Annals of Nuclear Energy最新文献_第2页

Physics analysis and design of heavy water reflected thermal test reactor 重水反射热试验反应堆的物理分析和设计

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-10-24 DOI: 10.1016/j.anucene.2024.110966

Hikaru Hiruta, Mark D. DeHart, Carlo Parisi

{"title":"Physics analysis and design of heavy water reflected thermal test reactor","authors":"Hikaru Hiruta, Mark D. DeHart, Carlo Parisi","doi":"10.1016/j.anucene.2024.110966","DOIUrl":"10.1016/j.anucene.2024.110966","url":null,"abstract":"<div><div>This work investigates the option of modifying the Advanced Test Reactor by replacing the current beryllium reflector with heavy water. Such a change may provide some potential benefits for not only increasing the thermal irradiation capabilities but also resolving other problems such as reflector integrity issues due to fast fluence damage, which is always a limiting factor in the lifetime of the current beryllium reflector. This paper presents the analysis and estimation of the ATR core physics parameters by replacing the current beryllium reflector with heavy water (D<sub>2</sub>O). The paper first describes the details of two selected conceptual designs, which are partially reflected with either beryllium or graphite, and how they are derived from the baseline beryllium reflector concept. Then, reactor physics performance parameters for the two new concepts are assessed by comparing with those of the baseline concept. The performance parameters considered in this paper include in-pile tube neutron and gamma fluxes and heating rates, maximum loop voiding reactivity, core power behavior with different power splits, predicted cycle length with a given fuel loading, and thermal hydraulic analysis with a higher lobe power split. It is important to note that this study focuses on the reactor physics aspects and does not delve into the engineering challenges associated with such a design modification.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142526956","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

CORTEX experiments – Part I: Modulation campaigns in AKR-2 & CROCUS for the validation of neutron noise codes CORTEX 实验--第一部分：AKR-2 和 CROCUS 的调制活动，用于验证中子噪声代码

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-10-23 DOI: 10.1016/j.anucene.2024.110928

Vincent Lamirand , Alexander Knospe , Klemen Ambrožič , Sebastian Hübner , Carsten Lange , Oskari Pakari , Fanny Vitullo , Adolfo Rais , Joachim Pohlus , Uwe Paquee , Christoph Pohl , Nicolas Weiss , Pavel Frajtag , Daniel Godat , Antonios Mylonakis , Axel Laureau , Thomas Ligonnet , Mathieu Hursin , Grégory Perret , Andreas Pautz

{"title":"CORTEX experiments – Part I: Modulation campaigns in AKR-2 & CROCUS for the validation of neutron noise codes","authors":"Vincent Lamirand , Alexander Knospe , Klemen Ambrožič , Sebastian Hübner , Carsten Lange , Oskari Pakari , Fanny Vitullo , Adolfo Rais , Joachim Pohlus , Uwe Paquee , Christoph Pohl , Nicolas Weiss , Pavel Frajtag , Daniel Godat , Antonios Mylonakis , Axel Laureau , Thomas Ligonnet , Mathieu Hursin , Grégory Perret , Andreas Pautz","doi":"10.1016/j.anucene.2024.110928","DOIUrl":"10.1016/j.anucene.2024.110928","url":null,"abstract":"<div><div>We present the experimental campaigns –namely, three per facility– carried out between 2018 and 2021 in the AKR–2 and CROCUS zero power reactors within the framework of the Horizon 2020 European project CORTEX. Their purpose was to produce high-quality and noise-specific experimental data for the validation of the neutron noise computational models developed in CORTEX. In both reactors, perturbations were induced by two devices, separately and altogether. In AKR–2, they consisted of a rotating absorber, i.e. an <em>absorber of variable strength</em>, and a linear oscillator, i.e. a <em>vibrating absorber</em>, both sets in horizontal channels close to the core. In CROCUS, the project benefited from the COLIBRI experimental program and its <em>fuel rods oscillator</em> set in the outer lattice; an additional <em>vibrating absorber</em> called POLLEN was set in a vertical air-channel at core center. The campaigns at both facilities consisted of neutron measurements with numerous detectors at reference static states, and with the addition of the mechanical perturbations to induce neutron reactivity modulation. The present article documents the experimental setups and measurements for each facility and perturbation type. A focus is set on the experimental designs and their evolution along the project, as well as motivations and learned lessons. Results are presented and discussed in details in associated papers.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142526947","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Generation and validation of a new WIMS-D library based on ENDF/B-VIII.0 生成并验证基于 ENDF/B-VIII.0 的新 WIMS-D 库

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-10-23 DOI: 10.1016/j.anucene.2024.110989

Jan Malec , Oscar Cabellos , Marjan Kromar , Andrej Trkov

{"title":"Generation and validation of a new WIMS-D library based on ENDF/B-VIII.0","authors":"Jan Malec , Oscar Cabellos , Marjan Kromar , Andrej Trkov","doi":"10.1016/j.anucene.2024.110989","DOIUrl":"10.1016/j.anucene.2024.110989","url":null,"abstract":"<div><div>The WIMSD-5B transport code is a deterministic tool for nuclear reactor core design and fuel management. It can efficiently handle pin-cell and supercell models and calculate homogenized cross sections, which are essential for reactor physics calculations. It is used by core design packages such as the CORD-2 package, developed at the Jožef Stefan Institute, and SEANAP developed by Universidad Politécnica de Madrid (UPM). The WLUP update project <span><span>https://www-nds.iaea.org/wimsd</span><svg><path></path></svg></span> demonstrated the way to update the WIMS-D libraries with different evaluated nuclear data libraries, including ENDF libraries up to version ENDF/B-VII.1. Using an updated version of the procedure, a new WIMS-D library based on the ENDF/B-VIII.0 data was developed to improve the accuracy of core design calculations. Several improvements to the library were made and the effects of each individual improvement was demonstrated using a 3×3 supercell benchmark model that is representative of a typical pressurized water reactor. Finally, the performance of the library over a diverse set of neutron transport problems was tested for, to ensure no regressions were introduced.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142526952","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Software-Based automation of burnup calculations for the NUR research reactor using SCALE/TRITON T6-DEPL sequence 利用 SCALE/TRITON T6-DEPL 序列对 NUR 研究堆进行基于软件的自动燃耗计算

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-10-23 DOI: 10.1016/j.anucene.2024.111007

Djahid Lababsa , Hakim Mazrou , Mohamed Belgaid , Tahar Zidi , Mohammed Azzoune , Azzeddine Ameur , Ahmed Guesmia , Leila Zamoun , Mohamed Boufenar

{"title":"Software-Based automation of burnup calculations for the NUR research reactor using SCALE/TRITON T6-DEPL sequence","authors":"Djahid Lababsa , Hakim Mazrou , Mohamed Belgaid , Tahar Zidi , Mohammed Azzoune , Azzeddine Ameur , Ahmed Guesmia , Leila Zamoun , Mohamed Boufenar","doi":"10.1016/j.anucene.2024.111007","DOIUrl":"10.1016/j.anucene.2024.111007","url":null,"abstract":"<div><div>This work represents a key milestone in the development of a Monte Carlo Burnup Calculation System (MCBCS) specially tailored for the NUR research reactor. Developed using Python and leveraging the 3D Monte Carlo TRITON depletion sequence (T6-DEPL) within the SCALE code, MCBCS accurately simulates the reactor’s operating history. The paper provides an overview of MCBCS, focusing on its components, verification, and validation.</div><div>The verification and validation process cover both fresh and burnt core conditions. For the fresh core, comparisons of excess reactivity, control rods worth, and critical configurations against experimental data and MCNP5 calculations showed good agreement. Burnup calculations were validated against measured core excess reactivity, reactivity worths of fuel assemblies, and neutron flux distribution. The system slightly underpredicted core excess reactivity by −2.95%, and discrepancies in reactivity worths remained within the 7% uncertainty range. Neutron flux distribution showed good consistency with minor location-specific deviations.</div><div>Overall, these findings confirm MCBCS as a reliable and accurate tool for burnup calculations of the NUR research reactor.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142526948","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Retention analysis of aerosol inside narrow channels of the containment 安全壳狭窄通道内的气溶胶滞留分析

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-10-22 DOI: 10.1016/j.anucene.2024.110987

Zhang Dandi , Wang Shanpu , Tong Lili , Cao Xuewu

{"title":"Retention analysis of aerosol inside narrow channels of the containment","authors":"Zhang Dandi , Wang Shanpu , Tong Lili , Cao Xuewu","doi":"10.1016/j.anucene.2024.110987","DOIUrl":"10.1016/j.anucene.2024.110987","url":null,"abstract":"<div><div>Aerosol retention inside narrow channels is the optimization direction of the leakage source term assessment for nuclear power plant containment. Based on the flow characteristics of carrier gas and the deposition characteristics of transported aerosol, a one-dimensional analysis method of aerosol retention in narrow channels is developed through considering different deposition mechanisms of inlet loss, gravity settlement, Brownian diffusion, turbulent deposition and steam condensation. The flow models of carrier gas and the retention models of aerosol are analyzed and verified, respectively. The flow of carrier gas deviates from laminar flow earlier through using the drag model of narrow channels. The prediction accuracy of aerosol penetration factor calculated by current analysis method in narrow channels is improved under laminar flow and turbulent flow through comparing with the previous calculation methods. Aerosol retention analysis is conducted on the narrow channels of steel containment under the typical severe accident. The turbulent deposition introduced by larger leakage channels increases the aerosols retention effect in narrow channels.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142526850","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Temperature fluctuation mitigation of heat pipe cooled reactor with closed Brayton cycle during load-following dynamic power regulation 采用封闭式布雷顿循环的热管冷却反应堆在负载跟随动态功率调节期间的温度波动缓解问题

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-10-22 DOI: 10.1016/j.anucene.2024.110986

Jingkang Li , Zunyan Hu , Zeguang Li , Liangfei Xu , Jianqiu Li

{"title":"Temperature fluctuation mitigation of heat pipe cooled reactor with closed Brayton cycle during load-following dynamic power regulation","authors":"Jingkang Li , Zunyan Hu , Zeguang Li , Liangfei Xu , Jianqiu Li","doi":"10.1016/j.anucene.2024.110986","DOIUrl":"10.1016/j.anucene.2024.110986","url":null,"abstract":"<div><div>Heat pipe cooled reactors (HPRs) offer the potential to achieve load-following control without the need for control rods or drums, thereby simplifying the control system. However, during load-following operation, HPRs experience fluctuations in temperature, which can impact safety. Limited research has focused on mitigating temperature fluctuations of HPRs during dynamic power regulation leveraging their inherent load-following capabilities. This study examines the characteristics of an HPR with closed Brayton Cycle (CBC), and develops a load-following control algorithm. A simplified CBC model is proposed to facilitate control strategy analysis. Model predictive control (MPC) is employed to suppress temperature fluctuations, revealing that the dynamic response of output power under MPC resembles that of a first-order inertial system. Consequently, a power control algorithm based on first-order inertial feedforward control is introduced. Simulation results demonstrate that the proposed algorithm, with a time constant ranging between 500 and 1000 s, significantly mitigates temperature and power fluctuations in HPRs during load-following dynamic power regulation.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142526945","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Verification of nuclear data libraries used to design molten salt blankets of a fusion neutron source 验证用于设计聚变中子源熔盐毯的核数据库

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-10-21 DOI: 10.1016/j.anucene.2024.110983

Yu.E. Titarenko, S.A. Balyuk, V.F. Batyaev, V.I. Belousov, I.A. Bedretdinov, V. Yu. Blandinskiy, V.D. Davidenko, I.I. Dyachkov, V.M. Zhivun, Ya.O. Zaritstkiy, M.V. Ioannisian, A.S. Kirsanov, A.A. Kovalishin, N.A. Kovalenko, B.V. Kuteev, V.O. Legostaev, M.R. Malkov, I.V. Mednikov, K.V. Pavlov, A. Yu. Titarenko, K.G. Chernov

{"title":"Verification of nuclear data libraries used to design molten salt blankets of a fusion neutron source","authors":"Yu.E. Titarenko, S.A. Balyuk, V.F. Batyaev, V.I. Belousov, I.A. Bedretdinov, V. Yu. Blandinskiy, V.D. Davidenko, I.I. Dyachkov, V.M. Zhivun, Ya.O. Zaritstkiy, M.V. Ioannisian, A.S. Kirsanov, A.A. Kovalishin, N.A. Kovalenko, B.V. Kuteev, V.O. Legostaev, M.R. Malkov, I.V. Mednikov, K.V. Pavlov, A. Yu. Titarenko, K.G. Chernov","doi":"10.1016/j.anucene.2024.110983","DOIUrl":"10.1016/j.anucene.2024.110983","url":null,"abstract":"<div><div>This study presents the results of testing nuclear data libraries by analyzing statistical criteria obtained from comparing experimental and calculated rates for (n,2n), (n,p), (n,pn), (n,nꞌγ) (n,α) and (n,γ) reactions measured on samples <sup>nat</sup>Ni, <sup>nat</sup>Zr, <sup>nat</sup>Nb, <sup>nat</sup>Cd, <sup>nat</sup>Ti, <sup>nat</sup>Co,<sup>63(96%), 65(99.70%)</sup>Cu, <sup>64(99.70%)</sup>Zn, <sup>nat</sup>In, <sup>nat</sup>Al, <sup>nat</sup>Mg, <sup>nat</sup>Fe, <sup>nat</sup>Au and <sup>nat</sup>Th, which were placed in the experimental channels of micromodels of the fusion blanket.</div><div>The “fast” (the cylinder Ø 230 mm and 520 mm length was filled with ∼ 67 kg of molten salt 0.52NaF + 0.48ZrF4) and the “thermal” blanket (the same cylinder was placed in a dry channel inside a cubic container filled with water with dimensions of 52.0 × 52.0 × 52.0 cm were investigated. The reaction rates were measured using the activation method.</div><div>Modeling with transport codes MCNP5, KIR, PHITS-3.31, SuperMC3.4.0 was performed using the ENDF/B-VII.0 library for neutron transport as well as seven neutron data libraries for reaction rates simulation, including: JEFF-3.3, JENDL-4.0, ENDF/B–VIII.0, ROSFOND-2010, FENDL-3.0, TENDL − 2019 and IRDFF-II.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142527012","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Insights into calculating Reference Discontinuity Factors with Serpent Monte Carlo code 使用蛇形蒙特卡洛代码计算参考不连续因数的启示

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-10-21 DOI: 10.1016/j.anucene.2024.110997

Emil Fridman , Jacob D. Smith , Dan Kotlyar

{"title":"Insights into calculating Reference Discontinuity Factors with Serpent Monte Carlo code","authors":"Emil Fridman , Jacob D. Smith , Dan Kotlyar","doi":"10.1016/j.anucene.2024.110997","DOIUrl":"10.1016/j.anucene.2024.110997","url":null,"abstract":"<div><div>This study explores the calculation of Reference Discontinuity Factors (RDFs) using the Serpent Monte Carlo code, focusing on the methodology and potential pitfalls. In two-step reactor analyses, consistently generated RDFs are crucial for aligning homogeneous nodal diffusion results with the reference heterogeneous transport solution. However, the Serpent internal diffusion solver, based on the Analytic Function Expansion Nodal (AFEN) method, may not be compatible with other nodal methods such as the Nodal Expansion Method (NEM). Additionally, the solver can suffer from instabilities, particularly in multi-group calculations, leading to erroneous RDFs. Despite these challenges, Serpent can generate the necessary raw data for RDF calculation, which can be accurately processed using external diffusion solvers. Two numerical examples − a 1D fuel-reflector model and a 2D SMR core model − illustrate the effects of consistent and inconsistent RDFs on simulation accuracy. The study emphasizes the importance of using compatible diffusion solvers and thoroughly assessing RDFs to avoid errors in reactor simulations.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142526943","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

A computational framework to support probabilistic criticality modelling for the geological disposal of radioactive waste 支持放射性废物地质处置临界概率建模的计算框架

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-10-21 DOI: 10.1016/j.anucene.2024.110965

E. Adam Paxton , Jiejie Wu , Tim Hicks , Slimane Doudou , David Applegate , Robert Mason , Andrew Price , Liam Payne

{"title":"A computational framework to support probabilistic criticality modelling for the geological disposal of radioactive waste","authors":"E. Adam Paxton , Jiejie Wu , Tim Hicks , Slimane Doudou , David Applegate , Robert Mason , Andrew Price , Liam Payne","doi":"10.1016/j.anucene.2024.110965","DOIUrl":"10.1016/j.anucene.2024.110965","url":null,"abstract":"<div><div>Nuclear Waste Services is tasked with disposal of the UK’s higher-activity radioactive waste in a Geological Disposal Facility. The disposal of fissile nuclides requires a demonstration that there is no significant concern from criticality, i.e. a fission chain reaction. While waste packages will initially be emplaced in a subcritical configuration, over the long timescales following closure there is potential for waste packages to degrade and for nuclides to be dispersed in the subsurface by groundwater, leading to the potential for a critical system forming. To facilitate modelling, a codebase has been developed which interfaces a probabilistic simulation tool (GoldSim) with a neutron transport code (MONK/MCNP). This allows large ensemble simulations to be run iteratively to determine limiting fissile masses which satisfy a criticality safety criterion. This paper documents the main algorithms and methodologies implemented within this framework, and provides background and example results illustrating the application to post-closure criticality modelling.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142526942","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Development of a friction factor correlation for a foam flow in a horizontal circular pipe 开发水平圆管中泡沫流的摩擦因数相关性

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2024-10-21 DOI: 10.1016/j.anucene.2024.110996

Hyoin Lee , Jaedeok Ko , Ji Hwan Jeong

引用次数: 0