乏燃料后处理溶解过程核临界安全初步分析

通过建立合理的空间分布模型,对后处理厂乏燃料溶解不同阶段的核临界安全问题进行分析,同时对重要的核临界安全参数给予影响评价。结果显示,在仅考虑易裂变核素形态转变的理想情况下,溶解初期为最危险状态;温度升高和硝酸浓度增大对系统的影响为负效应,影响均小于4%;可溶中子毒物的加入与燃耗信任制技术的应用能大幅提高系统的经济性,影响均可达到30%。     

最佳估算方法在核临界安全分析的应用研究

最佳估算方法可以同时对多个参数按概率分布进行抽样,从而模拟系统真实的物理状况,计算结果的容忍区间及置信水平与抽样数目有关。本文将最佳估算方法应用于压水堆核电站乏燃料贮存格架和燃料运输容器的临界安全分析,采用非参数抽样统计方法,多参数同时抽样,并对各抽样参数的敏感度进行分析。抽样计算的结果统计分析表明,最佳估算方法更接近真实值,证明原逐参数单独进行敏感性分析方法的保守性并得到相应的保守裕量;对于特定研究对象参数的敏感性排序是稳定的,主要取决于参数自身的敏感性,参数的范围及分布的影响较小,应在相关设备的设计与制造中重点关注敏感度高的参数。     

截短型燃料组件堆内贮存临界安全分析

基于小型压水堆特有的截短型燃料组件,建立乏燃料贮存水池几何模型,分析正常贮存及事故工况下的临界安全。选取合理的保守假设,建立适当的计算模型,分别计算了一区和二区正常贮存工况、地震事故工况、组件跌落事故工况、新组件误插入事故工况的反应性。计算得到事故工况下有效增值因子最大值为0.932 83。小型压水堆乏燃料贮存水池临界安全分析中,正常工况及事故工况下计算结果均小于0.95。该设计模型可确保燃料堆内贮存区域处于次临界状态,且安全可控。     

用蒙特卡罗方法进行核临界安全计算

本文介绍了蒙特卡罗程序ＡＭＰＸ－ＫＥＮＯ系统的铀富集厂核临界安全计算机中的应用,并为此作了大量临实验数据的验证计算和可用于实际生产的临界安全参数计算。     

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。     

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。     

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。     

基于SCALE的压水堆乏燃料贮存水池临界安全分析

基于国际先进的核设计与安全分析计算程序SCALE,针对我国自主研发的先进压水堆乏燃料贮存水池,建立恰当的计算模型,并选取合理的保守假设,计算乏燃料水池正常贮存及事故工况下的反应性,评估计算模型的临界安全,分析该程序对我国先进反应堆乏池计算的适用性。计算结果表明该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。SCALE计算程序适用于我国自主研发的先进压水堆乏燃料水池临界安全计算。     

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。     

PWR核电厂乏燃料贮存临界计算重要核素的选取

信用核素选取是基于燃耗信用制乏燃料贮存临界安全分析的关键一步。通过对不同富集度、燃耗深度及停堆冷却时间下典型PWR燃料组件分析,以核素中子吸收份额大小排序为依据,筛选出对总的中子吸收起主要贡献的核素。结果显示,47个核素即可包络停堆后0~20a内影响乏燃料贮存系统反应性的所有核素中的99%。通过核素敏感性因子分析证明依据中子吸收份额排序选取重要核素的方法是合理的,与基准算例的结果对比证明所筛选出的核素能足够代表影响系统反应性的所有重要核素。     

CEFR-MOX新燃料组件运输货包临界安全计算

易裂变材料运输过程中重要的安全问题之一是临界安全。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、事故后货包损伤对临界安全影响、最佳水慢化条件等因素。本文采用MCNP 程序针对CEFR-MOX新燃料组件运输货包进行了临界安全计算。计算结果表明:MCNP程序(采用核截面库为ENDF/B-V库)对本问题的次临界限值为0.924 6;正常运输条件下无限个运输货包的最大k_eff值为0.574 4,运输事故条件下无限个运输货包的最大k_eff值为0.659 7。根据临界安全指数的定义,确定CEFR-MOX新燃料组件运输货包的临界安全指数为0。     

浓缩铀柱形临界装置徒手装配临界安全检验

为完成徒手装配浓缩铀柱形临界装置临界安全检验实验,设计了基于薄膜装配的验证系统,以替代浓缩铀柱形临界装置上半部分结构。用MCNP程序计算了浓缩铀柱形临界装置上半部分及验证系统密合时的有效增殖因子keff。实验测得验证系统中心增殖不会超过12.50,满足徒手装配装置中心增殖限定值15的要求。实验结果表明,验证系统设计合理,徒手装配浓缩铀柱形临界装置是安全的。     

铀水系统核临界安全的蒙特卡罗计算

本工作涉及应用蒙特卡罗程序MCNP4B对铀水系统核临界实验数据进行验证计算和对740L容器取料时漏入CaCl₂盐水后形成的UO₂F₂-CaCl₂水溶液系统的有效增值系数k_eff的模拟计算。计算结果表明，MCNP4B程序对铀水系统核临界安全计算是有效的，漏入盐水后形成的均匀UO₂F₂-CaCl₂水溶液系统是核临界安全的。计算结果为实际生产中的核临界安全性提供了理论依据。      

Computation on Fuel Particle Size Capable of Being Regarded as Homogeneous in Nuclear Criticality Safety Analysis

For powdered fuel processed in the nuclear fuel facilities, flooding is often thought to be the severest condition regarding the nuclear criticality safety evaluation. The reactivity of such a heterogeneous system as powdered fuel in water should be almost equal to that of the homogeneous one, when the fuel particle size is very small. The neutron multiplication factor was calculated for an infinite cubic array of slightly enriched UO₂ sphere particles immersed in water with various enrichments, water to fuel ratios and fuel particle sizes. The calculations were performed with a computer code module based on the collision probability method to solve the ultra-fine energy group equations of neutrons. The change in the neutron multiplication factor from the homogeneous system is dominated first by the change in the resonance escape probability and second by the change in the thermal utilization factor; these changes and therefore their sum, depend almost completely on the mean uranium concentration (or water to fuel volume ratio) and rarely on uranium enrichment up to 10 wt% for a fuel particle size of 1mm. The dependence determines the fuel particle size regarded as homogeneous in proportion to the negligible relative error of the neutron multiplication factors.     

反应堆临界-燃耗耦合蒙特卡罗计算

基于连续点截面MCNP程序 ,研制了三维多群P3 中子输运蒙特卡罗程序MCMG ,并与栅元均匀化程序WIMS耦合 ,实现了临界 燃耗耦合计算。采用WIMS产生的 69群共振、自屏宏观中子截面和BUGLE 80u47群微观中子截面 ,分别计算了简单反应堆和临界实验堆问题 ,计算结果与其它输运方法的计算结果和试验结果一致。在相同计算精度下 ,MCMG的计算时间较MCNP的计算时间少     

Benchmark Analysis of Criticality Experiments in the TRIGA Mark II Using a Continuous Energy Monte Carlo Code MCNP

The criticality analysis of the TRIGA-II benchmark experiment at the Musashi Institute of Technology Research Reactor (MuITR, 100kW) was performed by the three-dimensional continuous-energy Monte Carlo code (MCNP4A). To minimize errors due to an inexact geometry model, all fresh fuels and control rods as well as vicinity of the core were precisely modeled. Effective multiplication factors (k_eff) in the initial core critical experiment and in the excess reactivity adjustment for the several fuel-loading patterns as well as the fuel element reactivity worth distributions were used in the validation process of the physical model and neutron cross section data from the ENDF/B-V evaluation. The calculated k_eff overestimated the experimental data by about 1.0Δk/k for both the initial core and the several fuel-loading arrangements (fuels or graphite elements were added only to the outer-ring), but the discrepancy increased to 1.8Δk/k for the some fuel-loading patterns (graphite elements were inserted into the inner-ring). The comparison result of the fuel element worth distribution showed above tendency. All in all, the agreement between the MCNP predictions and the experimentally determined values is good, which indicates that the Monte Carlo model is enough to simulate criticality of the TRIGA-II reactor.     

Modeling and criticality evaluation of the voloxidation process

Voloxidation is a necessary process in the dry reprocessing of spent nuclear fuels. The criticality evaluation plays a considerable role in the design of voloxidation apparatus. As conservative results are always preferred in a criticality evaluation, an optimized model was built in consideration of both the geometry of voloxidation apparatus and the occurring forms of evaluation material. The criticality evaluation of fresh UO₂ fuel and PWR spent fuel were then performed by employing Monte Carlo techniques, respectively. It is demonstrated that there is no criticality risk concerning the voloxidation process dealing with fresh UO₂ or PWR spent fuel if water does not intrude into the cell. However, if water intrudes and mixes with the fuels, the subcritical mass limit is 40.1 ± 0.1 kg for fresh UO₂ and 19,155 ± 50 kg for spent fuel. The contributions of ¹H and ²³⁵U were analyzed quantitatively by the TSUNAMI code to clarify the competition between ¹H moderation effect and its dilution effect on the concentration of ²³⁵U.     

Safety Concerning Transportation of Spent Fuel

The Central Research Institute of Electric Power Industry (CRIEPI) has been conducting, under contract with the Science and Technology Agency of Japan, the spent fuel transport cask reliability demonstration test since 1977 to verify the safety and reliability of spent fuel transport casks. The first phase of this test was completed in 1987.

In this demonstration test, both 50 t and 100 t class of casks, designed and manufactured by current techniques, were subjected to tests to verify the integrity and adequacy of the design and manufacturing techniques through observation of behavior of the cask under test conditions. The casks were subjected to tests under normal conditions and under the accident conditions specified in the Japanese regulations and the IAEA regulations, and also to pressure tests, which were performed from the viewpoint of safety in shipping, although by sea, this is not specified in the Japanese regulations.

From the test results, it was confirmed that the 1001 class cask maintained its integrity and characteristics in conformity with regulations even after accident condition tests. It is clear that the design concept and manufacturing procedure employed for this cask is adequate.     

基于启明星Ⅱ号的铅基堆核数据宏观检验

为提高铅基堆中子学模拟的可靠性,基于启明星Ⅱ号铅基零功率反应堆,开展铅基堆相关核数据的入堆宏观基准检验研究。采用周期法测量堆芯反应性,进而获得有效增殖因数keff为1001 14±0000 07。采用MCNP程序对铅基堆进行精细化建模,结合不同数据库内的中子评价核数据,计算实验燃料棒装载下的铅基堆芯的keff。比较结果可知,4种截面库计算的铅基堆keff模拟结果与实验结果吻合较好,最大相对偏差小于1%,其中,ENDF/B Ⅶ.1库的模拟结果与实验结果吻合最好,相对偏差和绝对偏差分别为025%和251 pcm。通过计算关键材料元素核数据引起keff的变化量,可知铅元素核数据引起的堆芯keff结果的波动量最大,在CENDL 31和JENDL 40中的铅元素引起keff的波动值分别为219 pcm和166 pcm。     

棱柱式高温气冷空间核反应堆初步方案设计与中子物理分析

空间核反应堆电源将核裂变能转换为电能,与太阳能、化学燃料电池等其他形式的电源相比,具有电功率大、系统比功率高、使用寿命长等优点,在太空探索中具有广阔的应用前景。以高温气冷堆技术为基础,提出了以氦氙混合气体作冷却剂,直接布雷顿循环的空间核反应堆电源方案。核反应堆是采用包覆颗粒燃料的小型棱柱式高温气冷堆,热功率为5 MW。采用蒙特卡罗方法进行了中子物理分析。结果表明,设计的反应堆满足10a以上的满功率运行寿期,具有负的反应性温度系数。通过布置B4C安全棒,使反应堆在发射失败引起的堆芯进水事故中能保证次临界。