MOX燃料组件装入现役M310堆芯对堆芯核设计的影响研究

国际上的MOX燃料技术目前已较为成熟,且已有在压水堆中运行的工程经验。本文对MOX燃料组件的中子学性能进行了分析,对其在我国现役M310堆芯应用的可行性进行了研究,得到了M310堆芯由全部使用UO2燃料组件向使用30%的MOX燃料组件过渡的堆芯燃料管理方案,并对使用MOX燃料组件的堆芯的部分中子学参数进行了初步分析。结果表明:使用30%的MOX燃料组件的堆芯可达到与全UO2堆芯相当的循环长度;堆芯反应性控制能力可满足要求;慢化剂温度系数、Doppler温度系数、Doppler功率系数、氙和钐的动态特性均趋向使堆芯运行更加安全和稳定。本文的研究结果可为MOX燃料在M310堆芯中应用的进一步研究提供参考。     

百万千瓦级环形MOX燃料堆芯设计

对环形UO2燃料及环形MOX燃料组件参数的计算方法进行了研究。设计了包含193盒环形UO2和MOX燃料组件的混合型长周期(18个月)堆芯方案。对设计的堆芯的重要物理参数进行了分析,并对各循环进行了燃耗计算。结果表明,装载约30%MOX组件的堆芯可在百万千瓦功率下实现长周期换料。堆芯从初装载可安全过渡到平衡循环,各循环的重要物理参数均满足设计要求,说明设计的堆芯及燃料管理方案是安全可行的。     

压水堆装载30%MOX燃料的燃料管理及经济性分析

初步分析我国现役大亚湾核电站M310堆芯的应用混合氧化物燃料(MOX)组件的可行性及经济性,给出M310堆芯由全堆装载UO2组件向使用30%MOX组件过渡的堆芯燃料管理方案。对使用MOX组件的堆芯的重要参数进行了分析,证明在现役大亚湾核电站M310堆芯应用MOX燃料是可行的。建立经济性分析模型,对所设计堆芯的燃料成本进行了具体分析。结果显示,MOX燃料的引入虽然大大提高了反应堆燃料成本,但仍存在较大的降低空间。     

大型压水堆装载50% MOX燃料方案初步研究

在保持堆内构件设计、燃料组件机械设计以及控制棒设计和布置不变的前提下,对大型压水堆应用MOX燃料进行初步研究。在遵守与UO_2堆芯相同的核设计准则的基础上,开展装载50%MOX燃料的堆芯燃料管理方案研究及核特性分析。分析结果表明,堆芯主要物理参数满足设计准则要求,能够实现堆芯运行和控制相关要求,具备装载50%MOX燃料的能力。混合堆芯的有效缓发中子份额比UO_2堆芯有所减小,其对弹棒事故的影响应予以重点关注。     

整个堆芯都采用MOX燃料的大间ABWR的堆芯设计

【日本《原子能视野》1999年10月刊第64～67页报道】1.整个堆芯采用MOX燃料的ABWR堆芯概要1.1基本考虑整个堆芯都采用MOX燃料的ABWR的燃料与堆芯设计的基本方针是,不改变以前ABWR的热功率、燃料组件的件数、控制棒的根数等基本规格,也不对以前的装铀燃料堆芯做大的变更。MOX燃料组件的基本构造与到目前为止已取得很好实绩的高燃耗8×8铀燃料(II级燃料)一样,在燃料组件的中央配置一根大口径挤水棒,在挤水棒周围配置排成8列8行的60根燃料棒。关于堆芯装MOX燃料组件问题,起初先装0～264根MOX燃料,然后阶段性地逐渐增加MOX燃料的比…     

基于MOX燃料组件的177混合堆芯装料方案数值研究

通过计算华龙一号（HPR1000）压水堆平均卸料燃耗得到乏燃料中钚（Pu）同位素的含量,以此成分比例来设计铀钚混合氧化物（MOX）燃料。采用离散型燃料组件设计,通过不同Pu含量的MOX燃料棒离散型布置来降低与UO₂燃料组件间的功率梯度。采用程序MCNP和COSLATC模拟堆芯功率分布和热中子注量率分布,采用分区分层的低泄漏装料方案,降低不同燃料组件间的功率梯度,展平堆芯的功率分布。在不考虑可燃毒物的前提下,利用3种Pu含量的MOX组件将混合堆芯的功率峰因子控制在1.77左右,明显优于原堆芯的功率峰因子,为国产三代压水堆引入MOX燃料提供了具有参考价值的装料方案。      

模块式小型压水堆ACP100堆芯燃料管理策略研究

为将多用途模块式先进小型压水堆(ACP100)堆芯换料周期提升至24个月,兼顾良好的燃料利用率并保证功率展平特性满足安全性的限值要求,本文使用具备工程经验的成熟软件,进行不同的批料数、不同富集度组合、不同径向装载模式等组合方案的燃料管理研究,通过研究掌握了不同策略特征并形成ACP100堆芯燃料管理推荐策略:选取3批次换料、24盒组件/批的倒料策略,结合部分低泄漏装载模式,作为ACP100堆芯燃料管理推荐策略,同时进一步提高燃料富集度以提升燃料经济性。     

聚变-裂变混合堆设计研究

利用MCNP5和MONK9A程序对聚变驱动裂变混合堆进行了初步研究,在等离子体第1壁外侧依次包覆长方体形状的燃料组件和产氚组件,形成裂变堆芯包层和产氚区.对分别装载贫铀、天然铀、贫铀MOX和天然铀MOX等4种燃料的混合堆进行了研究分析,其中,后两种燃料在整个运行寿期内的功率放大倍数和氚增殖比满足设计要求.通过随燃耗变化的同位素含量分析,初步探讨了混合堆的铀-钚燃耗循环策略.     

MOX燃料堆芯热工特性及设计限值研究

使用MOX燃料的快堆核电站以其线功率高、燃耗高、堆芯出口温度高等特点,对堆芯热工设计提出了新的问题.本文在对MOX燃料热工性能分析的基础上,给出了主要的热工设计限值,并以电功率870 MW电站为参考,初步分析了其堆芯热工特性和设计裕量.结果表明对于MOX燃料,较高的堆芯热工参数合理可行,且具有足够的裕量.     

压水堆内钍-铀增殖循环研究——堆芯设计

在全UOX（铀氧化物）堆芯平衡循环的基础上,研究了UOX/PuThOX（钚钍混合氧化物）混合堆芯和UOX/U3ThOX（工业级²³³U-钍混合氧化物）混合堆芯的燃料管理方案设计,实现了钍 铀增殖循环。U3ThOX燃料组件在堆内停留6个燃料循环,平均循环长度较参考的全UOX堆芯增加5 EFPD;U3ThOX燃料组件卸料后冷却1年时易裂变核素存量较装料时增加了7%。为比较分析,设计了UOX/MOX（钚铀混合氧化物）混合堆芯的燃料管理方案。核特性分析结果表明：1）装载PuThOX燃料对堆芯核特性产生的影响与装载MOX燃料类似,硼微分价值和控制棒价值减小、临界硼浓度变大、慢化剂温度系数更负、停堆裕量减小、多普勒亏损更大;2） UOX/U3ThOX混合堆芯和参考的全UOX堆芯具备相似的核特性。     

AP1000 core design with 50% MOX loading

The European uility requirements (EUR) document states that the next generation European passive plant (EPP) reactor core design shall be optimized for UO₂ fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO₂ core design and a mixed MOX/UO₂ core design, discussing relevant results related to reactivity management, power margin and fuel rod performance.     

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.     

Effective Utilization of Weapon-Grade Plutonium to Upgrade Repeatedly-Reprocessed Mixed-Oxide Fuel for Use in Pressurized Water Reactors

A “Multiple Recycling” mode of fuel management is proposed for effectively utilizing weapon-grade plutonium from discarded military material to compensate plutonium degradation in repeatedly-reprocessed mixed-oxide (MOX) fuel. Comparative calculations on core performance are undertaken for comparison between the proposed fuel management mode of Multiple Recycling—using recovered depleted plutonium upgraded by admixture with weapon-grade plutonium while retaining unincreased the total plutonium” content—and a reference mode of using repeatedly reprocessed spent MOX fuel with plutonium upgraded through increase of the plutonium content. Multiple Recycling proves all calculated safety parameters to be retained unimpaired through multiple cycles of MOX fuel reprocessing, whereas in the reference mode of refueling with spent MOX fuel reprocessed without upgrading with weapon-grade plutonium, many of the calculated safety parameters come to exceed stipulated limits with repetition of fuel cycles. Moreover, Multiple Recycling mode can be implemented with application solely of techniques already practiced in the fabrication of MOX fuel.     

Plutonium recycling in hydride fueled PWR cores

The classic approach to the recycling of Pu in PWR is to use mixed U-oxide Pu-oxide (MOX) fuel. The mono-recycling of plutonium in PWR transmutes less than 30% of the loaded plutonium, providing only a limited reduction in the long-term radiotoxicity and in the inventory of TRU to be stored in the repository. The primary objective of this study is to assess the feasibility of plutonium recycling in PWR in the form of plutonium hydride, PuH₂, mixed with uranium and zirconium hydride, ZrH_1.6, referred to as PUZH, that is loaded uniformly in each fuel rod. The assessment is performed by comparing the performance of the PUZH fueled core to that of the MOX fueled core. Performance characteristics examined are transmutation effectiveness, proliferation resistance of the discharged fuel and fuel cycle economics. The PUZH loaded core is found superior to the MOX fueled core in terms of the transmutation effectiveness and proliferation resistance. For the reference cycle duration and reference fuel rod diameter and pitch, the percentage of the plutonium loaded that is transmuted in one recycle is 53% for PUZH versus 29% for MOX fuel. That is, the net amount of plutonium transmuted in the first recycle is 55% higher in cores using PUZH than in cores using MOX fuel. Relative to the discharged MOX, the discharged PUZH fuel has smaller fissile plutonium fraction - 45% versus 60%, 15% smaller minor actinides (MA) inventory and more than double spontaneous fission neutron source intensity and decay heat per gram of discharged TRU. Relative to the MOX fuel assembly, the radioactivity of the PUZH fuel assembly is 26% smaller and the decay heat and the neutron yield are only 3% larger. The net effect is that the handling of the discharged PUZH fuel assembly will be comparable in difficulty to that of the discharged MOX assembly while the proliferation resistance of the TRU of the discharged PUZH fuel is enhanced.     

Design of an overmoderated fuel and a full MOX core for plutonium consumption in boiling water reactors

The use of uranium–plutonium mixed oxide fuel (MOX) in light water reactors (LWR) is nowadays a current practice in several countries. Generally 1/3 of the reactor core is loaded with MOX fuel assemblies and the other 2/3 with uranium assemblies. Nevertheless the plutonium utilization could be more effective if the full core could be loaded with MOX fuel. In this paper the design of a boiling water reactor (BWR) core fully loaded with an overmoderated MOX fuel design is investigated. The design of overmoderated BWR MOX fuel assemblies based on a 10×10 lattice are developed, these designs improve the neutron spectrum and the plutonium consumption rate, compared with standard MOX assemblies. In order to increase the moderator to fuel ratio two approaches are followed: in the first approach, 8 or 12 fuel rods are replaced by water rods in the 10×10 lattice; in the second approach, an 11×11 lattice with 24 water rods is designed with an active fuel length very close to the standard MOX assembly. The results of the depletion behavior and the main steady state core parameters are presented. The feasibility of a full core loaded with the 11×11 overmoderated MOX fuel assembly is verified. This design take advantage of the softer spectrum comparable to the 10×10 lattice with 12 water rods but with thermal limits comparable to the standard MOX fuel assembly.     

压水堆核燃料循环情景模式初步研究

根据我国核电发展现状和中长期发展规划及中长期（2030、2050）发展战略研究,假设2050年前我国压水堆核电发展规模,基于压水堆乏燃料后处理,回收的钚做成MOX燃料放入压水堆中使用,MOX燃料只使用1次的循环模式,进行核能发展情景研究。基于压水堆可装载30%比例MOX燃料的已有研究结果,考虑我国主要的两种压水堆堆型M310和AP1000,进行压水堆核燃料循环分析。利用核能发展情景动态分析程序DESAE-2,给出了不同情景模式下天然铀需求量、乏燃料累计量等。结果表明：至2050年,B₁和B₂模式较A模式分别节省天然铀4.1万t和2.9万t。     

Helios, current coupling collision probability method, applied for solving the NEA C5G7 MOX benchmark

As part of an effort to test the ability of current transport codes to treat reactor core problems without spatial homogenization, the lattice code HELIOS was employed to perform criticality calculations. The test consists in seven-group calculations of the C5 MOX fuel assembly problem specified by Cavarec et. al. [1]. This problem, known as C5G7 MOX Benchmark, is described in the Benchmark Specification [2] and comprises two cases — two and three-dimensional geometry. There are four fuel assemblies — two with MOX fuel, the other two with UO₂ fuel. Each fuel assembly is made up of a 17×17 lattice of square fuel-pin cells. Fuel pin compositions are specified in the Benchmark Specification, which also provides seven-group transport-corrected isotropic scattering cross-sections for U0₂, the three MOX enrichments, the guide tubes, the fission chamber and the moderator. This paper preset is the methodology employed in solving the C5G7 MOX Fuel Assembly Problem using the transport code HELIOS.