Progress in Nuclear Energy最新文献

LES study on thermal stratification characteristics of lead-cooled fast reactor 铅冷快堆热分层特性的LES研究

IF 3.2 3区工程技术

Progress in Nuclear Energy Pub Date : 2025-10-04 DOI: 10.1016/j.pnucene.2025.106067

Wang Yue , Liu Xuanming , Li Fengchen , Meng Shuqi , Mao Yulong , Li Qian , Cai Weihua

{"title":"LES study on thermal stratification characteristics of lead-cooled fast reactor","authors":"Wang Yue , Liu Xuanming , Li Fengchen , Meng Shuqi , Mao Yulong , Li Qian , Cai Weihua","doi":"10.1016/j.pnucene.2025.106067","DOIUrl":"10.1016/j.pnucene.2025.106067","url":null,"abstract":"<div><div>Lead-cooled fast reactor (LFR) is a fourth-generation nuclear energy system with great potential for development, but thermal delamination occurs after an emergency shutdown, which generates significant thermal stress on solid surface, thus endangering the reactor safety. In this paper, a simplified 1/6 upper plenum model is established and computational fluid dynamics software STAR-CCM+ is used to perform large-eddy simulation (LES) for the reactor shutdown, and the LES accuracy is verified based on experimental data. The results show that, in the steady-state operation condition, holes in the inner cylinder are not enough to affect the Lead-Bismuth Eutectic (LBE) flow in the upper plenum; after the shutdown, thermal stratification is gradually formed in 120 s and rises slowly with time, and the larger the outlet velocity of the reactor is, the faster the interface of thermal stratification rises, in which holes in the inner cylinder can significantly affect the LBE flow and slow down the rising rate of thermal stratification. The decrease of outlet velocity increases the non-uniformity velocity coefficient and thermal stratification number, thus affecting the formation and development of thermal stratification. It is of great significance to study the dynamic evolution of thermal stratification in the upper plenum after the reactor shutdown, so as to evaluate the reactor safety.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"191 ","pages":"Article 106067"},"PeriodicalIF":3.2,"publicationDate":"2025-10-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145220489","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

CFD analysis for optimization of aerodynamic barriers for severe accident consequence mitigation at a nuclear power plant 某核电站严重事故后果缓解气动屏障优化CFD分析

IF 3.2 3区工程技术

Progress in Nuclear Energy Pub Date : 2025-10-03 DOI: 10.1016/j.pnucene.2025.106063

Sana Ullah , Man-Sung Yim

{"title":"CFD analysis for optimization of aerodynamic barriers for severe accident consequence mitigation at a nuclear power plant","authors":"Sana Ullah , Man-Sung Yim","doi":"10.1016/j.pnucene.2025.106063","DOIUrl":"10.1016/j.pnucene.2025.106063","url":null,"abstract":"<div><div>One of the key post-Fukushima developments in nuclear safety is consideration of post-accident consequence mitigation to minimize the radiological consequences of a nuclear power plant severe accident. In our previous study, a conceptual approach based on aerodynamic barriers was successfully examined to confine and control the dispersion of fission products following a containment breach during a severe accident. This approach used a vortex-like air circulation within a defined boundary around the reactor containment with the induced flow directing the released radioactive aerosols toward strategically placed sanction intakes. To support practical implementation of the proposed aerodynamic barriers approach, this study investigated optimal configuration of the aerodynamic barriers using CFD analysis with respect to variations in environmental and accident conditions, and provide robust performance in capturing radioactive aerosols. The CFD analyses were based on coupled Euler–Lagrange method using OpenFOAM and simulated the release and transport of CsI as representative form of fission products under the influence of aerodynamic barriers. The results showed that controlling aerodynamic barrier installation distance and momentum ratio is very important to ensure radioactive aerosols capture. The results indicated that dynamic adjustment of aerodynamic barrier discharge speed and angle is important to handle changes in wind speeds. The results also indicated that successful air flow confinement and radioactive aerosol capture can be achieved with proper control of these key variables (e.g., maintaining the momentum ratio between 1 and 15) while showing minor impact of wind direction variations on the barrier performance.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"191 ","pages":"Article 106063"},"PeriodicalIF":3.2,"publicationDate":"2025-10-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145220490","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

Study on the condensation oscillation characteristics of steam-seawater low-velocity jet 蒸汽-海水低速射流凝结振荡特性研究

IF 3.2 3区工程技术

Progress in Nuclear Energy Pub Date : 2025-10-02 DOI: 10.1016/j.pnucene.2025.106072

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

{"title":"Study on the condensation oscillation characteristics of steam-seawater low-velocity jet","authors":"An Cao , Yongzhen Hua , Xinxing Liu , Zhaoming Meng , Guangming Fan","doi":"10.1016/j.pnucene.2025.106072","DOIUrl":"10.1016/j.pnucene.2025.106072","url":null,"abstract":"<div><div>Due to the similarity between the physical properties of seawater and freshwater, most scholars believe that the condensation oscillation characteristics of steam-seawater low-velocity jet are the same as those of steam-freshwater low-velocity jet, which leads to a lack of accurate understanding of the oscillation characteristics of steam-seawater jet. To explore the oscillation characteristics of steam-seawater low-velocity jet, a high-speed cameraand two high-frequency dynamic pressure sensorswere carried out to carry out a comparative study experiment between the steam-seawater low-velocity jet and the steam-freshwater low-velocity jet. Through experimental comparative analysis, it is found that the steam-seawater low-velocity jet and the steam-freshwater low-velocity jet have both similarities and obvious differences in phase interface structure, pressure oscillation intensity, pressure oscillation amplitude, and pressure oscillation spectrum. In particular, the amplitude of pressure oscillations in the pool is much smaller for steam-seawater low-velocity jet than for steam-freshwater low-velocity jet at higher water temperatures. The results of this study are helpful for understanding and mastering the condensation oscillation characteristics of steam-seawater low-velocity jet.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"191 ","pages":"Article 106072"},"PeriodicalIF":3.2,"publicationDate":"2025-10-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145220491","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

Study on the sliding, collision, and rocking behavior of free-standing spent fuel storage and transport baskets under seismic loading 地震荷载作用下独立式乏燃料储运篮滑动、碰撞和摇摆特性研究

IF 3.2 3区工程技术

Progress in Nuclear Energy Pub Date : 2025-09-30 DOI: 10.1016/j.pnucene.2025.106069

Daogang Lu , Yuchao Wang , Fei Zhao , Yu Liu

{"title":"Study on the sliding, collision, and rocking behavior of free-standing spent fuel storage and transport baskets under seismic loading","authors":"Daogang Lu , Yuchao Wang , Fei Zhao , Yu Liu","doi":"10.1016/j.pnucene.2025.106069","DOIUrl":"10.1016/j.pnucene.2025.106069","url":null,"abstract":"<div><div>To improve the efficiency of spent fuel transportation, a new free-standing storage and transport basket (hereafter referred to as the basket) has been developed. This basket uses a 3 × 3 storage cell layout, which makes it slimmer than the traditional 7 × 6 layout of spent fuel storage racks (hereafter referred to as the rack). Consequently, it exhibits more significant rocking motion during earthquakes. To enhance stability, 24 baskets are connected using dovetail grooves to form a 4 × 6 basket group. Since this type of connection has not been previously employed in spent fuel storage, a specialized seismic safety analysis is required. Previous experimental results indicate that under seismic loading, the baskets exhibit rocking, and occasionally the base plates interlock, preventing them from returning to their original positions. However, existing seismic analysis methods fail to accurately simulate the underwater rocking behavior of multiple baskets connected by dovetail grooves. To explore ways to reduce the rocking angle and to investigate the principle of baseplate interlock, a finite element model is developed. The model employs gap-spring elements to simulate collisions between baskets and fuel assemblies, and simplifies the dovetail connections into spring elements. It also represents fluid-structure interaction as an added mass matrix. Numerical results demonstrate good agreement with experimental data, validating the model's reliability. The influence of the vertical position of dovetail grooves on basket rocking angles and sliding displacements is investigated. The research content can provide important reference for reducing the rocking angle of the basket under earthquake and optimizing the design of the basket.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"191 ","pages":"Article 106069"},"PeriodicalIF":3.2,"publicationDate":"2025-09-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145220492","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

Study on the thermohydraulic reaction characteristics of carbon dioxide and liquid lead-bismuth eutectic 二氧化碳与液态铅铋共晶的热液反应特性研究

IF 3.2 3区工程技术

Progress in Nuclear Energy Pub Date : 2025-09-29 DOI: 10.1016/j.pnucene.2025.106064

Haonan Wang, Haicai Lyu, Weihao Xing, Xiaoyang Lun, Fang Liu, Ruixian Liang, Zhangpeng Guo, Fenglei Niu

{"title":"Study on the thermohydraulic reaction characteristics of carbon dioxide and liquid lead-bismuth eutectic","authors":"Haonan Wang, Haicai Lyu, Weihao Xing, Xiaoyang Lun, Fang Liu, Ruixian Liang, Zhangpeng Guo, Fenglei Niu","doi":"10.1016/j.pnucene.2025.106064","DOIUrl":"10.1016/j.pnucene.2025.106064","url":null,"abstract":"<div><div>The integration of a secondary loop supercritical carbon dioxide (CO<sub>2</sub>) Brayton cycle with a primary loop liquid lead-bismuth (Pb-Bi) fast reactor offers promising advantages, including high efficiency and compactness, making it suitable for diverse power supply applications. However, the safety analysis of heat exchanger tube rupture incidents is a critical aspect of nuclear safety design. In the event of such an accident, CO<sub>2</sub> from the secondary loop is introduced into the liquid Pb-Bi pool, raising the question of whether a thermohydraulic reaction occurs between these two substances. This study aims to investigate the potential thermohydraulic reaction between CO<sub>2</sub> and liquid Pb-Bi, by developing an experimental setup for analyzing the reaction characteristics. The experimental results indicate that a chemical reaction indeed occurs between liquid Pb-Bi and CO<sub>2</sub>, resulting in the formation of lead oxide (PbO) and lead carbonate (PbCO<sub>3</sub>), both of which exhibit a loose structure. The quantity of the reaction products is influenced by the gas phase conditions. These findings are significant for predicting incidents involving heat exchanger tube wall failure and for nuclear safety analysis in liquid Pb-Bi-CO<sub>2</sub> reactors.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"191 ","pages":"Article 106064"},"PeriodicalIF":3.2,"publicationDate":"2025-09-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145221221","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

Safety implications of multiple passive safety valve failures in an i-SMR design i-SMR设计中多个被动安全阀失效的安全含义

IF 3.2 3区工程技术

Progress in Nuclear Energy Pub Date : 2025-09-29 DOI: 10.1016/j.pnucene.2025.106065

Chang Hyun Song , Sung Joong Kim

{"title":"Safety implications of multiple passive safety valve failures in an i-SMR design","authors":"Chang Hyun Song , Sung Joong Kim","doi":"10.1016/j.pnucene.2025.106065","DOIUrl":"10.1016/j.pnucene.2025.106065","url":null,"abstract":"<div><div>Small Modular Reactors (SMRs) are being developed as next-generation nuclear options in response to increasing global power demand driven by data centers, artificial intelligence, and industrial electrification. Korea's innovative SMR (i-SMR) is designed as an integral pressurized water reactor with a thermal output of 520 MWth, aiming to achieve a core damage frequency below 1.0E-09 per year. To meet these goals, the i-SMR adopts multiple Passive Safety Systems (PSSs), including the Passive Emergency Core Cooling System (PECCS), which ensures core cooling through depressurization and natural recirculation. While previous studies have validated the effectiveness of PSSs under nominal conditions, their performance under multiple passive valve failures remains underexplored. In particular, the emergency depressurization valves, and emergency recirculation valves are critical components that enable passive cooling. If one or more of these valves fail to function as designed, coolant may be lost rapidly or recirculation hindered, accelerating fuel exposure. In this study, a detailed MELCOR input model of the i-SMR was developed to analyze accident progression behavior under multiple passive valve failures, particularly focusing on PECCS-related valves. MELCOR simulation results revealed that partial valve failures can lead to faster core degradation than complete system failure scenarios. These findings highlight the critical need to incorporate design redundancy and diversity in passive valve components to ensure reliable accident mitigation capabilities in the i-SMR.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"191 ","pages":"Article 106065"},"PeriodicalIF":3.2,"publicationDate":"2025-09-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145220473","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

Evaluation of COBRA-TF prediction performance enhanced by machine learning-based CHF models 基于机器学习的CHF模型增强COBRA-TF预测性能的评价

IF 3.2 3区工程技术

Progress in Nuclear Energy Pub Date : 2025-09-28 DOI: 10.1016/j.pnucene.2025.106066

Congshan Mao, Yue Jin

{"title":"Evaluation of COBRA-TF prediction performance enhanced by machine learning-based CHF models","authors":"Congshan Mao, Yue Jin","doi":"10.1016/j.pnucene.2025.106066","DOIUrl":"10.1016/j.pnucene.2025.106066","url":null,"abstract":"<div><div>This study investigates the integration of machine learning (ML) techniques into the thermal-hydraulic code COBRA-TF as a foundational advancement for predictive modeling in complex multiphase energy systems. Accurate prediction of critical heat flux (CHF) is essential not only for nuclear reactor safety and performance, but also for a wide range of high-efficiency energy systems that rely on advanced heat transfer, including concentrated solar power, supercritical CO<sub>2</sub> cycles, fission and fusion energy systems, and next-generation thermal desalination and storage technologies. Four ML models were developed and assessed, including two pure data-driven artificial neural networks (ANN) with 3- and 4-layer architectures, as well as two physics-informed machine learning (PIML) variants that embed physical constraints derived from the classical Zuber correlation. The models were trained and validated using the NRC CHF dataset and subsequently integrated into the COBRA-TF at the source-code level, enabling efficient and real-time hybrid modeling. Testing across diverse operating conditions demonstrated that all ML-enhanced COBRA-TF models substantially outperformed the legacy tool in both accuracy and stability. The PIML-enhanced models, particularly the PIML-4L model, achieve the lowest mean absolute error (MAE) and mean absolute percentage error (MAPE) among all tested models. While PIML-3L yields the lowest RMSE, PIML-4L performs best in normalized metrics such as rRMSE, indicating better control of extreme deviations. Error-based analysis further reveals that at least 80 % of predictions from all models fall within a ±15 % relative error range. PIML-3L achieves the highest accuracy under strict error tolerances (≤5 %), while PIML-4L performs best in the moderate error range (5–15 %), demonstrating superior robustness. Overall, it was concluded that integrating ML—especially PIML—into COBRA-TF significantly improves its CHF prediction capabilities, with PIML-4L offering the most comprehensive performance gain.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"191 ","pages":"Article 106066"},"PeriodicalIF":3.2,"publicationDate":"2025-09-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145220475","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 rigorous method for multi-scale coupling of pebble bed and fuel pebble in high temperature gas-cooled reactor 高温气冷堆卵石床与燃料卵石多尺度耦合的严谨方法

IF 3.2 3区工程技术

Progress in Nuclear Energy Pub Date : 2025-09-27 DOI: 10.1016/j.pnucene.2025.106055

Haojie Zhang, Ding She, Lei Shi

{"title":"A rigorous method for multi-scale coupling of pebble bed and fuel pebble in high temperature gas-cooled reactor","authors":"Haojie Zhang, Ding She, Lei Shi","doi":"10.1016/j.pnucene.2025.106055","DOIUrl":"10.1016/j.pnucene.2025.106055","url":null,"abstract":"<div><div>The thermal-hydraulic design of high temperature gas-cooled reactors (HTGRs) involves a multi-scale coupling problem between the solid temperature of macroscopic pebble bed and the microscopic fuel pebble temperature. In the traditional TINTE code, the Thin Shell method is employed to address this issue, but it fails to ensure energy conservation across scales. In this paper, rigorous formulations of the temperature and heat capacity relationships between the two scales are derived from an energy conservation perspective. Based on this, an Effective Heat Capacity (EHC) method is proposed. In the EHC method, the macroscopic heat capacity is redefined to ensure consistent internal energy variations between the scales. Therefore, the EHC method can significantly reduce the number of iterations and improve the computational efficiency. Based on the DAYU3D code, some numerical tests have been conducted to demonstrate the advantages of the EHC method. The results show that the EHC approach outperforms the Thin Shell method in terms of convergence and efficiency, with a reduction in computational time of up to 35 %. This study offers a theoretical foundation for optimizing multi-scale computational methods in thermal-hydraulic analysis software for HTGRs.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"191 ","pages":"Article 106055"},"PeriodicalIF":3.2,"publicationDate":"2025-09-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145221268","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

CANFLEX fuel bundle CFD analysis: Evaluation of different levels of modeling detailing CANFLEX燃料束CFD分析：不同级别建模细节的评估

IF 3.2 3区工程技术

Progress in Nuclear Energy Pub Date : 2025-09-27 DOI: 10.1016/j.pnucene.2025.106056

A.G. Moreira , C.S.L. de Carvalho , I.M. Gomide , I.K. Umezu , A.L. Costa

{"title":"CANFLEX fuel bundle CFD analysis: Evaluation of different levels of modeling detailing","authors":"A.G. Moreira , C.S.L. de Carvalho , I.M. Gomide , I.K. Umezu , A.L. Costa","doi":"10.1016/j.pnucene.2025.106056","DOIUrl":"10.1016/j.pnucene.2025.106056","url":null,"abstract":"<div><div>Nuclear energy plays a key role in sustainable development, with Computational Fluid Dynamics (CFD) simulations enhancing reactor safety and performance. In CANDU reactors, accurate fuel bundle modeling is crucial for reliable thermal-hydraulic analysis. While previous studies have investigated 37-rod and CANFLEX bundles, research on the 43-rod CANFLEX bundle in an Advanced CANDU Reactor-700 (ACR-700) remains limited, particularly using intermediate-scale modeling. This study addresses this gap by evaluating two CFD models in the ACR-700. The Porous Model treats the fuel bundle as a porous medium and a porous jump condition is applied in the transition between regions. The Rod Model represents the coolant flow around a simplified fuel bundle with no connecting pads, with end-plates also treated as porous jumps to reduce computational cost. Steady-state simulations were conducted in ANSYS R19.3 using <span><math><mrow><mi>k</mi><mo>−</mo><mi>ω</mi></mrow></math></span> SST turbulence model. The results indicate that the Porous Model captures global pressure drop and coolant temperature evolution while maintaining low computational cost. The Rod Model provides three-dimensional flow and thermal contours, capturing local effects while preserving agreement with design data. Both models showed mesh convergence, with Grid Convergence Index (GCI) analysis confirming the reliability of the Rod Model’s pressure predictions, that showed good agreement with experimental data, serving as an initial validation of the model’s accuracy. The findings contribute to the development of modeling strategies for CANDU reactors.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"191 ","pages":"Article 106056"},"PeriodicalIF":3.2,"publicationDate":"2025-09-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145221220","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

Analysis of Xi'an pulsed reactor large break loss of coolant accident 西安脉冲堆冷却剂大断损事故分析

IF 3.2 3区工程技术

Progress in Nuclear Energy Pub Date : 2025-09-27 DOI: 10.1016/j.pnucene.2025.106054

Sen Chen, Huaqi Li, Lixin Chen, Leitai Shi, Xiaofei Luo, Xiaoyan Tian, Lei Zhu, Da Li

{"title":"Analysis of Xi'an pulsed reactor large break loss of coolant accident","authors":"Sen Chen, Huaqi Li, Lixin Chen, Leitai Shi, Xiaofei Luo, Xiaoyan Tian, Lei Zhu, Da Li","doi":"10.1016/j.pnucene.2025.106054","DOIUrl":"10.1016/j.pnucene.2025.106054","url":null,"abstract":"<div><div>The Xi'an pulsed reactor (XAPR) severe accident caused by large break loss of coolant was modeled and computed by using ISSA-TRIGA program. The response characteristics and results of thermal hydraulic parameters, fuel rod mechanics and radionuclide release in the accident were analyzed. The results implied that the fuel temperature tend to rise firstly and then to decrease after the accident occurred. When the liquid level dropped to the upper grid plate, the core fuel began to be exposed and the fuel rod lost cooling, causing the fuel temperature start to rise. The first ring fuel rod clad was damaged at 1594 s, and the second and third ring fuel rod clad ruptured at 7642 s and 12999 s, respectively. The maximum fuel temperature and cladding temperature were 1462.3 K and 1452.6 K at 10720 s, respectively. After the fuel clad damaged, the nuclides in the cladding gap were released firstly, and part of fuel was oxidation subsequently. However, the average heat release power of the oxidation reaction was relatively small. In the process of the accident, the fuel temperature did not exceed its melting point and the fuel did not occur large-scale failure. The calculation results in this paper could be helpful for understanding the large break loss of coolant accident process of XAPR and provide important guidance for the subsequent accident mitigation.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"191 ","pages":"Article 106054"},"PeriodicalIF":3.2,"publicationDate":"2025-09-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145220474","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