CNSC乏燃料组件运输容器临界安全分析

临界安全作为乏燃料组件运输容器的一项重要安全指标,需经过计算和分析以判断其是否满足法规标准。为分析中国核工业集团有限公司(China National Nuclear Corporation,CNSC)乏燃料组件运输容器临界安全设计是否满足《放射性物品安全运输规程》的要求,使用蒙特卡罗程序MCNP(Monte Carlo N Particle Transport Code)构建了保守临界计算模型,对正常和事故工况下CNSC乏燃料组件运输容器进行了临界计算分析。分析表明:正常运输条件下单个货包和货包阵列的k_(eff)最大值为0.804 25,小于次临界限值,临界安全指数为0;事故工况下单个货包和货包阵列的k_(eff)最大值为0.813 17,小于次临界限值,临界安全指数为0。可见,正常和事故工况下,CNSC乏燃料组件运输容器的keff最大值均小于0.94的次临界限值,临界安全指数为0,满足法规标准要求。     

中国先进研究堆新燃料组件运输的临界安全分析

在新燃料组件运输过程中,临界安全是重点。使用MCNP程序对中国先进研究堆新燃料组件的运输进行临界安全计算分析,通过选取最不利临界安全的次临界限值、组件模型参数、事故工况来保证计算结果的保守性。结果表明,运输货包的临界安全指数可确定为0。该结果可为中国先进研究堆（CARR）的新燃料组件运输容器的研发提供参考依据。     

AP1000新燃料组件运输货包的临界安全计算

新燃料组件运输过程中最主要的核安全问题是临界安全。在对运输货包进行临界安全分析中必须要同时考虑多货包阵列形式、事故后货包损伤情况、最佳水慢化条件等因素。本文采用MCNP程序针对美国西屋公司XL型运输容器装载AP1000新燃料组件货包的实例进行了临界安全计算。结果表明,在XL型运输容器设计许可书中允许装载货包数N=75的限制条件下,临界安全是有保障的。     

UX-30型UF_6运输货包核临界安全分析

利用MONK-9A和MCNP程序对UX-30型UF6运输货包进行了正常与事故工况下的核临界安全分析与评价。首先选取国际公布的临界基准实验数据,验证并确定了MONK-9A和MCNP程序计算分析类似物料形态时的偏倚和次临界限值。其次采取较为包络的临界安全假设条件,计算分析了UX-30型UF6运输货包正常与事故工况下的中子有效增殖因数,评价了运输过程的安全性。计算结果表明,UX-30型UF6运输货包在最严重事故工况下最大的keff小于确定的次临界限值,处于次临界的安全状态。根据临界安全指数的定义,UX-30货包的临界安全指数CSI可定为0。     

高温气冷堆新燃料元件运输容器临界安全分析

采用基于蒙特卡罗方法的MCNP5程序对高温气冷堆所用的球形燃料元件进行描述;根据包覆燃料颗粒在燃料球内的分布性质构建了8种不同模型,并研究不同模型对有效增殖因子（keff）和计算时间的影响,获得了临界计算问题中最优的燃料球模型;运用MCNP5描述燃料球运输容器,并研究了容器中子吸收板厚度、外容器壁厚、缓冲层材料、反射层材料、容器形状、容器结构缺失和水密度等影响运输容器临界安全的因素。结果表明,所研究的高温气冷堆新燃料元件运输容器在正常运输条件下和事故运输条件下均处于临界安全状态,其临界安全指数（CSI）可定为0。      

高温气冷堆新燃料运输货包的撞击事故分析

介绍了高温气冷堆新燃料运输货包严重撞击事故的仿真计算分析方法。根据实际货包结构及运输条件,确定了分析的严重撞击事故景象。通过有限元法计算分析了货包在不同姿态、不同速度下的碰撞结果,给出了容器不同部分及所装载的燃料组件的损坏情况。在此基础上,计算了严重事故景象下有效增殖因子k_eff。     

燃料组件铁路运输栓系系统安全分析

某反应堆燃料组件的运输采用铁路运输,燃料组件运输容器的代号为MTR-D,采用栓系系统固定运输容器.针对燃料组件运输容器MTR-D,已经完成了正常和事故条件下的安全性分析.为论证栓系系统是否满足强度方面的要求,是否能够保证货包不会前后、左右以及垂直方向的移动,本工作采用经验公式,计算了运输过程中货包承受的力,同时校核了压紧螺杆的稳定性.计算结果表明,运输栓系系统能满足铁路运输燃料组件的要求.     

研究堆新燃料组件铁路运输货包固定系统安全分析

研究堆新燃料组件放置在专用货包内采用铁路运输,货包用木板、螺栓等进行紧固。对紧固系统在正常情况和极端情况下的受力情况进行了分析。强度校核计算表明,货包的固定方式能够保证其即使在10g的加速度下也不会在各个方向移动。稳定性校核计算表明,固定系统结构在此情况下也是稳定性的。实际运输监测中发现,全程最大加速度未超过10g,固定系统能够保障燃料组件运输容器的铁路运输安全。     

基于SCALE的RFA改进型燃料组件堆内贮存临界安全分析

RFA改进型燃料组件是西屋公司设计的能应用于大功率先进压水堆的改进型燃料组件。SCALE计算程序是一款在国际上得到广泛认可的综合性建模及模拟程序包,可用于核设计与核安全分析。基于SCALE计算程序,针对大功率先进压水堆的乏燃料贮存水池,建立恰当的计算模型,并选取合理的保守假设,分析乏燃料水池正常贮存及事故工况下的临界安全。计算结果表明一区正常贮存工况keff值为0.901 29,组件跌落事故工况下,有效增值因子为0.907 93。二区正常贮存工况下,计算模型keff值为0.909 98,新燃料组件误插入事故工况keff值为0.924 07。先进压水堆乏燃料贮存水池正常贮存工况及事故工况的有效增值因子均小于0.95,处于次临界状态。该设计模型可确保燃料堆内贮存区域临界状态安全可控。     

TP2008核数据库验证

TP2008是新研制的用于TPFAP程序的69群核数据库,本文利用IAEA压水堆棒状燃料组件基准问题和零功率临界实验结果对TP2008核数据库进行了验证分析。结果表明,燃料组件无限增殖因数k_∞与机构TUR的符合相对好;棒状燃料组件相对功率分布计算结果与参考程序的符合较好。零功率临界实验的堆芯有效增殖因数k_eff的相对偏差大部分在-0.5%以内,符合较好。     

运输容器临界安全评价要点剖析

易裂变物质的运输是堆外操作易裂变物质的主要活动之一,特别是随着越来越多核电厂、研究堆的投建或退役,新、乏燃料的运输临界安全问题备受关注。在对易裂变物质的运输进行临界安全评价时应遵循相关的法规要求,如GB 11806-2004《放射性物质安全运输规程》,这是我国易裂变材料运输要满足的强制性要求和准则。针对该标准制定的各项规定和要求,结合设计和评审中的工程实际经验,以1个新燃料运输容器的设计分析为例,探讨了易裂变物质运输时核临界安全评价的技术要求,为易裂变材料货包的设计、安全评审提供参考和建议。     

田湾核电站乏燃料干式贮存临界安全计算研究

以CASTOR 1000/19干式贮存容器装载田湾核电站六角形乏燃料组件为例,研究六角形乏燃料干式贮存的临界安全问题。基于新燃料假设,应用MONK9A程序对贮存容器满装载乏燃料进行不同工况下k_eff的计算。计算结果表明：正常工况下,k_eff远小于临界安全限值,是临界安全的;事故工况下,当²³⁵U富集度大于3.15%时,系统存在临界安全风险,须减少乏燃料装载量来确保临界安全。考虑燃耗信任制后,采用相同的模型计算得出贮存容器满装载的参考装载曲线,按此曲线要求装载能确保所有工况下的系统临界安全。采用燃耗信任制技术提高了贮存容器的利用率。该研究可为田湾核电站采用乏燃料干式贮存方案提供依据。     

Criticality assessments using polyurethane foam

Abstract

Rolls-Royce has designed a package to transport and store fresh fuel assemblies and anticipates approval from the regulators for the new package design in the near future. The space between the inner and outer steel shells is filled with shaped blocks of rigid polyurethane foam, of two different densities. The criticality safety case for the fresh fuel package had to consider single packages and arrays of packages under routine, normal and accident conditions. IAEA regulatory requirements state that the criticality assessment must include investigations on the effect on the neutron multiplication factor k _eff due to impacts, flooding and fire. Sensitivity studies must also be carried out to determine the effects on the k _eff due to any uncertainties in the composition of the fuel and container materials. An important part of the criticality safety case is the treatment of the foam. The approach adopted to model the polyurethane foam is the subject of this paper. The following were investigated: (1) the effect on the k _eff of varying the elemental composition of the foam, including the removal of hydrogen; (2) the experimental analysis of burnt foam; (3) the effect of addition of water to the foam to simulate water absorption; (4) a simple representation of crushed foam to simulate knock-back in the package; (5) extreme representations of burnt foam, such as replacing foam with solid carbon or as randomly distributed spheres of carbon to represent soot. These investigations were most informative and should be considered in any criticality assessments of transport packages containing large amounts of foam in the future.     

星球表面用核反应堆临界安全分析

核反应堆电源具有寿命长、可全天候工作等特点,可作为星球表面及其他深空探测任务的电源。针对星球表面用核反应堆电源在发射过程中重返地面的临界安全问题,提出了星球表面用核反应堆的临界安全分析要求、分析假设与模型,并对反应堆临界安全特性及采取的临界安全措施进行了计算分析。计算结果表明,不同假设掉落环境下的星球表面用核反应堆的有效增殖因数均小于0.98,满足临界安全要求。反应堆通过采用Mo-14%Re合金结构材料、设置相对较厚的堆芯反射层以及在反射层包壳和堆芯外围涂覆Gd₂O₃涂层等措施有利于确保反应堆在事故时处于次临界状态。     

Mechanical safety analysis for high burn-up spent fuel assemblies under accident transport conditions

Abstract

Transport packages for spent fuel have to meet the requirements concerning containment, shielding and criticality as specified in the International Atomic Energy Agency regulations for different transport conditions. Physical state of spent fuel and fuel rod cladding as well as geometric configuration of fuel assemblies are, among others, important inputs for the evaluation of correspondent package capabilities under these conditions. The kind, accuracy and completeness of such information depend upon purpose of the specific problem. In this paper, the mechanical behaviour of spent fuel assemblies under accident conditions of transport will be analysed with regard to assumptions to be used in the criticality safety analysis. In particular the potential rearrangement of the fissile content within the package cavity, including the amount of the fuel released from broken rods has to be properly considered in these assumptions. In view of the complexity of interactions between the fuel rods of each fuel assembly among themselves as well as between fuel assemblies, basket, and cask body or cask lid, the exact mechanical analysis of such phenomena under drop test conditions is nearly impossible. The application of sophisticated numerical models requires extensive experimental data for model verification, which are in general not available. The gaps in information concerning the material properties of cladding and pellets, especially for the high burn-up fuel, make the analysis more complicated additionally. In this context a simplified analytical methodology for conservative estimation of fuel rod failures and spent fuel release is described. This methodology is based on experiences of BAM acting as the responsible German authority within safety assessment of packages for transport of spent fuel.     

New approach to evaluate lattice expansion of light water reactor fuel elements on criticality safety of transport packages under impact accidents

Abstract

In order to safely transport packages containing light water reactor fuel assemblies, it is essential to maintain the fuel assemblies in a subcritical state in accidents during transport. To evaluate nuclear criticality safety, an estimator is required to determine an absolutely safe level based not only on hypothetical accidents but also on practical accident levels which, to some extent, are based on actual accidents. The purpose of the present study is to suggest the arrangement of the deformation range of the fuel assembly after an actual accident, and to obtain the maximum value of the neutron effective multiplication factor based on the criticality safety assessment for the transport cask. In the present study, two kinds of criticality calculations for the package were considered: large scale pin pitch shift and small scale pin pitch shift. For the large scale pin pitch shift, a parameter which determines the location of each fuel pin which constitutes the fuel assembly was introduced so that the criticality calculation for the fuel assembly with non-uniform lattice pitch can be performed parametrically. The result of the criticality calculation using the parameter made it clear that the fuel pin pitch is sensitive to the neutron reactivity because each of the fuel pin pitches is related to a ratio of the fissile to the moderator, and that the relationship of the ratio to the neutron reactivity depends on the type of the fuel assembly involved, i.e. the type of a nuclear reactor in which a fuel assembly is used. For the small scale pin pitch shift, the study focused on the small displacement of each fuel pin. The small displacement of each fuel pin pitch can be described probabilistically using the stochastic geometry routine in MCNP code. Using the scheme in combination with the scheme for the large scale pin pitch shift, the maximum value of the neutron effective multiplication factor of the package after an accident can be obtained. This scheme is useful to determine the maximum neutron effective multiplication factor for the criticality safety evaluation.     

Kilopower空间堆掉落事故临界安全问题研究

美国Kilopower空间堆在掉落事故下的k_eff不满足我国现行空间堆掉落临界安全要求。该反应堆在掉落过程中，若反射层外围的B₄C脱落，则存在瞬发超临界的严重安全隐患。针对此问题，本文对反应堆方案进行调整，提出3种解决方案，各方案均可满足掉落临界安全要求。此外，为研究各方案的优劣，从尺寸、质量、物理和热工运行特性等方面对各方案进行综合比较，提出了最优建议方案。     

高温气冷堆核电站示范工程新燃料运输辐射风险研究

探索了将概率安全评价（PSA）方法系统地应用于放射性物品运输的辐射风险评价,分析了高温气冷堆核电站示范工程（HTR-PM）新燃料元件公路运输的辐射风险。基于实际路况数据和可能的事故情景,选择货包辐射水平升高和临界两种事故工况进行了事故频率分析。分析表明：货包辐射水平升高事故的发生频率为4.21×10^-7（车•单次运输）^-1;临界事故的发生频率低于1×10^-13（车•单次运输）^-1,可不考虑其辐射后果。对事故后果估算的结果表明：货包辐射水平升高事故对应急人员造成的最大外照射剂量为0.55 mSv,对附近公众造成的最大外照射剂量为4.55×10^-3 mSv,其辐射影响是可接受的。总体辐射风险为1.24×10^-10人•Sv/(车•单次运输),其中撞击事故对风险的贡献最大。