Inert matrix and thoria fuels for plutonium elimination

In our previous studies we analysed the plutonium burning performance of various kinds of fuel, based on mixing plutonium oxide with thorium oxide (TOX), or with inert matrix (IMF), or with inert matrix plus a limited addition (doping) of thorium oxide (TD-IMF). The present paper includes the neutronic analysis of a metal-based fuel variant and of fissile material recycling in TOX fuels. If the recycled fuel is topped with weapons grade plutonium (WG-Pu) as fissile material, it is possible to sustain indefinitely a closed cycle.     

Dynamics aspects of plutonium burning in an inert matrix

Burnup calculations have been performed on a mini fuel assembly containing 21 fuel rods and four water holes at the corners. The fuel rod positions were filled with 4% enriched UO₂ fuel and with either reactor grade or weapons grade plutonium mixed in an inert matrix. The ratio between the UO₂ and the IMF rods was varied to investigate the influence of the UO₂ fuel on the dynamics of the assembly. From a simple reactor model with one delayed neutron group and first-order fuel and temperature feedback mechanisms, the linear transfer function from reactivity to reactor power was calculated that was subsequently used in a root-locus analysis. From this, it is concluded that only 20% of the fuel rods need to be made of UO₂ to have a fuel that is linearly stable up to 1000 days of irradiation.     

混合氧化物燃料应用前景及核燃料循环战略

本文叙述了不同国家在不同核电发展条件下的核燃料循环战略初步考虑,并着重叙述了在快堆商业化进程推迟后所发展的混合氧化物(MOX)燃料的设计、制造技术及定量经济分析的基本情况,并建议我国近期应主要采用中间贮存并配以建造后处理中间试验工厂,为今后的商业后处理厂做好技术准备,同时还要跟踪 MOX 燃料技术。     

聚变驱动乏燃料焚烧堆(FDS-SFB)燃料循环动态分析

针对聚变驱动乏燃料焚烧堆FDS-SFB燃料循环系统与一次通过燃料循环系统,利用系统动力学软件Vensim分别建立了这两种循环系统的动态分析模型,并根据假设的三种核电发展情景,分别计算了这两种燃料循环系统的资源需求、乏燃料累积量、钚累积量及次锕系元素累积量。初步计算结果表明:与一次通过式燃料循环系统相比,FDS-SFB燃料循环系统可减少天然铀需求量与乏燃料累积量,减少的程度与核电发展规模相关。     

Current status of researches on the plutonium rock-like oxide fuel and its burning in light water reactors

Intention of the ROX-LWR system research is to provide an option for utilization or disposition of surplus plutonium. Researches on inert matrix materials and irradiation performance shows that the most favorable candidate for the ROX fuel is a particle dispersed fuel where small particles consisted of yttria stabilized zirconia, PuO₂ and some additives are homogeneously dispersed in spinel matrix. Reactor safety analyses show that the ROX fueled PWR core has nearly the same performability as the existing UO₂ fueled PWR under both reactivity initiated accidents and loss of coolant accidents.     

快堆核燃料循环模式的经济性评价

以快堆核电厂的核燃料循环过程及核燃料循环模型为基础,利用注销法对2种核燃料循环方式进行经济性计算和分析;同时,也将快堆燃料循环经济性与压水堆(PWR)燃料"一次通过"的经济性进行对比。按目前价格水平计算,PWR"一次通过"的核燃料循环方式比快堆核燃料循环模式的经济性好,但随着天然铀价格的上涨以及燃料后处理技术水平的进步,快堆核燃料循环费用有望达到或低于PWR"一次通过"的核燃料循环费用。     

核燃料循环中的技术与工业演变

本文介绍了从天然铀开采到燃料生产直到后处理的整个燃料循环中的技术和工业演变。核燃料循环的焦点是燃料组件的设计和制造。燃料制造业最重要的发展是以轻水堆 MOX 燃料形式出现的钚再循环。高杰马集团在燃料循环工业中处于领先地位。     

大亚湾核电站实施18个月换料项目的可行性研究

介绍大亚湾核电站实施18个月换料的可行性研究工作的内容、步骤和主要成果。包括目标确定、效益成本评价。电网需求调查、风险分析、技术指标。     

三种堆型核燃料循环经济性比较

本文详细地分析了压水堆、高温气冷堆、快堆三种堆型核电厂的燃料循环过程,建立了燃料循环费用的计算模型和灵敏度分析模型。对三种堆型的核燃料费用进行了分析比较,指出快堆是燃料费用最经济的堆型。     

世界铀市场的供求现状和核燃料循环决策浅析

本文介绍了近几年来国际形势变化后引起的世界铀市场供求现状，指出铀市场疲软的状况还将持续相当长的一段时期。在此基础上分析了各国核燃料循环末端决策的三种现状和原因。     

Inert matrix fuel for the utilisation of plutonium

Utilising fuel resources responsibly, reducing waste volume and emissions as well as conflict potentials within the international community (non-proliferation, energy demand) are among the principles for the judgment of sustainable development. Utilising and burning plutonium in a light water reactor has been shown to be feasible for the disposition of the large amount of excess plutonium produced in today's power reactors and resulting from the disarmament efforts of the super powers. With regard to material technology aspects, efforts have concentrated on the evaluation of fabrication feasibility and on the determination of the physicochemical properties of a single phase zirconium/erbium/plutonium oxide material stabilised as a cubic solution by yttrium for plutonium and minor actinide incineration and transmutation. Due to the absence of uranium as origin for plutonium build-up, such a nuclear fuel is called Inert Matrix Fuel. Irradiation testing with a dedicated experiment in the material test reactor in Halden, Norway, is underway within the framework of the OECD-Halden Reactor Project.     

Transition toward thorium fuel cycle in a molten salt reactor by using plutonium

The molten salt reactor(MSR), as one of the Generation Ⅳ advanced nuclear systems, has attracted a worldwide interest due to its excellent performances in safety, economics, sustainability, and proliferation resistance. The aim of this work is to provide and evaluate possible solutions to fissile 233 U production and further the fuel transition to thorium fuel cycle in a thermal MSR by using plutonium partitioned from light water reactors spent fuel. By using an in-house developed tool, a breeding and burning(BB) scenario is first introduced and analyzed from the aspects of the evolution of main nuclides, net 233 U production, spectrum shift, and temperature feedback coefficient. It can be concluded that such a Th/Pu to Th/~(233)U transition can be accomplished by employing a relatively fast fuel reprocessing with a cycle time less than 60 days. At the equilibrium state, the reactor can achieve a conversion ratio of about 0.996 for the 60-day reprocessing period(RP) case and about 1.047 for the 10-day RP case.The results also show that it is difficult to accomplish such a fuel transition with limited reprocessing(RP is 180 days),and the reactor operates as a converter and burns the plutonium with the help of thorium. Meanwhile, a prebreeding and burning(PBB) scenario is also analyzed briefly with respect to the net 233 U production and evolution of main nuclides. One can find that it is more efficient to produce 233 U under this scenario, resulting in a double time varying from about 1.96 years for the 10-day RP case to about 6.15 years for the 180-day RP case.     

Elements of comparison between different inert matrix fuels as regards plutonium utilisation and safety coefficients

The current international trend is to focus towards the utilization of plutonium. The use of composite fuels in inert matrix (U-free) is a potentially efficient solution to this problem. This document deals with the cermet fuels, selected for their excellent behaviour under irradiation and their high thermal conductivity. The emphasis is placed on the study of kinetic coefficients. Comparisons are performed with other solutions that use other composite fuels, especially the Solid Solutions and ROXs. As core control requires a heterogeneous assembly, an assembly whose characteristics are compared to the APA reference is proposed.     

Composite materials and systems as alternative inert matrix fuel to dispose civil or weapon grade plutonium in light water reactors.

To complete the IMF cercer studies on the problem of Pu utilization in LWRs, a cermet fuel approach is presented. The advantages of cermet fuel are associated with high heat conductivity, ability to retain the fission products and a well-developed fabrication process. Attractive possibilities for the creation of new cermet fuels and cermet fuel elements are also presented. R&D activity aimed at the development of cermet fuel element with PuO₂-Zr composite was undertaken. As a result of this activity comparative analysis of thermodynamic calculations for UO₂-Zr and PuO₂-Zr composites was carried out, as well as an assessment of Pu loading and preliminary thermal calculations. As a consequence, it was concluded that the PuO₂-Zr cermet system could be considered as a possible variant of new cermet fuel and cermet fuel element for Pu burning in LWRs.     

Utilization of spent PWR fuel-advanced nuclear fuel cycle of PWR／CANDU synergism

High neutron economy, on line refueling and channel design result in the unsurpassed fuel cycle flexi-bility and variety for CANDU reactors. According to the Chinese national conditions that China has both PWR and CANDU reactors and the closed cycle policy of reprocessing the spent PWR fuel is adopted, one of the advanced nu-clear fuel cycles of PWR/CANDU synergism using the reprocessed uranium of spent PWR fuel in CANDU reactor is proposed, which will save the uranium resource (-22.5%), increase the energy output (-41%), decrease the quantity of spent fuels to be disposed (-2/3) and lower the cost of nuclear poower, Because of the inherent flexibility of nuclearfuel cycle in CANDU reactor, and the low radiation level of recycled uranium(RU), which is acceptable for CANDU reactor fuel fabrication, the transition from the natural uranium to the RU can be completed without major modifica-tion of the reactor core structure and operation mode.It can be implemented in Qinshan Phase Ⅲ CANDU reactors with little or no requirement of big investment in new design. It can be expected that the reuse of recycled uranium of spent PWR fuel in CANDU reactor is a feasible and desirable strategy in China.     

Another option for separated plutonium management: Storage MOX

Currently there is no plan for the management of civilian plutonium that does not create a stockpile of separated plutonium. As a result, a number of nations with nuclear technology hold a large stockpile (about 240 tonnes) of separated plutonium. This paper suggests a timely, cost-effective solution for managing this material: storage MOX. A storage MOX plan would use existing MOX fuel fabrication facilities to make a simple MOX waste form suitable for long-term storage. Alternative waste forms to MOX are also possible, such as zirconia and pyrochlore, which provide more reliable durability and radiation damage control over thousands to hundreds of thousands of years.     

钚的利用与核裂变能的可持续发展

简要分析了当今世界的能源结构 ,指出以化石燃料为主的能源供应不可持续。概述了乏燃料后处理与钚的循环对充分利用铀资源的贡献 ,指出钚和其他锕系元素的彻底焚烧 ,有可能最大限度地减少放射性废物量及其毒性 ,从而实现核裂变能的可持续发展     

含钚材料恐怖事件的后果分析

基于数值仿真计算手段,定量分析了含钚材料恐怖事件中炸药爆炸条件下钚材料气溶胶的转化份额;针对不同可吸入气溶胶比例,应用HotSpot核事故后果评价模型计算了有关辐射特征量。结果表明,不同可吸入气溶胶比例对于辐射危害防护及处置有着明显的影响。     

田湾核电站长周期换料冷却剂流量丧失事故分析

田湾核电站采用长周期燃料循环策略后,堆芯热工物理参数发生变化,最终安全分析报告的结论已不适用,需要对事故工况进行重新分析。本文给出了失流事故分析的主要假设和分析方法,采用瞬态计算程序DINAMIKA-97计算分析了失流事故。分析结果表明,所有失流事故均满足安全准则的要求,核电站的安全是能够保障的。     

Core characteristics of fast reactor cycle with simple dry pyrochemical processing

Fast reactor core concept and core nuclear characteristics are studied for the application of the simple dry pyrochemical processing for fast reactor mixed oxide spent fuels, that is, the Compound Process Fuel Cycle, large FR core with half of loaded fuels are recycled by the simple dry pyrochemical processing. Results of the core nuclear analyses show that it is possible to recycle FR spent fuel once and to have 1.01 of breeding ratio without radial blanket region. The comparison is made among three kinds of recycle fuels, LWR UO₂ spent fuel, LWR MOX spent fuel, and FR spent fuel. The recycle fuels reach an equilibrium state after recycles regardless of their starting heavy metal compositions, and the recycled FR fuel has the lowest radio-activity and the same level of heat generation among the recycle fuels. Therefore, the compound process fuel cycle has flexibility to recycle both LWR spent fuel and FR spent fuel. The concept has a possibility of enhancement of nuclear non-proliferation and process simplification of fuel cycle.