Volume 11: Mitigation Strategies for Beyond Design Basis Events最新文献

Simulation of MSLB+SBO Nuclear Accident Drill Sequence of 3rd-Generation Passive Nuclear Power Plant Based on Severe Accident Simulator 基于严重事故模拟器的第三代无源核电站MSLB+SBO核事故演练序列仿真

Volume 11: Mitigation Strategies for Beyond Design Basis Events Pub Date : 2022-08-08 DOI: 10.1115/icone29-90732

Yongyong Yang, Xiong Huang, Wei Wei, Yinyong Tao, Zhengquan Xie, Guoyang Ma

{"title":"Simulation of MSLB+SBO Nuclear Accident Drill Sequence of 3rd-Generation Passive Nuclear Power Plant Based on Severe Accident Simulator","authors":"Yongyong Yang, Xiong Huang, Wei Wei, Yinyong Tao, Zhengquan Xie, Guoyang Ma","doi":"10.1115/icone29-90732","DOIUrl":"https://doi.org/10.1115/icone29-90732","url":null,"abstract":"\u0000 The full scope simulator (FSS) is very important for operator training, but the related SA training can’t be conducted for lack of the severe accident models, so it is urgent to expand the severe accident model on the FSS in nuclear power plant (NPP). The full scope simulator of 3rd-generation passive NPP can perform more abundant and complete accident scenes for operator training and accident drills through coupling MAAP5 models, namely severe accident simulator (SA simulator). In this paper, the severe models (such as core, primary system, containment), the key technologies of coupling MAAP5 software with FSS are described. Then taking the MSLB+SBO as the initial hypothetical condition to simulate the whole process of the severe accident emergency drill scenario in 3rd-generation NPP and make a preliminary sensitivity analysis on the mitigation effect of the mitigation measure for RCS injection flow in the accident drill scenario. The results show that the whole severe accident drill process can be very simulated, the SA simulator can not only provide an effective mean for the simulation of NPP severe accident drill scenarios, but also can be used as an auxiliary tool for the verification of severe accident management guidelines (SAMG) and the optimization of drill scenarios in NPP, which provide some technical support for the safe and stable operation of nuclear power plant.","PeriodicalId":284948,"journal":{"name":"Volume 11: Mitigation Strategies for Beyond Design Basis Events","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"125506920","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

New Opening Strategy of Filtered Containment Venting System 过滤式安全壳排气系统的新开启策略

Volume 11: Mitigation Strategies for Beyond Design Basis Events Pub Date : 2022-08-08 DOI: 10.1115/icone29-92842

Hua Zheng, Shuhong Wei

{"title":"New Opening Strategy of Filtered Containment Venting System","authors":"Hua Zheng, Shuhong Wei","doi":"10.1115/icone29-92842","DOIUrl":"https://doi.org/10.1115/icone29-92842","url":null,"abstract":"\u0000 Filtered containment venting system (FCVS) is used in some nuclear power plants to avoid containment overpressure failure and uncontrolled radioactive release in severe accidents. Currently the opening strategy is mainly based on the containment pressure, such as the design pressure. The containment has a stronger capability. There are various deposition mechanisms or even plugging of aerosol particles passing through potential micro channels within the containment. There are various natural or engineered deposition mechanisms of aerosol particles in the containment. The qualification pressure of FCVS can be increased. If the FCVS is opened later, the total radioactive release into the environment can be decreased, even though the leakage may be increased. A new opening strategy of FCVS is proposed, which is not just based on the containment pressure, but considers the containment capability, various deposition mechanisms or even plugging of aerosol particles passing through potential micro channels within the containment, various natural or engineered deposition mechanisms of aerosol particles in the containment. This strategy uses the radioactive release into the environment and associated radiological consequence as the optimization objective, considering both radioactive release through potential micro channels before venting and radioactive release after venting, and can reduce the radioactive released into the environment as far as possible while maintaining the integrity of the containment.","PeriodicalId":284948,"journal":{"name":"Volume 11: Mitigation Strategies for Beyond Design Basis Events","volume":"68 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"126309951","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Research on Severe Accident Management Guidelines for HPR1000 NPP HPR1000核电站重大事故管理指南研究

Volume 11: Mitigation Strategies for Beyond Design Basis Events Pub Date : 2022-08-08 DOI: 10.1115/icone29-91849

Jiang Pingting, Chen Peng

{"title":"Research on Severe Accident Management Guidelines for HPR1000 NPP","authors":"Jiang Pingting, Chen Peng","doi":"10.1115/icone29-91849","DOIUrl":"https://doi.org/10.1115/icone29-91849","url":null,"abstract":"\u0000 The Severe Accident Management Guideline (SAMG) is an integrated document for a nuclear power plant (NPP) to deal with severe accidents effectively, with the basic objective to maintain the integrity of the containment in both short term and long term as far as possible. After Fukushima Daiichi nuclear accident, insights on severe accident prevention and mitigation are put forward in the process of a newly designed nuclear power plant. Hua-long Pressurised Reactor (HPR1000) is a generation-III NPP in China, with necessary dedicated system and equipment that designed at the beginning during severe accidents, and is also planned to have associated procedures after accidents. The paper focuses on development process of SAMG for HPR1000 nuclear power plant. Firstly, the characteristics of the plant are sorted, with both dedicated equipment required for severe accident mitigation and other equipment. Then the main objective and principles in developing SAMG are defined. After that, the structure of SAMG is discussed and finally determined referred to the latest insights all over the world and also considering the good practice in China. Additionally, diagnosis process and severe accident strategy are described. In relation to this, flexible strategy in HPR1000 NPP is discussed and included in the integrated SAMG.","PeriodicalId":284948,"journal":{"name":"Volume 11: Mitigation Strategies for Beyond Design Basis Events","volume":"125 1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"129671860","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Extension of the IVR Analysis Module in Modeling the Ex-Vessel Water Cooling 船前水冷却建模中IVR分析模块的扩展

Volume 11: Mitigation Strategies for Beyond Design Basis Events Pub Date : 2022-08-08 DOI: 10.1115/icone29-91888

Pengkun Yu, Yuer Zhao, Chen Wang, Xiaoming Yang, Rubing Ma, Yidan Yuan

{"title":"Extension of the IVR Analysis Module in Modeling the Ex-Vessel Water Cooling","authors":"Pengkun Yu, Yuer Zhao, Chen Wang, Xiaoming Yang, Rubing Ma, Yidan Yuan","doi":"10.1115/icone29-91888","DOIUrl":"https://doi.org/10.1115/icone29-91888","url":null,"abstract":"\u0000 In-vessel retention (IVR) through external reactor vessel cooling (ERVC) is an important severe accident mitigation strategy widely adopted in some advanced light water reactors, e.g. AP1000 and HPR1000. The reactor pressure vessel is supposed to be flooded externally by water, such that the decay heat could be efficiently removed and the radioactive corium materials retained inside the vessel. Lumped parameter models are used in engineering design to calculate the melt pool heat transfer and thus assess the effectiveness of IVR. In this methodology, a simple constant temperature is normally assumed for the ex-vessel surface accounting for the external water cooling. Since the RPV and cooling water are coupled in thermal behavior, modelling of the related heat transfer between them may be of interest. Under the PISAA framework, we developed the IVR module for the IVR analysis. In this paper, we further extend the code capability by considering the ex-vessel heat transfer. Possible conditions of the external cooling water e.g. pure liquid phase convection and boiling as well as the corresponding heat transfer correlations would be considered. The newly implemented model is then verified with acceptable results achieved. With this new feature, the coupled heat transfer analysis of the RPV and external cooling water is investigated. Compared with the original treatment with constant temperature boundary conditions, the new treatment with convective heat transfer affects the predictions of outer vessel temperature and residual wall thickness while the influence on the inner vessel temperature is negligible.","PeriodicalId":284948,"journal":{"name":"Volume 11: Mitigation Strategies for Beyond Design Basis Events","volume":"14 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"115303072","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Feasibility Analysis and Demonstration of In-Vessel-Injection in the Early Stage of Severe Accident 重大事故早期血管内注射的可行性分析与论证

Volume 11: Mitigation Strategies for Beyond Design Basis Events Pub Date : 2022-08-08 DOI: 10.1115/icone29-92736

Shaoxiong Xia, D. Zhan, Zijie Wu, Xinhai Zhao, Yiqun Yu

{"title":"Feasibility Analysis and Demonstration of In-Vessel-Injection in the Early Stage of Severe Accident","authors":"Shaoxiong Xia, D. Zhan, Zijie Wu, Xinhai Zhao, Yiqun Yu","doi":"10.1115/icone29-92736","DOIUrl":"https://doi.org/10.1115/icone29-92736","url":null,"abstract":"\u0000 The core of the nuclear reactor will melt during severe accidents, which may damage the integrity of the reactor pressure vessel and containment, and release radioactive materials to the environment. The third-generation pressurized water reactor is equipped with severe accident mitigation systems to prevent the high temperature (3000K) corium from melting through the reactor pressure vessel. The severe accident mitigation system mainly includes: primary depressurization system, reactor pit flooding system, containment combustible system, containment heat removal system, etc.. Benefit from these systems, large release frequency (LRF) is restricted to a low level. However, these systems cannot stop the process of core degradation. If severe accident happens, the core will melt and the whole reactor cannot be reused again, causing irreversible economic losses. In order to improve the economy and safety of nuclear power plants, this paper proposes in-vessel-injection (IVI) in the early stage of severe accident, and proves the effectiveness of this measure. According to the analysis, in-vessel-injection in the early stage of severe accident can prevent the large-scale melting of the core during severe accident, and the risk of hydrogen and source items will be greatly reduced.","PeriodicalId":284948,"journal":{"name":"Volume 11: Mitigation Strategies for Beyond Design Basis Events","volume":"362 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"122854326","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Investigation of Multifarious Zircaloy-Water Oxidation Models on Core Degadation in a Typical SBO Accident 典型SBO事故堆芯降解的多种锆水氧化模型研究

Volume 11: Mitigation Strategies for Beyond Design Basis Events Pub Date : 2022-08-08 DOI: 10.1115/icone29-92396

Hong-Ping Sun, Yuejian Luo, Jian Deng, Ming Zhang, Xiaoli Wu, Lili Liu, Youyou Xu

{"title":"Investigation of Multifarious Zircaloy-Water Oxidation Models on Core Degadation in a Typical SBO Accident","authors":"Hong-Ping Sun, Yuejian Luo, Jian Deng, Ming Zhang, Xiaoli Wu, Lili Liu, Youyou Xu","doi":"10.1115/icone29-92396","DOIUrl":"https://doi.org/10.1115/icone29-92396","url":null,"abstract":"\u0000 Cladding oxidation is a particularly important process which can affect the process of core melting. Synchronously, hydrogen released from zircaloy-water reaction is a threat to the integrity of containment. Up to now, multifarious zircaloy-water oxidation models are applied in different severe accident analysis programs. In order to investigate the influence of different cladding oxidation models on the core degradation process, a station blackout (SBO) accident model of CPR1000 nuclear power plant (NPP) was established and simulated by modifying the cladding oxidation model in this work. Five oxidation relations were chosen for evaluating thermal hydraulic, metal oxidation and core degradation analysis. Thermal hydraulic characteristics of five types remain consistent before core oxidation. For more concerned hydrogen production, the results show that the ratio of the maximum hydrogen production to the minimum is approximately 2.5 times in comparison of the different oxidation models. Four core nodes temperature distributions at different time points are selected to analyze the core node degradation process, which can help readers understand the effect of different models on core degradation more intuitively. The oxidation heat generated is consistent with that of hydrogen production, indicating that the core degradation time in the case of less hydrogen produced is delayed than that of more hydrogen produced. It follows that different oxidation models have great influence on the core oxidation behavior, which is of great significance to understand different oxidation models of zircaloy and the core degradation behavior of severe accident. It is critical to choose the appropriate model to simulate the whole severe accident.","PeriodicalId":284948,"journal":{"name":"Volume 11: Mitigation Strategies for Beyond Design Basis Events","volume":"18 3 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"126054950","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","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 Hydrogen Risk Venting Mitigation Measure of Small Steel Containment During Severe Accident 重大事故中小钢安全壳氢风险排气缓解措施研究

Volume 11: Mitigation Strategies for Beyond Design Basis Events Pub Date : 2022-08-08 DOI: 10.1115/icone29-91836

Z. Zhiqiang, Zhang Hang, Zhang Ming, P. Huanhuan, Xu Youyou, Deng Jian

{"title":"Study on Hydrogen Risk Venting Mitigation Measure of Small Steel Containment During Severe Accident","authors":"Z. Zhiqiang, Zhang Hang, Zhang Ming, P. Huanhuan, Xu Youyou, Deng Jian","doi":"10.1115/icone29-91836","DOIUrl":"https://doi.org/10.1115/icone29-91836","url":null,"abstract":"\u0000 The hydrogen risk in small steel containment is one of the most important safety problems of small modular reactor (SMR), effective measures need to be taken to mitigate the risk of high concentration hydrogen in a small free volume during severe accidents. In the early stage of a severe accident caused by a loss of coolant accident, the pressure in the containment may rise rapidly, probably triggering the immediate venting of the containment, which can protect the integrity of the containment in the stage of pressure rise in the short term. In the later stage of a severe accident, the generation of hydrogen from reactor core may cause hydrogen combustion or even explosion, which may cause the possibility of overpressure in the containment. In this paper, Computational Fluid Dynamics (CFD) GASFLOW-MPI code is used to analyze the effect of containment venting measures on hydrogen risk in small steel containment of SMR during severe accident initiated by direct vessel injection line break loss-of-coolant. The analysis results show that the venting action can effectively depressurize the containment, the larger the area of the venting pipe is, the more obvious the depressurization effect is. However, hydrogen cannot be effectively discharged in the middle and later stage of the accident, so the risk of hydrogen cannot be completely eliminated, and the possibility of flame acceleration in containment cannot be ruled out for the reason that the negative pressure generated by water vapor condensation will suck in the air outside the containment and form a combustible gas cloud again.","PeriodicalId":284948,"journal":{"name":"Volume 11: Mitigation Strategies for Beyond Design Basis Events","volume":"32 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"129878108","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Simulation Research on Real-Time Offsite Dose Radiation of Nuclear Power Plant 核电站实时场外剂量辐射模拟研究

Volume 11: Mitigation Strategies for Beyond Design Basis Events Pub Date : 2022-08-08 DOI: 10.1115/icone29-91568

P. Ou, Zhenghui Wang, Dongyu He, Pingting Jiang

{"title":"Simulation Research on Real-Time Offsite Dose Radiation of Nuclear Power Plant","authors":"P. Ou, Zhenghui Wang, Dongyu He, Pingting Jiang","doi":"10.1115/icone29-91568","DOIUrl":"https://doi.org/10.1115/icone29-91568","url":null,"abstract":"\u0000 With the rapid development of nuclear power, its safety has attracted more and more attention. In the event of a nuclear accident, once the leakage of radioactive material occurs, it will pose a huge threat to the surrounding environment and the economy and safety of human. In this case, the rapid and accurate evaluation of pollutant dispersion and radiation dose of offsite consequences can guide the decision of nuclear emergency, including the public emergency response and rescue operations. It can also provide the best solution of routes for public evacuation and rescue entrance. The limitation of radioactive accident consequences is essential to protect property security and public safety.\u0000 In this paper, a real-time evaluation program for offsite dose for nuclear emergency system is developed. The selected for atmospheric dispersion model is the segmented Gaussian model (SGM). Diffusion, advection, plume rise, dry deposition, wet deposition, decay, etc. are considered and FORTRAN language is used for the development.\u0000 For code-to-code comparison, C3X, the French general offsite dose evaluation program, is chosen as the reference. In the release of changing weather conditions, the simulation results are consistent with C3X, which demonstrates the validity and reliability of program. This paper makes also a nuclear emergency simulation for the changing weather conditions. The results show that the program can precisely calculate the plume nuclides concentration, ground concentration and dose in different time and space. It can also accurately divide the dose area for public emergency guide and rescue operations.","PeriodicalId":284948,"journal":{"name":"Volume 11: Mitigation Strategies for Beyond Design Basis Events","volume":"24 6 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"130282821","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","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 Passive Pulse Cooling Method of Secondary Side in PWR Nuclear Power Plant 压水堆核电站二次侧被动脉冲冷却方法研究

Volume 11: Mitigation Strategies for Beyond Design Basis Events Pub Date : 2022-08-08 DOI: 10.1115/icone29-93049

Zhenrong Wu, Junjun Xu, Wei-jiao Wang, Runcheng Li, Qi Tang

{"title":"Study on Passive Pulse Cooling Method of Secondary Side in PWR Nuclear Power Plant","authors":"Zhenrong Wu, Junjun Xu, Wei-jiao Wang, Runcheng Li, Qi Tang","doi":"10.1115/icone29-93049","DOIUrl":"https://doi.org/10.1115/icone29-93049","url":null,"abstract":"\u0000 Currently, the supplementary strategy of PWR nuclear power unit to deal with the Station Black-Out (SBO) and Total Loss of Feed-water (TLFW) accidents mainly depends on mobile facilities and manual operation. Under the SBO together with TLFW (SBO+TLFW) beyond design basis accident, the self-sustaining time after the loss of active accident mitigation measures (safety injection / auxiliary feed-water / containment spray) of the nuclear power plant is limited, and the operation time left to the emergency team is relatively less. Especially in the case of external disasters such as super typhoon, the reachability of temporary facilities and personnel is challenged. By analyzing the available resources of the secondary side of the unit after SBO together with TLFW accident, this paper puts forward the passive pulse cooling method of the secondary side: the steam generators are divided into power SG(s) and cooling SG(s), the high-pressure steam of the power SG(s) is used as the passive power source, inject the high-temperature deaerated water stored in the deaerator into the released cooling SG(s), and the core is cooled by pulse water injection and intermittent exhaust. Through qualitative analysis of cooling mechanism, operation strategy and operation mode are given to put into services, and the simulation machine and CATHARE model are used for verification test and simulation calculation respectively. The results show that the two calculation trends are basically the same, which can delay core melting time more than 7 hours after SBO+TLFW accident. According to the probabilistic safety analysis (PSA) of a plant, the core damage frequency (CDF) is reduced by about 7%. The research results in this paper will be helpful to optimize TLFW response strategies and avoid putting into “Feed and bleed mode” too early and contribute to provide mitigation measures for delaying core melting under SBO+TLFW accident, and to gain time for putting external mitigation measures into service.","PeriodicalId":284948,"journal":{"name":"Volume 11: Mitigation Strategies for Beyond Design Basis Events","volume":"303 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"124327000","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Sensitivity Analysis on Key Parameters of Severe Accident Source Term of PWR 压水堆重大事故源项关键参数敏感性分析

Volume 11: Mitigation Strategies for Beyond Design Basis Events Pub Date : 2022-08-08 DOI: 10.1115/icone29-91730

P. Ou, Yongzheng Chen, Dongyu He, Peng Chen

{"title":"Sensitivity Analysis on Key Parameters of Severe Accident Source Term of PWR","authors":"P. Ou, Yongzheng Chen, Dongyu He, Peng Chen","doi":"10.1115/icone29-91730","DOIUrl":"https://doi.org/10.1115/icone29-91730","url":null,"abstract":"\u0000 The complicated severe accident phenomena in typical Pressurized Water Reactor (PWR) Generation III may have a strong influence on source term release into environment and radiological consequence. The study on sensitivity analysis is beneficial to the identification of important factors in severe accident source term analysis and the quantification of their impact.\u0000 ASTEC, the integral code of severe accident analysis developed by IRSN, is used to analyze the sensitivity of key parameters of severe accident source term for typical PWR Generation III, with the simulation of safety systems and source term phenomena, in the representative sequence with fast accident progression, Large Break Loss of Coolant Accident (LBLOCA). With the consideration of the design features of typical PWR Generation III and research status of severe accident source term, the key parameters for sensitivity analysis are identified and selected based on the whole process of radionuclides release, including gaseous iodine mass release fraction from primary circuit to containment, silver iodide reaction, dose rate and pH value in sump, washing effect, etc.\u0000 The sensitivity is quantified by iodine release mass to containment, which is one of the most dangerous radionuclides due to its threat to environment and human thyroid after inhalation and ingestion. The gaseous iodine mass release fraction from primary circuit to containment, silver iodide reaction and washing effect are presented in results as the major contributors to the variation of severe accident source term evaluation.","PeriodicalId":284948,"journal":{"name":"Volume 11: Mitigation Strategies for Beyond Design Basis Events","volume":"265 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"115207653","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0