Development and verification of an MC/MOC two-step scheme for neutronic analysis of FCM-fueled micro gas-cooled reactor

IF 1.9 3区 工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY Annals of Nuclear Energy Pub Date : 2024-09-29 DOI:10.1016/j.anucene.2024.110940
Kuaiyuan Feng , Qufei Song , Yuyang Shen , Lei Lou , Yao Xiao , Hui Guo , Hanyang Gu
{"title":"Development and verification of an MC/MOC two-step scheme for neutronic analysis of FCM-fueled micro gas-cooled reactor","authors":"Kuaiyuan Feng ,&nbsp;Qufei Song ,&nbsp;Yuyang Shen ,&nbsp;Lei Lou ,&nbsp;Yao Xiao ,&nbsp;Hui Guo ,&nbsp;Hanyang Gu","doi":"10.1016/j.anucene.2024.110940","DOIUrl":null,"url":null,"abstract":"<div><div>Gas-cooled microreactors are known for compact designs, high thermal-to-electric efficiencies, long refueling cycles, and flexible deployment capabilities, representing a groundbreaking solution to address the energy requisites of special scenarios. Fully ceramic microencapsulated (FCM) fuel is widely used in gas-cooled microreactors, bringing challenges to neutronic analysis methods. In this paper, an Monte Carlo/Method of Characteristics (MC/MOC) two-step scheme is developed and verified based on the reference case. In this scheme, the continuous-energy Monte Carlo calculations are used for reference calculation and multi-group cross-section generation. The multi-group Monte Carlo calculations are used for multi-group cross-section verification and the MOC solver verification. Fuel multi-group cross-sections are generated with the explicit fuel assembly model by continuous-energy Monte Carlo calculations, and structure multi-group cross-sections are generated with the simplified core model by continuous-energy Monte Carlo calculations. The core calculations are conducted with the MOC calculations. For verification, parameters such as power distribution, neutron spectrum, and control devices worth will be compared. Core calculation results show that the relative errors of MOC results are within −329 pcm, ±5%, ±6%, and ± 2 % for K<sub>eff</sub>, power distribution, neutron spectrum, and control device worth, separately. Moreover, the computation cost of MOC is only 6.6 % of the reference computation cost. The figure of merit results show that the MC-MOC scheme exhibits improved computational efficiency for neutronic analysis of FCM-fueled micro gas-cooled reactor.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9000,"publicationDate":"2024-09-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":"0","resultStr":null,"platform":"Semanticscholar","paperid":null,"PeriodicalName":"Annals of Nuclear Energy","FirstCategoryId":"5","ListUrlMain":"https://www.sciencedirect.com/science/article/pii/S0306454924006030","RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":null,"EPubDate":"","PubModel":"","JCR":"Q1","JCRName":"NUCLEAR SCIENCE & TECHNOLOGY","Score":null,"Total":0}
引用次数: 0

Abstract

Gas-cooled microreactors are known for compact designs, high thermal-to-electric efficiencies, long refueling cycles, and flexible deployment capabilities, representing a groundbreaking solution to address the energy requisites of special scenarios. Fully ceramic microencapsulated (FCM) fuel is widely used in gas-cooled microreactors, bringing challenges to neutronic analysis methods. In this paper, an Monte Carlo/Method of Characteristics (MC/MOC) two-step scheme is developed and verified based on the reference case. In this scheme, the continuous-energy Monte Carlo calculations are used for reference calculation and multi-group cross-section generation. The multi-group Monte Carlo calculations are used for multi-group cross-section verification and the MOC solver verification. Fuel multi-group cross-sections are generated with the explicit fuel assembly model by continuous-energy Monte Carlo calculations, and structure multi-group cross-sections are generated with the simplified core model by continuous-energy Monte Carlo calculations. The core calculations are conducted with the MOC calculations. For verification, parameters such as power distribution, neutron spectrum, and control devices worth will be compared. Core calculation results show that the relative errors of MOC results are within −329 pcm, ±5%, ±6%, and ± 2 % for Keff, power distribution, neutron spectrum, and control device worth, separately. Moreover, the computation cost of MOC is only 6.6 % of the reference computation cost. The figure of merit results show that the MC-MOC scheme exhibits improved computational efficiency for neutronic analysis of FCM-fueled micro gas-cooled reactor.
查看原文
分享 分享
微信好友 朋友圈 QQ好友 复制链接
本刊更多论文
开发并验证用于 FCM 燃料微型气冷堆中子分析的 MC/MOC 两步方案
气冷式微反应器以设计紧凑、热电联供效率高、加注周期长和部署灵活而著称,是解决特殊场景能源需求的开创性解决方案。全陶瓷微胶囊(FCM)燃料被广泛用于气冷式微反应器,这给中子分析方法带来了挑战。本文基于参考案例,开发并验证了蒙特卡罗/特征方法(MC/MOC)两步方案。在该方案中,连续能量蒙特卡罗计算用于参考计算和多组截面生成。多组蒙特卡罗计算用于多组截面验证和 MOC 求解器验证。燃料多组截面是通过连续能量蒙特卡洛计算使用显式燃料组件模型生成的,结构多组截面是通过连续能量蒙特卡洛计算使用简化堆芯模型生成的。堆芯计算与 MOC 计算一起进行。为了进行验证,将对功率分布、中子谱和控制装置价值等参数进行比较。堆芯计算结果表明,MOC 计算结果在 Keff、功率分布、中子谱和控制设备价值方面的相对误差分别在-329 pcm、±5%、±6%和±2%以内。此外,MOC 的计算成本仅为参考计算成本的 6.6%。优越性结果表明,MC-MOC 方案提高了 FCM 燃料微型气冷堆中子分析的计算效率。
本文章由计算机程序翻译,如有差异,请以英文原文为准。
求助全文
约1分钟内获得全文 去求助
来源期刊
Annals of Nuclear Energy
Annals of Nuclear Energy 工程技术-核科学技术
CiteScore
4.30
自引率
21.10%
发文量
632
审稿时长
7.3 months
期刊介绍: Annals of Nuclear Energy provides an international medium for the communication of original research, ideas and developments in all areas of the field of nuclear energy science and technology. Its scope embraces nuclear fuel reserves, fuel cycles and cost, materials, processing, system and component technology (fission only), design and optimization, direct conversion of nuclear energy sources, environmental control, reactor physics, heat transfer and fluid dynamics, structural analysis, fuel management, future developments, nuclear fuel and safety, nuclear aerosol, neutron physics, computer technology (both software and hardware), risk assessment, radioactive waste disposal and reactor thermal hydraulics. Papers submitted to Annals need to demonstrate a clear link to nuclear power generation/nuclear engineering. Papers which deal with pure nuclear physics, pure health physics, imaging, or attenuation and shielding properties of concretes and various geological materials are not within the scope of the journal. Also, papers that deal with policy or economics are not within the scope of the journal.
期刊最新文献
The nucleation characteristics of geyser boiling in sodium heat pipes European research reactor strategy derived in the scope of the towards optimized use of research reactors (TOURR) project Analysis of internal flow excitation characteristics of reactor coolant pump based on DMD On the Neutron Kinetics during a Promptcritical Accident in a Heavy Liquid Metal Fast Reactor and the Importance of Low-Energy Neutrons Machine-learned force fields for thermal neutron scattering law evaluations
×
引用
GB/T 7714-2015
复制
MLA
复制
APA
复制
导出至
BibTeX EndNote RefMan NoteFirst NoteExpress
×
×
提示
您的信息不完整,为了账户安全,请先补充。
现在去补充
×
提示
您因"违规操作"
具体请查看互助需知
我知道了
×
提示
现在去查看 取消
×
提示
确定
0
微信
客服QQ
Book学术公众号 扫码关注我们
反馈
×
意见反馈
请填写您的意见或建议
请填写您的手机或邮箱
已复制链接
已复制链接
快去分享给好友吧!
我知道了
×
扫码分享
扫码分享
Book学术官方微信
Book学术文献互助
Book学术文献互助群
群 号:481959085
Book学术
文献互助 智能选刊 最新文献 互助须知 联系我们:info@booksci.cn
Book学术提供免费学术资源搜索服务,方便国内外学者检索中英文文献。致力于提供最便捷和优质的服务体验。
Copyright © 2023 Book学术 All rights reserved.
ghs 京公网安备 11010802042870号 京ICP备2023020795号-1