MATXS格式多群核截面数据加工平台研制与CMGC1.0数据库验证  

Development of MAXTS Format Multi-group Cross-section Processing Platform and Verification of CMGC1.0 Library

作　　者：杨寿海 曹南凤 刘杰 熊军 祖铁军 徐宁 曹良志 YANG Shouhai;CAO Nanfeng;LIU jie;XIONG Jun;ZU Tiejun;XU Ning;CAO Liangzhi(State Key Laboratory of Nuclear Power Safety Technology and Equipment,Shenzhen of Guangdong Prov.518172,China;School of Nuclear Science and Technology,Xi’an Jiaotong University,Xi’an of Shaanxi Prov.710049,China)

机构地区：[1]核电安全技术与装备全国重点实验室,广东深圳518172 [2]西安交通大学核科学与技术系,陕西西安710049

出　　处：《核科学与工程》2023年第6期1250-1257,共8页Nuclear Science and Engineering

基　　金：国家重点实验室长期基础项目(K-A2018.424)。

摘　　要：中广核工程有限公司与西安交通大学在NECP-Atlas程序的基础上联合开发了MATXS格式多群截面加工平台(CXROS),用于处理ENDF-6格式的评价核截面数据。研制过程通过需求分析、理论算法说明、程序设计和编码、测试与验证等流程的控制保证了研制过程的高可靠性。基于新研制的多群核截面数据加工平台,采用ENDF/B-Ⅶ.1评价核截面数据库,开发了适用于压水堆核电厂和干法贮存容器临界计算的361群中子的MATXS格式多群截面数据库CMGC1.0,并使用DRAGON4程序以及WLUP临界基准题对其进行基准验证。验证结果表明,CMGC1.0数据库的临界基准平均偏差为0.93%,最大偏差为3.68%,可满足压水堆乏燃料组件干法贮存容器临界设计的工程应用需求。本工作可以为核截面加工平台和截面数据库的加工与验证提供借鉴。China Nuclear Power Engineering Co.,Ltd and Xi'an Jiaotong University jointly developed MATXS format multi-group cross-section processing platform(CXROS) based on NECP-Atlas program,which is used to process evaluated nuclear data in ENDF-6 formats.The development process of CXROS ensures the high reliability of the platform through the control of requirements analysis,theoretical algorithm explanation,program design and coding,testing and verification,etc.A MATXS format multi-group library set named CMGC1.0 with 361-neutron group based on ENDF/B-Ⅶ.1 evaluated nuclear data and CXROS is used for the criticality analyses of Pressurized Water Reactor Nuclear Power Plant and spent fuel dry storage container.DRAGON4 program and WLUP criticality benchmarks are used to verify the validation of the library.The verification results show that the average deviation of CMGC 1.0 library is 0.93%,and the maximum deviation is 3.68%,which can meet the engineering application requirements of criticality design for spent fuel dry storage container used for storing spent fuel assemblies from Pressurized Water Reactor.This work can provide reference for the processing and verification of nuclear cross-section processing platform and cross-section database.

关 键 词：MATXS格式多群截面加工平台 ENDF/B-Ⅶ.1评价库 MATXS格式多群截面数据库 WLUP临界基准题 

分 类 号：TL48[核科学技术—核技术及应用]