GFR equilibrium cycle analysis with the EQL3D procedure

Advanced fast reactors of the fourth generation should be capable to breed their own fuel from ²³⁸U feed and to recycle the actinides from their own spent fuel. This recycling or virtually the closure of fuel cycle can converge to an equilibrium fuel cycle and has impact on the safety-related parameters. The goals of this study are: (i) to apply an equilibrium cycle procedure EQL3D to the Gas cooled Fast Reactor (GFR), (ii) to simulate and confirm the GFR neutronics capability for closed fuel cycle, and (iii) to evaluate the safety-related parameters of the equilibrium cycle.Equilibrium cycle method for considering the homogeneous recycling of actinides is a known approach. However, in EQL3D the equilibrium method is newly applied for hexagonal-z 3D core geometry and 33 energy-groups neutron-flux calculation. This geometry enables to characterize the equilibrium cycle for complex reloading patterns within a multi-batch cycle.Two GFR geometries were studied, the first based on an international neutronics benchmark with a simple set-up and the second based on more advanced core design. For the advanced design, three reloading patterns within a multi-batch cycle with four different feeds were compared.The GFR neutronics capability for closed cycle was proved. The negative impact of the fuel cycle closure on safety-related parameters was confirmed and quantified. The GFR core with closed fuel cycle could serve after prospective optimization as a sustainable and clean energy source.     

Fuel performance evaluation of rock-like oxide fuels

The concept of the rock-like oxide (ROX) fuel has been developed for the annihilation of excess plutonium in light water reactors. Irradiation tests and post-irradiation examinations were carried out on candidate ROX fuels. The ternary fuel of YSZ–spinel–corundum system, the single-phase fuels of YSZ, the particle-dispersed fuels of YSZ in spinel or corundum matrix, and the blended fuels of YSZ and spinel or corundum matrix were fabricated and submitted to irradiation testings. The fuels containing spinel showed chemical instabilities with the vaporization of MgO component, which caused fuel restructuring. The swelling behavior was improved with the particle-dispersed fuels. However, the particle-dispersed fuels showed higher fractional gas release (FGR) than blended type fuels. The FGR of YSZ single-phase fuels were comparable to what would be expected for UO₂ fuel at the similar fuel temperatures. The YSZ single-phase fuel showed the best irradiation performance among the ROX fuels investigated.     

Metallic fuels for advanced reactors

In the framework of the Generation IV Sodium Fast Reactor Program, the Advanced Fuel Project has conducted an evaluation of the available fuel systems supporting future sodium cooled fast reactors. This paper presents an evaluation of metallic alloy fuels. Early US fast reactor developers originally favored metal alloy fuel due to its high fissile density and compatibility with sodium. The goal of fast reactor fuel development programs is to develop and qualify a nuclear fuel system that performs all of the functions of a conventional fast spectrum nuclear fuel while destroying recycled actinides. This will provide a mechanism for closure of the nuclear fuel cycle. Metal fuels are candidates for this application, based on documented performance of metallic fast reactor fuels and the early results of tests currently being conducted in US and international transmutation fuel development programs.     

Hertzian indentation of advanced lmfbr fuels with simulated burn-up

Hertzian indentation fracture of advanced fast breeder reactor fuel materials [mixed carbonitrides, (U_0.8, Pu_0.2)C_0.8N_0.2, and nitrides (U_0.8Pu_0.2)N was evaluated to yield the fracture surface energy, γ, and the fracture toughness, K_Ic. Both technological grade fuels and fuels with added fission products to chemically simulate burn-up values of 3 and 10 at.% were used. As in previous self-diffusion studies on the same materials, identical behavior (identical critical loads, P_c for crack formation) was observed for 3 and 10% b.u. Simulated M(C, N), whereas the 10% b.u. Simulated MN showed a cracking behavior identical with that of the undoped MN. In contrast, the 3 at.% b.u. Simulated MN showed lower P_c values. This is compatible with differences in fission product solubilities in these materials. The effect of fission products on γ was < 20% whereas γ increased from (U, Pu)(C, N) to (U, Pu)N by up to 80 to 90%, depending on content in fission products.     

Fuel cycle analysis of TRU or MA burner fast reactors with variable conversion ratio using a new algorithm at equilibrium

Partitioning and Transmutation (P&T) strategies assessment and implementation play a key role in the definition of advanced fuel cycles, in order to insure both sustainability and waste minimization. Several options are under study worldwide, and their impact on core design and associated fuel cycles are under investigation, to offer a rationale to down selection and to streamline efforts and resources. Interconnected issues like fuel type, minor actinide content, conversion ratio values, etc. need to be understood and their impact quantified. Then, from a practical point of view, studies related to advanced fuel cycles require a considerable amount of analysis to assess performances both of the reactor cores and of the associated fuel cycles. A physics analysis should provide a sound understanding of major trends and features, in order to provide guidelines for more detailed studies. In this paper, it is presented an improved version of a generalization of the Bateman equation that allows performing analysis at equilibrium for a large number of systems. It is shown that the method reproduces very well the results obtained with full depletion calculations. The method is applied to explore the specific issue of the features of the fuel cycle parameters related to fast reactors with different fuel types, different conversion ratios (CR) and different ratios of Pu over minor actinide (Pu/MA) in the fuel feed. As an example of the potential impact of such analysis, it is shown that for cores with CR below 0.8, the increase of neutron doses and decay heat can represent a significant drawback to implement the corresponding reactors and associated fuel cycles.     

Fuel cycle transients under power mismatch

The transient behaviour of a single stream fueling system is examined using two simple representations of power mismatch between batches for a reloading scenario in which the cycle time varies. It is found that in one case the convergence rate of the cycle sequence slows with increasing mismatch, while the other, more realistic, model yields faster convergence with greater mismatch.     

Purex co-processing of spent LWR fuels: comparative fuel cycle cost analyses

Nuclear fuel cycle costs for purex reprocessing and co-processing cycles are to be evaluated, by calculating unit costs of recovered, accordingly treated and fabricated products and then comparing those to the unit cost of fresh uranium fuel ready to be loaded into a typical LWR on the once-through cycle.     

Fuel cycle strategies and plutonium management in Europe

An overview of current nuclear power generation and fuel cycle strategies in Europe is presented, with an emphasis on options for the management of separated plutonium in the medium to long term. Countries which have opted for reprocessing of spent fuel have had to contend with increasing inventories of separated plutonium. Of the various potential options for utilisation or disposition of these stockpiles, only light water reactor (LWR) mixed-oxide (MOX) fuel programmes are sufficiently technologically mature to be fully operational in several European countries at present. Such reprocessing-recycling programmes allow for a stabilisation of the overall separated plutonium stocks, but not for a significant reduction in the stockpile. Moreover, the quality of recycled plutonium decreases at each potential step of re-irradiation. Therefore, optimised or new ways of managing the plutonium stocks in the medium to long term are required. In the present overview we consider the most promising options for reactor utilisation of plutonium in both near-term future reactor and Generation IV systems.     

VVER-1000应用PC级燃料的燃料管理

核电厂应用后处理制造的燃料组件,便可完成核燃料的"闭式循环",可以增加燃料的利用率和缓解乏燃料的储存难题。PC级燃料是利用后处理产品铀进行浓缩及加工所获得的一种核燃料,俄罗斯的VVER-1000机组中使用PC级燃料已有20年的历史。田湾核电厂1号机组已确定从第10燃料循环开始使用PC级燃料。使用KASKAD程序包,对VVER-1000使用PC级浓缩铀制造的TVS-2M组件展开研究设计,分析其应用的可行性,给出优化的燃料管理方案。     

Dissolution and analysis of erbium doped cermet fuels

Cermets are suggested as new kind of nuclear fuel to reduce global costs. They need high enriched fuel and thus use of burnable poison. Special pellets were developed and irradiated to test such concepts. Some pellets consist of a cermet fuel. With an improved fuel thermal conductivity (by using metal matrix), lower temperatures than standard fuel are obtained. Some pellets were made of cermet and erbium in small quantity. Studies on erbium were launched to determine the influence of this neutron poison. Standard dissolutions (HNO₃, HF) on cermet (Mo-UO₂) and on erbium doped cermet show a large amount of insoluble matter. Tests have been carried out in order to establish a procedure for a complete dissolution of active pellets. Consequently, an optimal process was defined. Irradiated pellets from experimental reactor SILOE will be dissolved. Analytical chemistry studies were undertaken. Thermal Ionization Mass Spectrometry (TIMS) and Glow Discharge Mass Spectrometry (GDMS) have been applied. The U and Er isotopic composition has been determined on different samples.     

Spectrum shifting as a mechanism to improve performance of VHTRs with advanced actinide fuels

Reprocessing of spent LWR fuel is an intrinsic part of the closed fuel cycle. While current technologies treat recovered minor actinides as high level wastes, the primary objective of one of the U.S. DOE Nuclear Energy Research Initiative (NERI) projects is to assess the possibility, advantages and limitations of designing a single-batch (no-refueling) very high temperature reactor (VHTR) configuration that utilizes transuranic nuclides (TRU) as a fuel component. Since both VHTR core design concepts, pebble bed and prismatic block assembly, permit flexibility in component configuration, fuel utilization and management, it is possible to improve fissile properties by neutron spectrum shifting through configuration adjustments. The presented analysis is focused on the TRU-impact on the single-batch mode (no-refueling) VHTR core lifetime. As a result of the analysis, promising performance characteristics have been demonstrated. The TRU-core configurations are expected to be suitable for long-term autonomous operation without intermediate refueling.     

Characterization of radiation fields and dose assessment from fuels manufacturing for advanced fuel cycles

The purpose of this study was to examine radiological hazards introduced to workers from the fabrication of fuels with minor actinides (MA) and reprocessed uranium (RepU) and determine the feasibility of using shielded gloveboxes instead of remotely controlled operations in hot cells. Of particular concern is the increase in photon source term from the daughter products in RepU and the mixed neutron and photon fields introduced by MA recycle. In the interest of keeping the glovebox worker's radiation dose as low as reasonable achievable (ALARA), dose rates were calculated for various typical and bounding fuel compositions for light water reactors (LWR) and fast reactor (FR) mixed oxide (MOX) fuels with and without MA. The impact of varying the separation efficiency of americium (Am) and curium (Cm) was examined because current separation processes in reprocessing and recycling do not allow for the complete separation of Am from Cm. The additional Cm will cause a significant increase in the neutron source term. The sensitivity of the fuels to aging time was also examined by decaying the recycled feedstocks from 6 months to 3 years to simulate the effect of delays in reuse of recycled materials. The highest photon and neutron sources were used to calculate the additional shielding requirements that would be needed for fuel fabrication. Through insight gained from this study, it can be concluded that a standard glovebox with one quarter inch stainless steel walls can be used to fabricate fuels with Am with little to no additional shielding. The introduction of small amounts of Cm in the fuels will require the fuel fabrication to be preformed remotely in hot cells. Thus, stressing the importance of developing methods to increase the separation efficiency of Am from Cm.     

Combustion synthesis of γ-lithium aluminate by using various fuels

Combustion synthesis, which is a quick and straightforward preparation process to produce homogeneous, crystalline and unagglomerated multicomponent oxide ceramic powders without the intermediate decomposition and/or calcining steps, was used to prepare γ-lithium aluminate. Lithium nitrate and aluminium nitrate were used as the starting materials and various organic compounds, such as citric acid, urea, carbohydrazide, glycine and alanine, as the fuels. The mixture of nitrate and fuel could be ignited at 450 °C, but only urea and carbohydrazide could be reacted with the mixed nitrates to result in dry, loose and white γ-LiAlO₂ powders. In this study, the effects of fuel type and ratio of fuel to nitrates on the phase formation of γ-LiAlO₂ powder were investigated and also discussed. Additionally, the phase and morphology of the γ-LiAlO₂ powder synthesized by the combustion reaction were compared with that by the conventional solid state reaction.     

Analysis of the reactivity coefficients of the advanced high-temperature reactor for plutonium and uranium fuels

The conceptual design of the advanced high-temperature reactor (AHTR) has recently been proposed by the Oak Ridge National Laboratory, with the intention to provide and alternative energy source for very high temperature applications. In the present study, we focused on the analyses of the reactivity coefficients of the AHTR core fueled with two types of fuel: enriched uranium and plutonium from the reprocessing of light water reactors irradiated fuel. More precisely, we investigated the influence of the outer graphite reflectors on the multiplication factor of the core, the fuel and moderator temperature reactivity coefficients and the void reactivity coefficient for five different molten salts: NaF, BeF₂, LiF, ZrF₄ and Li₂BeF₄ eutectic. In order to better illustrate the behavior of the previous parameters for different core configurations, we evaluated the moderating ratio of the molten salts and the absorption rate of the key fuel nuclides, which, of course, are driven by the neutron spectrum. The results show that the fuel and moderator temperature reactivity coefficients are always negative, whereas the void reactivity coefficient can be set negative provided that the fuel to moderator ratio is optimized (the core is undermoderated) and the moderating ratio of the coolant is large.     

Heterogeneous fuels for minor actinides transmutation: Fuel performance codes predictions in the EFIT case study

Plutonium recycling in new-generation fast reactors coupled with minor actinides (MA) transmutation in dedicated nuclear systems could achieve a decrease of nuclear waste long-term radiotoxicity by two orders of magnitude in comparison with current once-through strategy.     

A parametric analysis of heat transfer in composite nuclear fuels

The composite nuclear fuel described in this paper consists of a heat generating fuel matrix containing cylindrical metal and fibers uniformly aligned throughout the matrix. Exact analytical solutions were found for temperature distributions in the fiber and matrix for a composite cell modeled as concentric cylinders. A parametric study is presented of composite overall thermal conductivities and temperature distributions as a function of fiber-to-matrix conductivity ratios, cell length-to-radius ratios, and fiber-to-cell radius ratios. For composite cells in which length-to-radius ratios exceed 10, axial temperature distributions may be calculated assuming a homogeneous material.