Reactor pressure vessel damage (dpa/s) calculation and testing of Fe-56 data libraries based on PCA benchmark model simulations using the SuperMC 3.4 code

doi:10.1016/j.anucene.2021.108694

HFCAS OpenIR

	Reactor pressure vessel damage (dpa/s) calculation and testing of Fe-56 data libraries based on PCA benchmark model simulations using the SuperMC 3.4 code
	Baidoo, I. K.1,2,3; Li, Bin 1; Wu, Bin1 ; Hao, Lijuan 1; Song, Jing 1; Shitsi, Edward 3
	2022-02-01
发表期刊	ANNALS OF NUCLEAR ENERGY
ISSN	0306-4549
通讯作者	Baidoo, I. K.(baidooisaac51@yahoo.co.uk)
摘要	Reactor Pressure Vessel (RPV) damage determination is important for reactor (power plant) safety and power economics, as its determination informs plant integrity and end-of-life extension. Irradiation damage determination is indirectly derived from the measurement/calculation of neutron fluence at or close to the pressure vessel system. The labyrinth procedure for experimentally determining the fluence by online operating plants makes theoretical (calculation) applications attractive. However, the calculation process must be validated via benchmark analysis, particularly according to the U.S Nuclear Regulatory Guide (Guide 1.190) for RPV dosimetry methods. One such related benchmark is the Pool Critical Assembly (PCA) of Oak Ridge. In this work, a detailed 3D model of the PCA facility and the calculated threshold reaction rate(s) have been achieved using the SuperMC code. Since iron (i.e., Fe-56) forms the major components of RPV, its evaluated data is key for benchmark calculations. Therefore, the calculation compares the Fe-56 data from ENDF/BVII.1, ENDF/BVIII.0, and a modified Fe-56-data by CIELO group (postENDF/BVIII.0). The calculated threshold reaction rates were achieved within 5%-10% deviation. The deviations were also identified to be dependent on the selected detectors' energy response spectrum. A detailed demonstration of displacement-per-atom rate (dpa/s) calculation is also presented, it showed contribution of about 5% and 95% for neutron energy <0.1 MeV and >0.1 MeV respectively. (C) 2021 Published by Elsevier Ltd.
关键词	SuperMC Neutron Fluence Rate RPV NRC Guide 1.190
DOI	10.1016/j.anucene.2021.108694
关键词[WOS]	DESIGN
收录类别	SCI
语种	英语
资助项目	Projects including the Special Program for Informatization of the Chinese Academy of Sciences[XXH13506-104] ; project of HIPS[KP-2019-13] ; Chinese Academy of Sciences and World Academy of Science (CAS-TWAS)'s Ph.D. scholarship scheme ; Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui
项目资助者	Projects including the Special Program for Informatization of the Chinese Academy of Sciences ; project of HIPS ; Chinese Academy of Sciences and World Academy of Science (CAS-TWAS)'s Ph.D. scholarship scheme ; Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui
WOS研究方向	Nuclear Science & Technology
WOS类目	Nuclear Science & Technology
WOS记录号	WOS:000709517500006
出版者	PERGAMON-ELSEVIER SCIENCE LTD
引用统计
文献类型	期刊论文
条目标识符	http://ir.hfcas.ac.cn:8080/handle/334002/125769
专题	中国科学院合肥物质科学研究院
通讯作者	Baidoo, I. K.
作者单位	1.Chinese Acad Sci, Inst Nucl Energy Safety Technol, Key Lab Neutron & Radiat Safety, Hefei 230031, Anhui, Peoples R China 2.Univ Sci & Technol China, Hefei 230027, Anhui, Peoples R China 3.Ghana Atom Energy Commiss, Natl Nucl Res Inst, Nucl Reactors Res Ctr, Box LG 80, Legon, Accra, Ghana
第一作者单位	中国科学院核能安全技术研究所
通讯作者单位	中国科学院核能安全技术研究所
推荐引用方式 GB/T 7714	Baidoo, I. K.,Li, Bin,Wu, Bin,et al. Reactor pressure vessel damage (dpa/s) calculation and testing of Fe-56 data libraries based on PCA benchmark model simulations using the SuperMC 3.4 code[J]. ANNALS OF NUCLEAR ENERGY,2022,166.
APA	Baidoo, I. K.,Li, Bin,Wu, Bin,Hao, Lijuan,Song, Jing,&Shitsi, Edward.(2022).Reactor pressure vessel damage (dpa/s) calculation and testing of Fe-56 data libraries based on PCA benchmark model simulations using the SuperMC 3.4 code.ANNALS OF NUCLEAR ENERGY,166.
MLA	Baidoo, I. K.,et al."Reactor pressure vessel damage (dpa/s) calculation and testing of Fe-56 data libraries based on PCA benchmark model simulations using the SuperMC 3.4 code".ANNALS OF NUCLEAR ENERGY 166(2022).