采用TVS-2M组件的VVER堆芯燃料管理研究

使用KASKAD程序包,对田湾核电站从首循环开始使用TVS-2M组件展开研究,提出相应的燃料组件设计。以此为基础展开燃料管理研究,提出3个燃料管理方案(年换料方案和两个长周期换料方案)。对每个方案中堆芯的安全参数及其他重要参数进行分析,结果表明各种安全参数均满足设计要求。长周期的换料方案是从首循环就开始使用TVS-2M组件,并且只经过2个循环的过渡,寿期长度便达到了长周期的要求。长周期换料方案可提高电厂的年均能力因子,并在整个堆芯寿期内减少大修次数,因而每年节约30.8%的大修费用,因此电厂的经济效益得以提高。     

田湾核电站3、4号机组的长周期燃料管理

田湾核电站3、4号机组计划从首循环起使用TVS-2M组件,采用由年换料过渡到长周期换料的燃料管理方案,并且业主对平衡循环长度提出了更高的要求,目标是510 EFPD。本文使用KASKAD程序包,对从年换料快速过渡到长周期换料展开研究,并给出了两个燃料管理方案。对方案中堆芯的安全参数及其他重要参数进行了分析,并对平衡循环的经济性作了简单的分析评价。结果显示,2个燃料管理方案的安全参数均满足设计要求。该研究成果具有工程应用价值,可应用在田湾核电站3、4号机组和VVER-1000堆型中,提高电站经济性。     

田湾核电站18个月换料燃料管理策略

为实现田湾核电站1、2号机组长周期换料项目,制定了过渡到18个月换料的燃料管理策略。参考俄罗斯核电站管理经验,长周期换料项目需采用TVS-2M新型燃料组件。在正式向长周期燃料循环过渡前通过TVS-2M先导组件运行验证了混合堆芯相容性。TVS-2M组件入堆替换AFA组件将分成两个阶段,即前两个过渡循环装入带有包覆层的TVS-2M组件,后继循环装入不带包覆层的TVS-2M组件。田湾核电站两台机组经历4个过渡循环,逐步延长运行时间,最终达到480 EFPD的循环长度。过渡循环和平衡循环均采用部分低泄漏堆芯装载,降低了对反应堆压力容器的中子辐照。田湾核电站18个月换料燃料管理策略提高了机组能力因子和经济性并具有灵活的循环长度。     

田湾核电站1号机组第6循环堆芯装载策略及验证

田湾核电站1号机组第5次换料大修期间,根据燃料组件检查结果,开展了紧急换料设计。1号机组第6循环堆芯装载策略具有不同于正常换料的特点,例如燃料装载不对称、部分辐照过的燃料组件移动到对称象限、堆芯功率分布不对称等。另外,堆芯装载策略考虑了TVS-2M先导燃料组件的位置要求。经第6循环寿期初物理试验和堆内测量系统验证,堆芯装载方案设计结果满足各项测量准则要求,且堆芯运行参数符合设计预期。     

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。     

田湾核电厂长周期混合堆芯稳态热工水力分析

田湾核电厂1、2号机组计划自2014年开始向长周期燃料循环过渡,在AFA型燃料组件组成的堆芯中逐步装入TVS-2M新型燃料组件,经过3个燃料循环的过渡,堆芯将全部装载TVS-2M型燃料组件,以实现长周期燃料循环。燃料组件结构的改变使原堆芯热工水力分析不再适用。本文以长周期燃料循环过渡时期的5种典型堆芯组成情况为例,介绍了VVER机组稳态热工水力分析的程序和方法,对混合堆芯的稳态热工水力特性进行了重新分析。结果表明,混合堆芯稳态设计仍满足热工水力设计准则。     

采用TVS-2M组件的堆芯燃料管理及其经济性分析

田湾核电站3、4号机组正在考虑使用TVS-2M组件来提高经济性。本文使用KASKAD程序包,对田湾核电站从首循环起使用TVS-2M组件进行研究设计,给出了改进型的燃料管理方案。对采用和未采用TVS-2M组件的两种燃料管理方案进行了经济性分析。分析结果显示,采用TVS-2M组件可显著提高电站经济性。     

一体化先进压水堆小型核电站堆芯燃料管理设计

采用SCIENCE核程序包进行装载方案的设计计算,确定了满足设计准则的各个过渡循环至平衡循环的堆芯.选择合理的平衡循环堆芯燃料的富集度、换料燃料组件数以及各循环的装载和换料方式,使平衡循环达到预定的2 a换料循环长度.堆芯采用低泄漏"内-外"式布置,旧燃料组件布置于堆芯外区.第一循环堆芯,高富集度的组件置于堆芯外区,低富集度的组件排列在堆芯内区.第二循环堆芯装入44个富集度为4.95%的新燃料组件,同时卸出44个旧燃料组件,旧燃料组件布置于堆芯外区.第三循环开始到反应堆寿期内的所有堆芯,都只使用含0、12和20根载钆燃料棒的燃料组件.各循环燃料组件最大卸料燃耗满足设计准则要求.     

压水堆多堆联合堆芯装载设计技术应用研究

在压水堆之间开展的多堆联合堆芯装载设计技术能提高堆芯装载设计的灵活性,研究其对燃料经济性和机组在燃料组件损坏情况下应对能力的影响。分析了多堆联合堆芯装载设计技术对来自其他机组燃耗过燃料组件的相容性要求。计算了燃耗过燃料组件的余热与冷却时间的关系,列出了可用于燃耗过燃料组件运输的大负荷乏燃料运输容器。分析了首循环堆芯和换料堆芯面对燃料组件损坏后堆芯装载设计的应对能力,以及在采用多堆联合堆芯装载设计技术后的应对能力改善情况。开展了多堆联合首循环堆芯装载模拟设计并分析了其经济性。研究结果表明,首循环出现燃料组件损坏的风险比后续换料堆芯大,而由全新燃料组件组成的首循环应对燃料组件损坏的能力最低;通过多堆联合堆芯装载设计技术可以把首循环转化为换料堆芯,不仅能凭借换料堆芯的设计灵活性提高应对燃料组件损坏的能力,还能节省约3.2亿元人民币的燃料费。因此,多堆联合堆芯装载设计技术能提高首循环的燃料经济性和机组面对燃料组件损坏后的应对能力。     

田湾核电站TVS-2M先导组件堆芯热工水力分析

田湾核电站一号机组于第5燃料循环装入6组TVS-2M先导燃料组件,并将经历从第5燃料循环到第8燃料循环4年的堆内运行。本文通过对先导燃料组件堆芯热工水力分析,堆芯运行实际试验测量以及组件变形检查,验证了热工水力设计程序计算模型的合理性以及计算结果与试验结果的符合性。结果表明,TVS-2M燃料组件与AFA燃料组件具有良好的相容性,从而证实了过渡循环条件下反应堆运行的安全性和可靠性。     

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。     

核燃料

在法国核然料工业组织中,法杰马公司主要销售燃料组件。法比燃料公司(FBFC)的3个从属工厂都负责燃料组件的制造,该公司每年生产装铀量为1500t 的燃料组件。自1985年以来,法杰马公司又销售先进燃料组件(AFA)。该 AFA 的主要特点是使用了锆合金定位格架和可拆式上、下管嘴。大亚湾核电站要用的燃料组件正是该种与一般组件不同的先进燃料组件。法杰马公司采用钆作可燃毒物,以保证燃料组件的良好特性。近来该公司又推出混合氧化物燃料组件(MOX)。由于法杰马公司在设计和制造的各阶段都严格遵守了质量保证和质量控制制度,所以其产品质量优良、可靠性好。展望未来,法杰马公司将与法国核燃料工业中的其它集团一起,努力为用户提供尽可能好的产品。     

Fuel R & D to Improve Fuel Reliability

Light water reactor fuel is operating in an increasingly challenging environment. Fuel burnup extension and cycle length increase both can increase the local duty. Reactor water chemistry modifications for the purpose of protection the plant system materials have the potential of increasing fuel surface deposition and cladding corrosion and hydriding. The status of fuel performance in US reactors is summarized and an update of the “Fuel Reliability Program” established by the utility industry to ensure reliability is provided.     

Reactivity Effect of Non-Uniformly Distributed Fuel in Fuel Solution Systems

A numerical method to determine the optimal fuel distribution for minimum critical mass, or maximum k-effective, is developed using the Maximum Principle in order to evaluate the maximum effect of non-uniformly distributed fuel on reactivity. This algorithm maximizes the Hamiltonian directly by an iterative method under a certain constraint—the maintenance of criticality or total fuel mass. It ultimately reaches the same optimal state of a flattened fuel importance distribution as another algorithm by Dam based on perturbation theory.

This method was applied to two kinds of spherical cores with water reflector in the simulating reprocessing facility. In the slightly-enriched uranyl nitrate solution core, the minimum critical mass decreased by less than 1% at the optimal moderation state. In the plutonium nitrate solution core, the k-effective increment amounted up to 4.3Δk within the range of present study.     

Characterization of High Burnup Fuel for Safety Related Fuel Testing

Fuel Assemblies designed and fabricated by Westinghouse Electric Sweden (WSE) to reach high burnup have been operated in the Leibstadt nuclear power plant (KKL) for seven cycles attaining an assembly average burnup above 60 MWd/kgU. The irradiation conditions in KKL featured linear heat generation rates ranging from 250 W/cm early in life down to 100W/cm in the last cycle and normal water chemistry with zinc injection. Selected rods have been extracted at both intermediate and final irradiation stages and hot cell examinations have been carried out. The results show that the fuel is well suited for high burnup applications and rod segments have been provided to the OECD Halden Reactor Program, the OECD Studsvik Cladding Integrity Program and the Japanese ALPS program for dedicated high burnup tests with regard to fission as release and cladding lift-off as well as behavior under power transient, RIA and LOCA.     

Fuel transient deformation

Cladding strains resulting from fuel-cladding mechanical interaction in a transient tested fresh fuel pin are assessed against laboratory measurements of high-temperature creep and hot pressing of mixed-oxide fuel pellets.A fuel pin containing nine different fuel sections with varying fuel pellet geometry and density was transiently tested in a MARK IIIA flowing sodium loop under conditions typical of a 1$/s overpower transient. Post-test cladding strain measurements indicated that the largest strains were generated by solid pellets with small gaps while large gap annular pellets generated the smallest strains.High-temperature creep and hot pressing tests on mixed-oxide fuel have been performed at temperatures up to 2600°C. The results indicate that at temperatures above 2300°C, an additional component of creep is operative; while the densification due to hot pressing was considerably less than expected by extrapolating the low-temperature behavior.Both the in- and out-of-reactor data suggest that fuel creep into the center void or hole of a fuel pin is a more effective means of reducing fuel-cladding stress than densification by hot pressing into fuel pellet porosity.     

Fuel Temperature Coefficients of Reactivity for Heavy-Water-Moderated,Cluster-Type Fuel Lattices

Temperature coefficients of reactivity have been measured up to 600°C on cluster-type UO₂ fuel for three kinds of ²³⁵U enrichment and on a hollow cluster of sus-cladding tubes by using a hot He gas loop in a heavy-water-moderated, pressure-tube-type critical assembly. A new experimental method has been developed which accurately eliminates the reactivity disturbance caused by heat leakage in the measurement of an extremely small change in reactivity. The fuel (fuel pellet, cladding land pressure-tube) temperature coefficients of reactivity obtained for the temperature range below 300°C are +1.00±0.04, ?3.48±0.13 and - 6.36±0.25 in the unit of l0^-5% Δk/k.°C for 0.2%, 0.7% and 1.5%²³⁵U enrichment, respectively. In the higher temperature region above 300°C, each coefficient shifts to positive side by about 2x10^-5 Δk/k.°C. Temperature coefficient of reactivity for the hollow cluster of sus-cladding tubes (cladding and pressure-tube) has a large constant value with positive sign, + (6.42±0.26) x 10^-5 Δk/k.°C, all through the temperature range. A calculational model to analyze a hot-loop-type measurement of temperature coefficients with use of WIMS-D code was proposed and could be successfully applied to the present measurement.     

157组燃料组件组成的堆芯燃料管理研究

本文应用SCIENCE程序包对157组燃料组件组成的压水堆堆芯进行换料优化燃料管理研究,给出了3个年换料和2个18个月换料共5个设计方案,每个设计方案给出了从首循环到第8循环共8个循环的主要计算结果,并进行了分析比较。综合来看,OUT-IN装载的设计方案功率峰值偏低,IN-OUT装载的设计方案功率峰值偏高,但均在设计限值以内;1/4堆芯换料设计方案的平均卸料燃耗最深,表明其组件燃耗得最充分,经济性较好。