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中国评价核数据库最新版CENDL-3.2(Chinese Evaluated Nuclear Data Library)已于2020年6月发布,对包括核工程计算中常用的235U、238U、239Pu、56Fe等134个核素的中子反应数据重新进行了评价和计算,与CENDL-3.1相比,CENDL-3.2数据种类和数据质量均有大幅提高。Be由于其散射截面大、吸收截面小,常被用作熔盐堆燃料载体盐成分之一,其反应截面数据的准确性在熔盐堆设计中不容忽视。基于CENDL-3.2评价核数据库,采用NJOY制作了199群中子、42群光子的MATXS格式多群截面库,挑选了35个含Be快临界基准对其进行检验分析,并与基于ENDF/B-7.1和JENDL-4.0的多群截面库计算结果进行对比。分析表明:基于CENDL-3.2多群截面库计算的26个基准题(74.29%)的结果与实验值偏差在0.5%以内,整体上优于ENDF/B-7.1和JENDL-4.0。表明CENDL-3.2中的Be数据和基于CENDL-3.2的多群截面库及其制作方法是可靠的,能够用于熔盐堆相关设计计算。 相似文献
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MCNP程序用热中子散射数据制作和检验 总被引:2,自引:2,他引:0
基于ENDF/B-Ⅶ.0评价库,以前已陆续研制了可供MCNP程序使用的连续截面库,以及多套多个温度、多组邦达连柯背景截面修正的多群参数库。本文采用NJOY程序以及ENDF/B-Ⅶ.0评价库热散射子库,完成了MCNP程序使用热中子散射数据库S(α,β)的制作和检验。比较了自制库与MCNP自带基于ENDF/B-Ⅵ版热散射数据库(sab2002),对改进较明显的重要介质“轻水中氢”和“重水中氘”给出了分析说明。通过48个基准装置keff计算结果可看出,MCNP程序自带热中子散射库sab2002与自制库thb70计算的keff整体上偏差不大,keff平均偏差约65pcm。 相似文献
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《原子能科学技术》2015,(Z1)
对于采用水冷方式的聚变堆,主要的放射性源项是水冷回路中的活化腐蚀产物,它会对反应堆的屏蔽设计、人员防护以及事故后果产生重要影响。本文为水冷聚变堆开发活化腐蚀产物源项分析程序CATE,该程序基于两项均匀模型构建浓度平衡方程组,全面考虑了活化腐蚀产物在水冷回路中的主要行为,包括腐蚀、释放、溶解、沉积、活化、衰变、净化等,并采用Runge-Kutta方法对浓度平衡方程组进行数值求解。使用CATE程序对国际热核聚变实验堆(ITER)的偏滤器冷却回路进行建模仿真,计算得到了活化腐蚀产物的成分和放射性活度在冷却剂中和管壁上的分布以及随时间的变化规律。与国际上同类程序PACTITER和TRACT相比,CATE程序的计算结果无论是在数值上还是趋势上都是合理的,可为ITER和CFETR(中国聚变工程实验堆)等的技术评审提供源项数据支持,在增加相应数据库后,还可应用于液态金属冷却反应堆的源项分析。 相似文献
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活化产物为压水堆核电站中主要辐射源,有必要对其建立分析手段。分析了压水堆核电站堆芯外材料中活化产物源项的产生途径,建立了压水堆核电站堆芯外材料中活化产物源项的计算模型,并分别基于矩阵指数法和切比雪夫有理近似法求解所建立的计算模型。开发了具有良好人机界面的计算程序CPAP,并采用典型材料活化例题与国外同类软件进行了对比测试。测试结果表明:CPAP程序对于测试算例的计算结果与国外同类软件的计算结果之间的偏差在工程可接受的范围内。CPAP程序具有人机界面友好以及求解器可选的优点,可广泛应用于压水堆核电站的设计、运行和退役阶段。 相似文献
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《辐射防护》2015,(Z1)
研究反应堆相关结构材料活化源项,对核电厂设计、运行及退役都有十分积极的意义和价值。本文利用离散纵标程序DORT计算反应堆堆腔内的中子注量率空间分布情况,通过数值解析的方法计算反应堆堆腔内主要结构材料中活化产物的活度浓度,进而计算活化源强(即γ射线源强,表征γ射线发射率与γ射线能量的关系),分析并建立一套空间分布活化源项研究体系,并与基于点燃耗模型的ORIGEN程序计算结果进行比较。计算结果表明,在活化源强计算中,基于离散纵标法的活化源强计算方法,在堆内构件等中子注量率变化明显之处拥有显著的精度,而ORIGEN程序则比较适合于厂房空间及主设备等中子注量率变化不明显之处。 相似文献
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轻水堆燃料组件计算程序包TPFAP 总被引:4,自引:4,他引:0
TPFAP是一个同时适用于PWR和BWR的穿透几率法燃料组件燃耗计算程序包。它首先利用碰撞几率方法在库能群结构下完成三区或四区圆环几何的栅元输运计算。载钆燃料棒或硼棒可燃毒物栅元的有效吸收截面由微燃耗程序CMB产生,两维穿透几率法组件计算是在(x,y)几何下进行。基模计算用来考虑中子泄漏修正。根据反应率等效,计算组件等效扩散参数。在每一燃料棒和可燃毒物棒进行燃耗计算,TPFAP给出每一燃耗步的组件和栅元少群截面、功率分布,提供核设计和安全分析所需参数。 相似文献
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压水堆核电厂乏燃料组件源项计算分析 总被引:1,自引:1,他引:0
核燃料贮存、运输以及后处理过程中的安全是构成核与辐射安全的重要内容,为保证安全性,提高运输经济性,减小后处理厂对环境的排放,须获得乏燃料组件的包络源项,因此,采用ORIGEN-ARP程序分析组件运行历史、初始富集度、燃耗深度等参数对源项的影响。运行历史在卸料初期对源项略有影响,可采用合适的保守因子予以包络,在冷却一定时间后,其影响可忽略不计;初始富集度、燃耗深度均不同的组件须经对比计算以获得包络源项。计算表明:在目前核电厂乏燃料组件中,235U初始富集度为4.45%、燃耗深度为55 GW•d/tU的AFA-3G型组件源项是包络的,可作为乏燃料水池、运输容器设计,以及后处理厂排放源项分析的初始源项。 相似文献
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An effective homogenization method has been developed for heterogeneous assemblies such as fuel assemblies with and without control blades in BWR and control-rod channels in FBR. Effective homogenized cross sections are calculated so as to preserve the integrated reaction rates in a heterogeneous assembly in each group by iteratively changing the cross section used in homogeneous super-cell calculations in a model composed of the heterogeneous assembly and a fuel region. The method has been applied to the rod-worth calculation for pin rods in the fast critical assembly ZPPR-10 and to the power-density calculation of a test BWR core. 相似文献
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全陶瓷微密封(FCM)燃料是一种弥散颗粒燃料。由于弥散颗粒燃料存在双重非均匀性,传统的确定论方法及蒙特卡罗方法皆难以处理这种双重非均匀效应以获得有效多群截面。本文基于超细群方法建立FCM燃料的有效多群截面计算方法。为描述燃料棒内TRISO颗粒的非均匀性,在共振能量段,通过采用超细群方法求解包含TRISO颗粒的一维球模型得到超细群缺陷因子,通过超细群缺陷因子修正所有核素的超细群截面即可将颗粒和基质均匀化。由于TRISO颗粒在热能区也存在较强的自屏效应,在热能区,利用穿透概率及碰撞概率等价得到多群缺陷因子,通过多群缺陷因子修正所有核素的多群截面将燃料和基质均匀化。均匀化后的FCM燃料组件即可视为普通压水堆燃料组件进行共振计算。利用丹可夫修正因子等价得到FCM燃料组件各燃料棒的等效一维棒模型,对一维棒模型求解超细群慢化方程从而得到共振能量段的有效自屏截面。数值结果表明,该方法能有效处理FCM燃料的双重非均匀性,得到精确的有效自屏截面。 相似文献
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Yuki Honda Sadao Uchikawa Yoshiaki Oka 《Journal of Nuclear Science and Technology》2013,50(5):645-655
A fast and thermal neutron coupled core adopts blanket fuel assemblies with zirconium hydrides in the core for negative coolant void reactivity. Conventional neutronics calculation methods have been developed for analysis of a fast core or thermal core, in which the coarse-group macroscopic cross sections of fuel assemblies are prepared without including the effect of the surrounding fuel assemblies. However, such methods are not adequate for analyzing fast and thermal neutron coupled cores where the intra-assembly and inter-assembly heterogeneity effects must be precisely taken into account. Recently, a concept of reconstruction of cell homogenized macroscopic cross sections has been proposed to take into account effects of inter-assembly heterogeneities on macroscopic cross sections used in the reactor core analysis and successfully applied based on a Monte Carlo method. In the present study, a reconstruction method of cell homogenized coarse-group macroscopic cross section for analyzing fast and thermal coupled cores is developed based on a deterministic neutronics calculation code system, SRAC. Three types of fixed source calculations for unit assembly cell geometry are performed independently of the specific core layouts and their results are combined with the results of core analysis to produce cell homogenized coarse-group macroscopic cross sections. Numerical results show that the heterogeneity effects can be adequately reflected in the reconstructed macroscopic cross sections with the proposed method. When the number of energy groups is small, the proposed method gives poor results in the transitional energy groups from resonance to thermal energy. Therefore, it is necessary to increase the number of energy groups in this energy range. 相似文献
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《Journal of Nuclear Science and Technology》2013,50(7):1087-1093
The method for the establishment of an equilibrium core model proposed in the previous paper and the source term calculation method proposed in this paper for the characterization of decommissioning waste were verified by comparing the nuclide inventory estimated by MCNP/ORIGEN2 simulations with the measured nuclide inventory according to a chemical assay in an irradiated pressure tube discharged from Wolsong Unit 1 in 1994. At first, the time-average pseudoequilibrium full-core model of Wolsong Unit 1 was developed on the basis of the previously proposed modeling method for the activation of in-core and ex-core structural components. Then, the application level of the neutron flux and cross section in the radionuclide buildup calculation were compromised. Fourteen major actinides and fission products were considered to represent the irradiated fuel condition, and a geometry simplification was also introduced in the burned full-core model for MCNP simulation. The assumption of a constant neutron flux and capture cross section as a function of the irradiation time was applied in the radionuclide buildup calculation in ORIGEN2. As a result, the values estimated from the analysis system agreed with the measured data within a difference range of 30%. Therefore, it was found that the MCNP/ORIGEN system and source term characterization method proposed can be viable to estimate the source terms of the decommissioning waste from a CANDU reactor. 相似文献
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OpenMC是麻省理工大学计算反应堆物理组开发的开源蒙特卡罗程序,能够方便地制作适用于特定堆芯中子能谱分布的多群反应截面及高阶勒让德散射截面以用于离散坐标输运程序ANISN的计算。本文基于ENDF/B-Ⅶ.1和CENDL-3.1评价数据库,利用OpenMC计算制作了ANSIN格式的多群截面并通过基准题的计算验证计算结果的准确性。通过截面转换程序的编写,将OpenMC给出的堆芯各阶勒让德散射分量,堆芯中子能谱分布,散射、吸收反应率以及裂变中子产生速率等信息转换为ANISN程序可读取的截面库格式。采用制作的截面库利用ANINS计算有效中子增殖因子及堆芯中子通量分布。结果表明,ANISN确定论的计算结果与OpenMC给出的蒙特卡罗计算结果相吻合,验证了这种方法可有效地为ANISN提供截面数据,将来可推广应用于二维、三维确定论中子输运计算。 相似文献
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为验证光纤激光用于燃料组件解体和燃料棒切割的可行性,研究光纤激光用于热物性差别很大的UO2芯块 不锈钢包壳管复合结构的切割和铀芯块的切割质量,本文采用光纤激光切割UO2芯块 316Ti包壳管元件棒,并通过扫描电子显微镜、能谱和X射线衍射对UO2芯块的切断面进行微观表征分析,研究激光切割过程对铀芯块切断的表面微观形貌、元素组成及物相的影响。研究结果表明,光纤激光可用于切割UO2芯块 316Ti包壳管元件棒,激光切割过程虽会造成铀芯块切断面出现大量微孔和碎渣,但不会造成UO2的相变。以上结果表明,光纤激光可用于UO2芯块 316Ti包壳管元件棒的切割,通过后续对激光切割系统的抗辐射屏蔽防护,可应用于乏燃料组件解体和乏燃料棒切割。 相似文献
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混合能源堆裂变包层核燃料成本分析 总被引:1,自引:0,他引:1
混合能源堆裂变包层燃料管理策略是:对乏燃料做后处理,得到的回收燃料作为下一循环的燃料,据此开展裂变包层的燃耗性能分析。在此基础上建立了针对混合能源堆的燃料循环成本分析模型:建立核燃料循环图,进行物料衡算,并分析燃料管理方案的单位发电量的燃料消耗量,根据市场价格,得到最终的核燃料成本。根据燃料循环成本分析结果,对影响较大的因素,如天然铀采购单价、乏燃料后处理单价、燃料制造单价等参数进行敏感性分析,得到燃料成本根据各价格参数变化规律。 相似文献
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A shielding analysis was performed which included a source term analysis of an Advanced Voloxidation Process where a spent oxide fuel is oxidized at a high temperature and vacuum condition by removing from it some volatile and semi-volatile elements such as tritium, krypton, xenon, and cesium. The source terms were estimated based on a spent PWR fuel using the ORIGEN-ARP calculation by considering the removal rate of the specific isotopes in the Advanced Voloxidation Process. The shielding problem was constructed by using a simple concrete box containing the spent fuel source which was processed by the Advanced Voloxidation Process, and the surface dose rates were estimated by Monte Carlo calculations. In order to quantify the advantage of the Voloxidation Process, the thicknesses of the concrete wall before and after the Advanced Voloxiation Process were compared. It was found that the Advanced Voloxidation Process provided a slight reduction in the burden on the radiation shielding. 相似文献