Utilization of rock-like oxide fuel in the phase-out scenario

Utilization of rock-like oxide (ROX) fuel in light water reactors for plutonium (Pu) burning was studied by nuclear material balance (NMB) analysis for a case of Japanese phase-out scenario under investigation after the Fukushima accident. For the analysis, the NMB code was developed with features of accurate burn-up calculation, flexible combination of reactors and fuels, and an ability to estimate waste and repository. Three scenario groups of once-through Pu burning in mixed oxide (MOX) fuel and in ROX fuel were analyzed. Using two full-MOX or full-ROX reactors the Pu amount is reduced to about one-half and the isotopic vector of Pu deteriorated for being used as a nuclear weapon, especially in terms of spontaneous fission neutron generation. Effects of ROX reactors are more significant than MOX reactors in terms of both reduction in the Pu amount and deterioration of the isotopic vector. Repository footprint and potential radiotoxicity are not reduced by the MOX and ROX reactors because the heat and toxicity of MOX and ROX spent fuels are considerably high.     

Core concept of compound process fuel cycle

This paper presents fast reactor core concept and its feasibility as a part of newly proposed compound process fuel cycle in which spent fuels of light water reactor are multi-recycled without conventional reprocessing but with only pyrochemical processing, fuel re-fabrication and reloading to the fast reactor core. Results of the core survey analyses in order to find out the feasibility of this concept, taking example for BWR MOX spent fuel of 60 GWd/t burn-up, show that four times recycling of LWR spent fuel with the burn-up of more than 300 GWd/t can be achieved without increasing MA content. Such benefits will be expected in this concept as reduction of fuel cycle cost due to simplified reprocessing procedure, reduction of environmental impacts due to reduced high level waste, efficient utilization of nuclear fuel resources, enhancement of nuclear non-proliferation, and suppression of LWR spent fuel pile-up.     

MOX Fuel Performance Experiment Specified for Japanese PWR Utilisation in the HBWR

In order to obtain high burn-up MOX fuel irradiation performance data, SBR and MIMAS MOX fuel rods with Pufissile enrichment of about 6 wt% have been irradiated in the HBWR. In-pile performance data of MOX have been obtained, and the peak burn-up of MOX pellet have reached to 66 GWd/tM as of October 2004. MOX fuel temperature is confirmed to have no significant difference compared to UO₂, if taking into account adequately for thermal conductivity degradation due to PuO₂ addition and burn-up development, and measured fuel temperature agrees well with HB-FINE code calculation up to high burn-up region. Fission gas release of MOX is possibly larger than UO₂ based on temperature and pressure assessment. No significant difference is confirmed between SBR and MIMAS MOX on FGR behaviour. MOX fuel swelling rate agrees well with solid swelling rate. Cladding elongation data shows onset of PCMI in high power region. Ramp test data from other experiment programs with various types of MOX fabrication route confirms superior PCI resistance of MOX compared to UO₂, due to enhanced creep rate of MOX. The irradiation is expected to continue until achieving of 70 GWd/tM (MOX pellet peak).     

Neutronic performance comparation of MOX, nitride and metallic fuel based 25–100 MWe Pb–Bi cooled long life fast reactors without on-site refuelling

In this paper the performance of 25–100 MWe Pb–Bi cooled long life fast reactors based on three type of fuels: MOX, Nitride and Metal are compared and discussed. In general MOX fuel (UO₂–PuO₂) has lower atomic density compared to the nitride or metal fuel, but MOX fuel has some advantages such as higher Doppler coefficient, high melting point and availability. Nitride fuel has advantages such as high density, high thermal conductivity, and high melting point, but need N-15 to avoid C-14 problems.

The results show that nitride fuel as well as MOX fuel can be used to develop 25–100 MWe (75–300 MWth) Pb–Bi cooled long life reactors without on-site fuelling. The results show that nitride fuels have more superior neutronic characteristics compared to that of MOX fuel due to higher density. However, in the large power level both fuels can be easily applied. In lower power level the MOX fuel need higher fuel volume fraction to reach the comparable target of nitride fuelled cores.     

High burn-up operation and MOX burning in LWR; Effects of burn-up and extended cooling period of spent fuel on vitrification and disposal

Looking ahead to final disposal of high-level radioactive waste arising from further utilization of nuclear energy, the effects of high burn-up of light-water reactors (LWR) with UO₂ and MOX fuel and extended cooling period of spent fuel on waste management and disposal were discussed. It was assumed that the waste loading of waste glass is restricted by three factors: heat generation rate, MoO₃ content, and platinum group metal content. As a result of evaluation for effects of extended cooling period, the waste loading of waste glass from both UO₂ and MOX spent fuel could be increased in the current vitrification technology. For the storage of waste glass from MOX spent fuel with higher waste loading, however, those waste glass require long storage period prior to geological disposal because decay heat of ²⁴¹Am contributes significantly. Therefore, the evaluation of effects of Am separation on the storage period was performed. Furthermore, heat transfer calculation was carried out in order to evaluate the temperature of buffer material in a geological repository. The results showed, 70 to 90% of Am separation is sufficiently effective in terms of thermal feasibility of a repository.     

LWR spent fuel transmutation with fusion-fission hybrid reactors

In this paper the transmutation of light water reactors (LWR) spent fuel is analyzed. The system used for this study is the fusion-fission transmutation system (FFTS). It uses a high energy neutron source produced with deuterium-tritium fusion reactions, located in the center of the system, which is surrounded by a fission region composed of nuclear fuel where the fissions take place. In this study, the fuel of the fission region is obtained from the recycling of LWR spent fuel. The MCNPX Monte Carlo code was used to setup a model of the FFTS. Two fuel types were analyzed for the fissile region: the mixed oxide fuel (MOX), and the inert matrix fuel (IMF). Results show that in the case of the MOX fuel, an important Pu-239 breeding is achieved, which can be interesting from the point of view of maximal uranium utilization. On the contrary, in the case of the IMF fuel, high consumption of Pu-239 and Pu-241 is observed, which can be interesting from the point of view of non-proliferation issues. A combination of MOX and IMF fuels was also studied, which shows that the equilibrium of actinides production and consumption can be achieved. These results demonstrate the versatility of the fusion-fission hybrid systems for the transmutation of LWR spent fuel.     

Thermal conductivity of homogeneous and heterogeneous MOX fuel with up to 44 MWd/kgHM burn-up

New thermal diffusivity data for homogeneous SBR and heterogeneous MIMAS and OCOM MOX fuels are reported. No significant difference between the thermal diffusivity of the homogeneous and heterogeneous fuels was found at the burn-up up to 44 MWd/kgHM. These measurements, combined with previously published results or correlation functions for irradiated UO₂ and MOX were compared and it was found that separate correlations for these two fuels are not justified. A correlation for the thermal conductivity of irradiated UO₂ and MOX as a function of burn-up and irradiation temperature is proposed.     

MOX燃料在轻水堆核电站中的应用

目前MOX燃料已成为一种可用于轻水堆核电站成熟的核燃料。简要介绍了国外该领域的发展状况以及MOX燃料对反应堆性能的主要影响和应对措施。探讨了MOX燃料在国内压水堆核电站中的应用问题。     

Preparation and characterization of the simulated burnup americium-containing uranium–plutonium mixed oxide fuel

In order to investigate the effect on fuel thermophysical properties when adding americium and selected fission products to uranium–plutonium mixed oxide (MOX) fuel, simulated low decontamination MOX fuel with high burn-ups to 250 GWd/t, has been prepared and subjected to characterization tests, elastic moduli measurements and melting temperature measurement. Elastic moduli for the simulated low decontamination MOX fuel were almost the same level as fuel without americium and fission products and decrease in the moduli was slight with increasing simulated burn-up. The melting temperature of high burn-up, low decontamination MOX fuel may be estimated by using the findings on the effect of americium, plutonium addition and fission products accumulation.     

裂变核全套中子数据评价

裂变核全套中子评价数据对反应堆设计和安全运行、乏燃料少锕系核素嬗变率、嬗变系统及高燃耗反应堆设计提供重要的基础数据。文章对核数据分类、现行主要全套评价数据库及全套核数据评价方法进行阐述,并对234U(n,f)和237Np(n,2n)反应截面的实验数据进行评价。完成的裂变核全套中子数据整体满足用户需求,比原评价结果有较明显改进。     

Evaluation of high plutonia (44% PuO2) MOX as a fuel for fast breeder test reactor

Uranium plutonium mixed oxide (MOX) containing up to 30% plutonia is the conventional fuel for liquid metal cooled fast breeder reactor (LMFBR). Use of high plutonia (>30%) MOX fuel in LMFBR had been of interest but not pursued. Of late, it has regained importance for faster disposition of plutonium and also for making compact fast reactors. Some of the issues of high plutonia MOX fuels which are of concern are its chemical compatibility with liquid sodium coolant, dimensional stability and low thermal conductivity. Available literature information for MOX fuel is limited to a plutonium content of 30%. Thermodynamic assessment of mixed oxide fuels indicate that with increasing plutonia oxygen potential of the fuel increases and the fuel become more prone to chemical attack by liquid sodium coolant in case of a clad breach. In the present investigation, some of these issues of MOX fuel have been studied to evaluate this fuel for its use in fast reactor. Extensive work on the out-of-pile thermo-physical properties and fuel-coolant chemical compatibility under different simulated reactor conditions has been carried out. Results of these studies were compared with the available literature information on low plutonia MOX fuel and critically analyzed to predict in reactor behaviour of this fuel containing 44% PuO₂. The results of these out-of-pile studies have been very encouraging and helped in arriving at a suitable and achievable fuel specification for utilization of this fuel in fast breeder test reactor (FBTR). As a first step of test pin irradiation programme in FBTR, eight subassemblies of the MOX fuel are undergoing irradiation in FBTR.     

Multigroup neutron diffusion equation with the finite volume method in reactors using MOX fuels

The use of mixed oxide (MOX) fuel to partially fill the cores of commercial light water reactors (LWRs) gives rise to a reduction of the radioactive waste and production of more energy. However, the use of MOX fuels in LWRs changes the physics characteristics of the reactor core, since the variation with energy of the cross sections for the plutonium isotopes is more complex than for the uranium isotopes. Although the neutron diffusion theory could be applied to reactors using MOX fuels, more emphasis on treatment of the energy discretization should be placed. This energy discretization could be typically 4–8 energy groups, instead of the standard 2-energy group approach. In this work, the authors developed a finite volume method for discretizing the general multigroup neutron diffusion equation. This method solves the eigenvalue problem by using Krylov projection methods, in which the size of the vectors used for building the Krylov subspace does not depend on the number of energy groups, but it can solve the multigroup formulation with upscattering and fission production terms in several energy groups. The method was applied to MOX reactors for its validation.     

Preliminary study of the tight lattice pressured heavy water reactor loaded with Pu／U and Th／U mixed fuels

To improve nuclear fuel utilization efficiency and prolong fuel cycle burn-up,a tight ptich lattice pressured heavy water reactor was investigated as an alternative of next generation of power reactors.It is shown that the high conversion ratio and negative coolant void reactivity coefficient are challenges in the reactor core physics designs.Various techniques were proposed to solve these problems.In this work.a tight pitch lattice and mixed fuel assemblies pressured heavy water reactor concept was investigated.BY utilizing numerical simulation technique,it is demonstrated that reactor core mixed with Pu/U and Th/U assemblies can achieve high conversion ratio(0.98) ,long burn-up(60GWD/t)and negative void reactivity coefficients.     

Safety performance comparation of MOX, nitride and metallic fuel based 25–100 MWe Pb–Bi cooled long life fast reactors without on-site refuelling

In this paper the safety performance of 25–100 MWe Pb–Bi cooled long life fast reactors based on three types of fuels: MOX, nitride and metal is compared and discussed. In the fourth generation NPP paradigm, especially for Pb–Bi cooled fast reactors, inherent safety capability is necessary against some standard accidents such as unprotected loss of flow (ULOF), unprotected rod run-out transient over power (UTOP), unprotected loss of heat sink (ULOHS). Selection of fuel type will have important impact on the overall system safety performance.

The results of safety analysis of long life Pb–Bi cooled fast reactors without on-site fuelling using nitride, MOX and metal fuel have been performed. The reactors show the inherent safety pattern with enough safety margins during ULOF and UTOP accidents. For MOX fuelled reactors, ULOF accident is more severe than UTOP accident while for nitride fuelled cores UTOP accident may push power much higher than that comparable MOX fuelled cores.     

New Package for the Transport of Fresh MOX Fuel Assemblies in Europe

Abstract

The use of Mixed Oxide Fuels (MOX) in commercial reactors has increased significantly over the past 10 years as an effective way of using stocks of plutonium produced from reprocessing uranium fuels. Now, with advances in fuel design, MOX can give performance approaching that of enriched uranium fuel. To meet demand from European and Japanese utilities, British Nuclear Fuels are in the process of commissioning a large capacity plant at Sellafield to assemble MOX fuels. This has required a new transport package to be developed capable of carrying high specification fuels to customers in Europe whilst complying with the 1996 IAEA ST-1 Transport Regulations. This package is known as Euromox and currently under development to enter service in 2003. Relatively few packages exist for the transport of MOX fuels and Euromoxis the first designed by BNFL for shipments to Europe. Euromox has provided several technical challenges in its development arguably exceeding those typically encountered during the development of new package for irradiated fuel transports.     

Investigation on spent fuel characteristics of reduced-moderation water reactor (RMWR)

The spent fuel characteristics of the reduced-moderation water reactor (RMWR) have been investigated using the SWAT and ORIGEN codes. RMWR is an advanced LWR concept for plutonium recycling by using the MOX fuel. In the code calculation, the ORIGEN libraries such as one-group cross-section data prepared for RMWR were necessary. Since there were no open libraries for RMWR, they were produced in this study by using the SWAT code. New libraries based on the heterogeneous core modeling in the axial direction and with the variable actinide cross-section (VXSEC) option were produced and selected as the representative ORIGEN libraries for RMWR. In order to investigate the characteristics of the RMWR spent fuel, the decay heat, the radioactivity and the content of each nuclide were evaluated with ORIGEN using these libraries. In this study, the spent fuel characteristics of other types of reactors, such as PWR, BWR, high burn-up PWR, full-MOX-PWR, full-MOX-BWR and FBR, were also evaluated with ORIGEN.

It has been found that about a half of the decay heat of the RMWR spent fuel comes from the actinides nuclides. It is the same with the radioactivity. The decay heat and the radioactivity of the RMWR spent fuel are lower than those of full-MOX-LWRs and FBR, and are the same level as those of the high burn-up PWR. The decay heat and the radioactivity from the fission products (FPs) in the spent fuel mainly depend on the burn-up and the burn-up time rather than the reactor type. Therefore, the decay heat and the radioactivity from FPs in the RMWR spent fuel are smaller, reflecting its relatively long burn-up time resulted from its core characteristics with the high conversion ratio. The radioactivity from the actinides in the spent fuel mainly depends on the ²⁴¹Pu content in the initial fuel, and the decay heat mainly depends on ²³⁸Pu and ²⁴⁴Cm. The contribution of ²⁴⁴Cm is much smaller in RMWR than in MOX-LWRs because of the difference in the spectrum. In addition, from the waste disposal point of view, the characteristics of the heat generation FP elements, the platinum group metals, Mo and the long-lived FPs (LLFPs) were also investigated.     

A nondestructive method for discriminating MOX fuel from LEU fuel for safeguards purposes

Plutonium-rich mixed oxide fuel (MOX) is increasingly used in thermal reactors. However, spent MOX fuel could be a potential source of nuclear weapons material and a safeguards issue is therefore to determine whether a spent nuclear fuel assembly is of MOX type or of LEU (Low Enriched Uranium) type.     

Benchmark Problem Suite for Reactor Physics Study of LWR Next Generation Fuels

This paper proposes a benchmark problem suite for studying the physics of next-generation fuels of light water reactors. The target discharge burnup of the next-generation fuel was set to 70GWd/t considering the increasing trend in discharge burnup of light water reactor fuels. The UO₂ and MOX fuels are included in the benchmark specifications. The benchmark problem consists of three different geometries: fuel pin cell, PWR fuel assembly and BWR fuel assembly. In the pin cell problem, detailed nuclear characteristics such as burnup dependence of nuclide-wise reactivity were included in the required calculation results to facilitate the study of reactor physics. In the assembly benchmark problems, important parameters for in-core fuel management such as local peaking factors and reactivity coefficients were included in the required results. The benchmark problems provide comprehensive test problems for next-generation light water reactor fuels with extended high burnup. Furthermore, since the pin cell, the PWR assembly and the BWR assembly problems are independent, analyses of the entire benchmark suite is not necessary: e.g., the set of pin cell and PWR fuel assembly problems will be suitable for those in charge of PWR in-core fuel management, and the set of pin cell and BWR fuel assembly problems for those in charge of BWR in-core fuel management.     

Burning of MOX fuels in LWRs; fuel history effects on thermal properties of hull and end piece wastes and the repository performance

The thermal impacts of hull and end piece wastes from the reprocessing of MOX spent fuels burned in LWRs on repository performance were investigated. The heat generation rates in MOX spent fuels and the resulting heat generation rates in hull and end piece wastes change depending on the history of MOX fuels. This history includes the burn-up of UO₂ spent fuels from which the Pu is obtained, the cooling period before reprocessing, the storage period of fresh MOX fuels before being loaded into an LWR, as well as the burn-up of the MOX fuels. The heat generation rates in hull and end piece wastes from the reprocessing of MOX spent fuels with any of those histories are significantly larger than those from UO₂ spent fuels with burn-ups of 45 GWd/THM. If a temperature below 80°C is specified for cement-based materials used in waste packages after disposal, the allowable number of canisters containing compacted hull and end pieces in a package for 45 and 70 GWd-MOX needs to be limited to a value of 0.4–1.6, which is significantly lower than 4.0 for 45 GWd-UO₂.     

Estimation of Spent Fuel Compositions from Light Water Reactors

The compositions and quantities of minor actinide (MA) and fission product (FP) in spent fuels will be diversified with the use of high discharged burnup fuels and MOX fuels in LWRs which will be a main part of power reactors in future.

In order to investigate above diversities, we have studied on the calculation method to be used in the estimation of spent fuel compositions and adopted the real irradiation calculation in which axial burnup and moderator distribution are considered in the burnup calculation.

On the basis of the calculations, compositions and burnup quantities of various LWR spent fuels (reactor type: PWR and BWR, discharged burnup: 33, 45 and 60 GWd/tHM, fuel type: U0₂ and MOX) are apparently estimated among various forms of fuels. As an example, it is shown that there are considerable discrepancy in MA burnup between PWR and BWR spent fuels.