Development of a MCNP5 and ORIGEN2 based burnup code for molten salt reactor

doi:10.1007/s41365-016-0070-1

CAS OpenIR > 中科院上海应用物理研究所2011-2018年

	Development of a MCNP5 and ORIGEN2 based burnup code for molten salt reactor
	Sun, GM; Cheng, MS; Sun, GM (reprint author), Chinese Acad Sci, Shanghai Inst Appl Phys, Shanghai 201800, Peoples R China.; Sun, GM (reprint author), Univ Chinese Acad Sci, Beijing 100049, Peoples R China.
	2016
发表期刊	NUCLEAR SCIENCE AND TECHNIQUES
ISSN	1001-8042
卷号	27 期号:3 页码:-
文章类型	期刊论文
摘要	The Molten Salt Reactor (MSR) is one of the six advanced reactor nuclear energy systems for further research and development selected by Generation IV International Forum (GIF), which is distinguished by its core in which the fuel is dissolved in molten fluoride salt. Because fuel flow in the primary loop, the depletion of MSR is different from that of solid-fuel reactors. In this paper, an MCNP5 and ORIGEN2 Coupled Burnup (MOCBurn) code for MSR is developed under the MATLAB platform. Some new methods and novel arrangements are used to make it suitable for fuel flow in the MSR. To consider the fuel convection and diffusion in the primary loop of MSR, fuel mixing calculation is carried out after each burnup time step. Modeling function for geometry with repeat structures is implicated for reactor analysis with complex structures. Calculation for a high-burnup reactor pin cell benchmark is performed using the MOCBurn code. Results of depletion study show that the MOCBurn code is suitable for the traditional solid-fuel reactors. A preliminary study of the fuel mixture effect in MSR is also carried out.
关键词	Molten Salt Reactor Burnup Mocburn Mcnp Origen2
DOI	10.1007/s41365-016-0070-1
收录类别	SCI
语种	英语
WOS记录号	WOS:000380038900016
引用统计
文献类型	期刊论文
条目标识符	http://ir.sinap.ac.cn/handle/331007/26474
专题	中科院上海应用物理研究所2011-2018年
通讯作者	Sun, GM (reprint author), Chinese Acad Sci, Shanghai Inst Appl Phys, Shanghai 201800, Peoples R China.; Sun, GM (reprint author), Univ Chinese Acad Sci, Beijing 100049, Peoples R China.
推荐引用方式 GB/T 7714	Sun, GM,Cheng, MS,Sun, GM ,et al. Development of a MCNP5 and ORIGEN2 based burnup code for molten salt reactor[J]. NUCLEAR SCIENCE AND TECHNIQUES,2016,27(3):-.
APA	Sun, GM,Cheng, MS,Sun, GM ,&Sun, GM .(2016).Development of a MCNP5 and ORIGEN2 based burnup code for molten salt reactor.NUCLEAR SCIENCE AND TECHNIQUES,27(3),-.
MLA	Sun, GM,et al."Development of a MCNP5 and ORIGEN2 based burnup code for molten salt reactor".NUCLEAR SCIENCE AND TECHNIQUES 27.3(2016):-.

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