC-reviewed modular sodium-cooled PRISM reactor, only a single reactor will be needed and the cost and risk minimized in the initial phase of the program. This paper outlines a process and a schedule to deploy the necessary domestic infrastructure to achieve the goals of GNEP. It is recognized that there are many types of public private partnerships that could demonstrate the technologies to close the fuel cycle. It is important to acknowledge the realities in the marketplace when developing an approach to advance nuclear technologies that can gain widespread commercial acceptance.

4598308 Easily repairable, low cost, high speed electromechanical assembly of integrated circuit die

Japan Atomic Energy Agency (JAEA) initiated “Fast Reactor Cycle Technology Development (FaCT)” project since 2005 to accomplish necessary R&Ds goals for Japan sodium-cooled fast reactor (JSFR) cycle (the loop type sodium-cooled fast reactor with MOX fuel and its related fuel cycle system). Since the concept of multi-transuranic (TRU) fuel recycling has been adopted to reduce environmental burden and to enhance proliferation resistance, decay heat and radiation of MA, etc. became major consideration in designing fresh fuel shipping cask.The use of dry cask and helium gas filling for its high thermal conductivity has been selected, since water as a coolant is not preferable in order to avoid corrosion of fuel steel by sodium–water reaction products. Hence the helium gas cask with five fuel subassemblies storage was designed. To assure the thermal integrity, the cladding temperature of the fuel assembly has to be kept below the creep temperature to prevent any creep damage before fuel loaded into the core. In this study, a whole cask analysis using three-dimensional FEM model has been carried out to confirm absence of creep damage on fuel cladding and thermal integrity complying with the IAEA safety standards. Furthermore, detailed fuel subassembly analysis using two-dimensional FEM model estimated the maximum cladding tube temperature in detail to confirm the heat transport performance. In addition, the cask filled with water is assessed by Monte Carlo method for its integrity at criticality.