Progress in Nuclear Energy最新文献_第2页

Numerical study on influence of LFR water-cooled reactor vessel cooling system under unprotected loss of heat sink Accident 无保护散热器损失情况下LFR水冷堆容器冷却系统影响的数值研究

IF 3.3 3区工程技术

Progress in Nuclear Energy Pub Date : 2025-07-24 DOI: 10.1016/j.pnucene.2025.105933

Yue Lin , Dalin Zhang , Yuzhuo Cai , Yapeng Liu , Peng Du , Zhixin Niu , Bo Wang , Wenxi Tian , Suizheng Qiu , Guanghui Su

{"title":"Numerical study on influence of LFR water-cooled reactor vessel cooling system under unprotected loss of heat sink Accident","authors":"Yue Lin , Dalin Zhang , Yuzhuo Cai , Yapeng Liu , Peng Du , Zhixin Niu , Bo Wang , Wenxi Tian , Suizheng Qiu , Guanghui Su","doi":"10.1016/j.pnucene.2025.105933","DOIUrl":"10.1016/j.pnucene.2025.105933","url":null,"abstract":"<div><div>If both normal and emergency heat removal systems fail in a nuclear reactor after an accident occurs, safety and integrity will be severely threatened due to decay heat inside reactor. Passive external cooling outside the reactor vessel could be adopted for heat removal. And its specific design and influences are researched here. The Advanced Coupling code for Evaluating Fast spectrum Nuclear reactor Accidents(ACENA) developed by Xi'an Jiaotong University coupling point-kinetics model is used to simulate thermal hydraulics and neutronics phenomenon. STAR-LM is chosen as the research object for its integral air-cooled Reactor Vessel Cooling System(RVCS) design. Based on the schematic, water-cooled RVCS is modeled and analyzed. STAR-LM is modeled first in ACENA and parameters under steady state are compared with design values, validating the applicability and accuracy of related models. An Unprotected Loss of Heat Sink(ULOHS) accident is taken to analyze changing trends of temperatures inside the reactor which are references for following researches on RVCS. Influence factors including water mass flow rate and operating timing of RVCS are researched. According to simulation, a smaller mass flow rate of water injected causes worse cooling effects. But large and continuous water injection needs to avoid reactor power recovering. Additionally, delaying operation timing of RVCS could bring better cooling effects for nuclear reactor when satisfying safety limitation values. It provides a reliable tool for liquid metal fast reactor safety analysis. And certain experiences are obtained for RVCS design in lead-bismuth eutectic cooled fast reactors.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"189 ","pages":"Article 105933"},"PeriodicalIF":3.3,"publicationDate":"2025-07-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144694299","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

Phenomena Identification and Ranking Table (PIRT) for heat pipes 热管现象识别与排序表（PIRT）

IF 3.3 3区工程技术

Progress in Nuclear Energy Pub Date : 2025-07-24 DOI: 10.1016/j.pnucene.2025.105929

Ilyas Yilgor , Mauricio Tano , Katrina M. Sweetland , Joshua E. Hansel , Piyush Sabharwall , Mark H. Anderson , Zachary D. Sellers , Lise Charlot , Jeremy L. Hartvigsen , Victor Petrov , Yinbin Miao , Stephen Bajorek , Tarek Zaki

{"title":"Phenomena Identification and Ranking Table (PIRT) for heat pipes","authors":"Ilyas Yilgor , Mauricio Tano , Katrina M. Sweetland , Joshua E. Hansel , Piyush Sabharwall , Mark H. Anderson , Zachary D. Sellers , Lise Charlot , Jeremy L. Hartvigsen , Victor Petrov , Yinbin Miao , Stephen Bajorek , Tarek Zaki","doi":"10.1016/j.pnucene.2025.105929","DOIUrl":"10.1016/j.pnucene.2025.105929","url":null,"abstract":"<div><div>Heat pipes are advanced passive thermal management devices that utilize phase change and capillary action to achieve efficient heat transfer. However, due to the complexity of the phenomena coupled in heat pipes, including capillary, phase change, turbulence, and compressibility effects, there are high uncertainties in the predictability of their operational regimes and performance. This PIRT exercise, conducted as a collaborative effort involving the Department of Energy (DOE) Microreactor Program (MRP), the Nuclear Regulatory Commission (NRC), and university partners systematically identifies, reviews, and prioritizes critical phenomena affecting the operation of heat pipes based on their importance and knowledge levels. Additional analyses and discussions are provided for phenomena with high importance and low knowledge, such as wick de-wetting, critical heat flux, contact angles, and pressure dynamics. The discussions include recognizing challenges and proposing future research directions for both modeling/simulation and experimental efforts. Additionally, the article addresses phenomena with medium importance and low knowledge that could impact heat pipe operation during non-normal or transient conditions, including frozen startup, laminar to turbulent transition, geyser boiling, wick priming, under-filling conditions, surface roughness of the wick, non-condensable gases trapped in the wick, and the timescales of startup and shutdown. This comprehensive evaluation serves as a valuable resource for guiding future research and development efforts, supporting the successful integration of heat pipes into nuclear reactors, and contributing to the advancement of heat pipe technologies in safety-critical industries.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"189 ","pages":"Article 105929"},"PeriodicalIF":3.3,"publicationDate":"2025-07-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144694300","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

Atomic-scale modeling assessing the impact of defects on the thermal conductivity of UN 原子尺度模型评估缺陷对UN导热性的影响

IF 3.3 3区工程技术

Progress in Nuclear Energy Pub Date : 2025-07-23 DOI: 10.1016/j.pnucene.2025.105923

C.O.T. Galvin , A. Schneider , M. AbdulHameed , B. Beeler , R.W. Grimes

引用次数: 0

Study on the condensation pattern of steam jet under different back pressure conditions 不同背压条件下蒸汽射流凝结规律的研究

IF 3.3 3区工程技术

Progress in Nuclear Energy Pub Date : 2025-07-22 DOI: 10.1016/j.pnucene.2025.105944

An Cao , Xinxing Liu , Yongzhen Hua , Zhaoming Meng , Guangming Fan

{"title":"Study on the condensation pattern of steam jet under different back pressure conditions","authors":"An Cao , Xinxing Liu , Yongzhen Hua , Zhaoming Meng , Guangming Fan","doi":"10.1016/j.pnucene.2025.105944","DOIUrl":"10.1016/j.pnucene.2025.105944","url":null,"abstract":"<div><div>In the practical application of engineering, the phenomenon of steam underwater immersion jet under high back pressure conditions (>0.1 MPa) exists. However, most of the current research on steam immersion jet condensation flow patterns and their boundaries is conducted under atmospheric pressure conditions (0.1 MPa). The applicability of condensation flow patterns and related prediction formulas derived under atmospheric pressure to high back pressure conditions remains to be validated. Based on the experimental research, the condensation pattern of steam underwater immersion jet under different back pressure conditions is identified, and the condensation behavior of the jet under different back pressure conditions is clarified by comparing and analyzing the visual images and dynamic pressure characteristics of different condensation flow patterns, and the influence law of back pressure and other parameters on the condensation flow pattern boundary is mastered. This study fills the gap in the research of steam underwater immersion jet under the condition of high back pressure, which has important scientific research significance and engineering application value.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"189 ","pages":"Article 105944"},"PeriodicalIF":3.3,"publicationDate":"2025-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144679852","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

Classification of radioactive material release timing for emergency preparedness 应急准备放射性物质释放时间分类

IF 3.3 3区工程技术

Progress in Nuclear Energy Pub Date : 2025-07-21 DOI: 10.1016/j.pnucene.2025.105942

Wasin Vechgama , Jaehyun Cho

{"title":"Classification of radioactive material release timing for emergency preparedness","authors":"Wasin Vechgama , Jaehyun Cho","doi":"10.1016/j.pnucene.2025.105942","DOIUrl":"10.1016/j.pnucene.2025.105942","url":null,"abstract":"<div><div>Following the Fukushima accident in 2011, concerns about nuclear accident risk and consequence management and emergency response intensified. For such severe conditions, promptly implementing a proper emergency response and effectively minimizing the impact on the public and environment is essential. To model it efficiently, a number of elements must be taken into account. In particular, the release starting time after the initiating event and the release behavior characteristics of radioactive materials (e.g., Cs-137) are essential for accurate modeling. In the radiological emergency plan, a conservative estimate for the release starting time of 4 h is assigned in the emergency actions for the public. However, for specific NPPs, the characteristics of the release starting times and Cs-137 release points in each accident are different. Therefore, to construct proper radiological emergency plans for specific NPPs, this study aimed to cluster the proper ranges of release starting times and estimate the confidential Cs-137 release points under the uncertainty analysis scheme of the Optimized Power Reactor-1000 (OPR-1000) during a station blackout (SBO) and Total Loss Of Component Cooling Water (TLOCCW) accident using the Dirichlet process mixture clustering approach. As a result, the SBO scenario having the long range of release starting time showed the highest Cs-137 release in the range of 11–19 h which could help provide more flexible emergency actions and timelines. The TLOCCW scenario showed the overall short range of release starting time already agreed with the conservative estimation at 4 h to consider as the priority of the radiological emergency plan and the decision making.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"189 ","pages":"Article 105942"},"PeriodicalIF":3.3,"publicationDate":"2025-07-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144672326","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

A neutronic model to find promising candidates as burnable poisons in fast reactors 在快堆中寻找有希望的候选可燃毒物的中子模型

IF 3.3 3区工程技术

Progress in Nuclear Energy Pub Date : 2025-07-19 DOI: 10.1016/j.pnucene.2025.105940

Roberto Pergreffi , Francesco Lodi , Giacomo Grasso , Alessia Di Francesco

{"title":"A neutronic model to find promising candidates as burnable poisons in fast reactors","authors":"Roberto Pergreffi , Francesco Lodi , Giacomo Grasso , Alessia Di Francesco","doi":"10.1016/j.pnucene.2025.105940","DOIUrl":"10.1016/j.pnucene.2025.105940","url":null,"abstract":"<div><div>Small modular liquid-metal-cooled fast reactors may require dedicated provisions to reach very long core lifetime, or to reduce the reactivity swing if minimizing control requirements is needed. In such cases, the use of burnable poisons may be a necessary route. Unfortunately, to date very few materials are known that may be used as burnable poisons in fast reactors. Goal of this paper is therefore to propose a model by which the expected neutronic behavior of a candidate burnable poison material in fast spectrum can be estimated without the need for extensive neutronic calculations, for which a detailed description of the system in terms of materials and geometry would be required. The neutronic behavior of a material is estimated by considering separately its performance in terms of poisonousness, defined as the ability to provide enough negative reactivity to compensate for the initial excess reactivity of the fuel, from its performance in terms of burnability, defined as the ability to balance in time the reactivity change due to fuel depletion. Although the values calculated by the model equations do not have physical meaning per se, they can be used to compare different materials with each other in order to make a preliminary assessment of their neutronic behavior. The model was first tested with seven candidates (Eu<sub>2</sub>O<sub>3</sub>, Gd<sub>2</sub>O<sub>3</sub>, Dy<sub>2</sub>O<sub>3</sub>, Er<sub>2</sub>O<sub>3</sub>, NpO<sub>2</sub>, AmO<sub>2</sub> and B<sub>4</sub>C), and then assessed against the results obtained from simulations with MCNP6.1 for a lead fast reactor fuel assembly with UO<sub>2</sub> enriched in <sup>235</sup>U at 19.75 wt% in which, one at a time, the various candidates have been added. Despite some small differences from the simulations especially at low burnups, the comparison confirmed the prediction capability of the model up to a poison content of 10 at.%.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"189 ","pages":"Article 105940"},"PeriodicalIF":3.3,"publicationDate":"2025-07-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144662649","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

An approach to validating abnormal event diagnoses with Explainable AI 一种用可解释人工智能验证异常事件诊断的方法

IF 3.3 3区工程技术

Progress in Nuclear Energy Pub Date : 2025-07-19 DOI: 10.1016/j.pnucene.2025.105939

Ji Hyeon Shin, Junyong Bae, Seung Jun Lee

{"title":"An approach to validating abnormal event diagnoses with Explainable AI","authors":"Ji Hyeon Shin, Junyong Bae, Seung Jun Lee","doi":"10.1016/j.pnucene.2025.105939","DOIUrl":"10.1016/j.pnucene.2025.105939","url":null,"abstract":"<div><div>Nuclear power plant (NPP) operators must diagnose abnormal events accurately and promptly to prevent reactor trips and ensure safety. While artificial intelligence (AI) models can support operators in this complex task, their limited explainability often create trust issues. This study proposes a novel approach to validate AI diagnostic results using Explainable AI, specifically Layer-wise Relevance Propagation with epsilon, to identify whether the model's explanations are consistent with the diagnosed event. By introducing a sub-model trained on explanation patterns, the approach can detect potentially incorrect diagnoses where explanations differ from expected patterns. Case studies using NPP simulator data demonstrated that the proposed method successfully identified approximately 80 % of misdiagnosed cases as untrustworthy, effectively reducing operator confusion from model errors. This validation process enhances both NPP safety and operator trust in AI diagnostic systems by providing an independent verification mechanism for model outputs, thereby reducing the risk of operators relying on incorrect diagnostic information. The proposed approach represents a significant step forward in making AI-based diagnostic systems more reliable and trustworthy in safety-critical nuclear applications.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"189 ","pages":"Article 105939"},"PeriodicalIF":3.3,"publicationDate":"2025-07-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144662650","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

Numerical research of flow and heat transfer characteristics of high-Prandtl-number fluoride salt in the pebble bed channel 高普朗特数氟化物盐在卵石床通道内流动和传热特性的数值研究

IF 3.3 3区工程技术

Progress in Nuclear Energy Pub Date : 2025-07-18 DOI: 10.1016/j.pnucene.2025.105936

Limin Liu , Jiaming Zhang , Tenglong Cong , Yao Xiao , Hanyang Gu

{"title":"Numerical research of flow and heat transfer characteristics of high-Prandtl-number fluoride salt in the pebble bed channel","authors":"Limin Liu , Jiaming Zhang , Tenglong Cong , Yao Xiao , Hanyang Gu","doi":"10.1016/j.pnucene.2025.105936","DOIUrl":"10.1016/j.pnucene.2025.105936","url":null,"abstract":"<div><div>Fluoride-salt-cooled High-temperature Reactors (FHRs) offer promising energy options for high-temperature industrial applications and electricity generation. The thermal-hydraulic characteristics of fluoride salt in the pebble bed reactor core are complex due to the molten salt's high Prandtl number and pebble bed structure. The research on it is important for the design and safe operation of the FHR reactor core. This work investigates the flow and heat transfer of fluoride salt in the pebble bed channel through computational fluid dynamics, with the Reynolds number ranging from 50 to 600. The results reveal that the critical Reynolds numbers for the laminar-to-transitional and transitional-to-turbulent flow regimes are 100 and 180, respectively. The evolution of vortex structures significantly influences the flow characteristics, directly impacting the turbulence intensity distribution and, consequently, the flow transition process. In the laminar regime, the narrow gaps between pebbles severely hinder fluid flow, leading to localized hot spots and elevated fluid temperatures in the vicinity of the gaps. Increasing the Reynolds number mitigates this effect, resulting in substantially reduced temperature peaks and a more uniform temperature distribution. Existing pebble-bed flow and heat transfer correlations show considerable deviation from the present numerical results at lower Reynolds numbers, with the new correlations developed based on the simulation data. This work can provide important thermal-hydraulic data and empirical correlations for the FHR reactor design and operation.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"189 ","pages":"Article 105936"},"PeriodicalIF":3.3,"publicationDate":"2025-07-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144662648","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

Experimental study on the heat transfer characteristics of steam-water two-phase flow in a 2×2 wire-wrapped rod bundle 2×2包丝棒束中蒸汽-水两相流换热特性的实验研究

IF 3.3 3区工程技术

Progress in Nuclear Energy Pub Date : 2025-07-18 DOI: 10.1016/j.pnucene.2025.105935

Han Wang , Jinghui Wu , Wukun Zhu , Dajun Fan , Qincheng Bi

{"title":"Experimental study on the heat transfer characteristics of steam-water two-phase flow in a 2×2 wire-wrapped rod bundle","authors":"Han Wang , Jinghui Wu , Wukun Zhu , Dajun Fan , Qincheng Bi","doi":"10.1016/j.pnucene.2025.105935","DOIUrl":"10.1016/j.pnucene.2025.105935","url":null,"abstract":"<div><div>The Supercritical Water-Cooled Reactor (SCWR) is recognized as a promising nuclear reactor design owing to its elevated thermal efficiency and economic advantages. During transient operations or accident scenarios, reactor cores may experience subcritical pressure conditions. Therefore, the heat transfer characteristics of steam-water two-phase flow under subcritical pressure are crucial for the thermal-hydraulic design and safety assessment of SCWRs. The heat transfer characteristics of steam-water two-phase flow in a 2 × 2 wire-wrapped rod bundle were experimentally investigated. The effects of system parameters on heat transfer characteristics were analyzed. As the heat flux increased, the wall temperature in the single-phase region increased accordingly, but the heat transfer coefficient remained unchanged. In the saturated boiling region, the wall temperature stayed relatively constant, but the heat transfer coefficient improved with heat flux. The mass flux had a relatively weak effect, whereas an increase in pressure led to higher values in the wall temperature and heat transfer coefficient within the two-phase region. Several heat transfer correlations were selected for evaluation. It was found that the Kandlikar (1990) correlation showed the highest accuracy for saturated flow boiling. The critical heat flux associated with heat transfer deterioration was also examined. As the pressure increased, the critical heat flux and critical steam quality decreased, suggesting that higher pressure will exacerbate heat transfer deterioration.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"189 ","pages":"Article 105935"},"PeriodicalIF":3.3,"publicationDate":"2025-07-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144654559","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

Numerical simulation and experimental validation of DLOFC transients in a high temperature helium loop embedded with AM geometry 嵌入AM几何结构的高温氦回路中DLOFC瞬态的数值模拟与实验验证

IF 3.3 3区工程技术

Progress in Nuclear Energy Pub Date : 2025-07-17 DOI: 10.1016/j.pnucene.2025.105910

Ketan Ajay , Broderick Sieh , Hitesh Bindra

{"title":"Numerical simulation and experimental validation of DLOFC transients in a high temperature helium loop embedded with AM geometry","authors":"Ketan Ajay , Broderick Sieh , Hitesh Bindra","doi":"10.1016/j.pnucene.2025.105910","DOIUrl":"10.1016/j.pnucene.2025.105910","url":null,"abstract":"<div><div>The Transformational Challenge Reactor (TCR) draws inspiration from High-Temperature Gas-Cooled Reactors (HTGRs) and leverages the advantages of additive manufacturing technology, particularly by incorporating additively manufactured fuel elements. Similar to HTGRs, the TCR incorporates engineered passive safety features designed to effectively remove heat during a Loss of Forced Cooling (LOFC) event. A Depressurized Loss of Forced Cooling (DLOFC) accident is a design basis accident in the TCR, occurring when a breach in the coolant system leads to depressurization. This allows air to enter the core or plenum through mechanisms such as diffusion, gravity currents, or natural circulation. After several hours following the depressurization, the Onset of Natural Circulation (ONC) may occur, driving bulk air ingress into the core. This paper presents a numerical study aimed at investigating the underlying physics and predicting the influence of Additively Manufactured (AM) geometry during air ingress transients. These simulations were performed for two different geometric configurations: one without the AM part, referred to as the base geometry, and the other with the AM part. The test geometry and boundary conditions used in the analysis were obtained from the experiment. Transient fluid temperature profiles and helium concentration distributions were obtained. A thorough comparison between the predicted ONC times and temperatures with the experimental data was conducted. The model results were found to be in good agreement with the temperature distribution and ONC times reported in the experiment.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"189 ","pages":"Article 105910"},"PeriodicalIF":3.3,"publicationDate":"2025-07-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144654558","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