Nuclear Science and Technology Open Research最新文献

Concept design, application and optimization of operational parameters for new forced safety injection tank system 新型强制安全喷射罐系统的概念设计、应用和运行参数优化

Nuclear Science and Technology Open Research Pub Date : 2024-07-08 DOI: 10.12688/nuclscitechnolopenres.17546.1

Asad Ullah Amin Shah, Junyung Kim, Ali Mansoor, H. Kang

{"title":"Concept design, application and optimization of operational parameters for new forced safety injection tank system","authors":"Asad Ullah Amin Shah, Junyung Kim, Ali Mansoor, H. Kang","doi":"10.12688/nuclscitechnolopenres.17546.1","DOIUrl":"https://doi.org/10.12688/nuclscitechnolopenres.17546.1","url":null,"abstract":"Background The engineered safety systems are designed to execute fundamental safety functions encompassing reactivity confinement, reactivity control and decay heat removal. Failure of any one of these functions can result in severe accident conditions. Passive systems have been implemented as a better option for plant safety. This research proposes a modification of existing safety injection tanks to the new forced safety injection tanks (FSITs) that utilize the backpressure from the pressurizer or steam generator to drive coolant into the reactor core under high pressure conditions. Methods FSITs aim to extend the coping-time during accidents like Station Blackout, providing an extended timing window for deployment of FLEX systems. The mathematical design is proposed and implemented into the thermal-hydraulic input of a nuclear power plant to demonstrate the system’s applicability that was demonstrated by leveraging the conventional PRA approach and risk quantification. Results The suggested system useful when used in the accident scenario of Station black out in-coincident with turbine-driven pumps fail to run. As the proposed design is a conceptual design, the optimization of associated operational parameters and set points is necessary. This optimization is performed using the risk analysis and virtual control environment-based Dynamic Probabilistic Risk Assessment framework. The method and findings of this study affirm that the coping-time for Station Blackout can be significantly extended, ensuring a substantial margin for the effective deployment of the FLEX. Conclusions A concept design of a passive forced safety injection system was suggested and demonstrated by integrating the mathematical model into a thermal-hydraulic model of a nuclear power plant. The parameters were optimized and the results demonstrated that the new system was effective in recovering the nuclear power plant from the accidents such as SBO with turbine-driven pumps fail to operate.","PeriodicalId":516082,"journal":{"name":"Nuclear Science and Technology Open Research","volume":"4 7","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-07-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141667974","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

Influence of neutron decay on nuclear reactor materials 中子衰变对核反应堆材料的影响

Nuclear Science and Technology Open Research Pub Date : 2024-06-07 DOI: 10.12688/nuclscitechnolopenres.17547.1

J. C. Philipps, Jeff Terry

引用次数: 0

CFD model for predicting spent nuclear fuel drying conditions 预测乏核燃料干燥条件的 CFD 模型

Nuclear Science and Technology Open Research Pub Date : 2024-06-06 DOI: 10.12688/nuclscitechnolopenres.17492.1

Fakhrul Islam, Nathaniel Cooper, Alexander W. Abboud, Tanvir Farouk, Jamil Khan, D. Guillen, Travis Knight

{"title":"CFD model for predicting spent nuclear fuel drying conditions","authors":"Fakhrul Islam, Nathaniel Cooper, Alexander W. Abboud, Tanvir Farouk, Jamil Khan, D. Guillen, Travis Knight","doi":"10.12688/nuclscitechnolopenres.17492.1","DOIUrl":"https://doi.org/10.12688/nuclscitechnolopenres.17492.1","url":null,"abstract":"Background A significant quantity of aluminum-clad spent nuclear fuel (ASNF) is currently managed by the United States (U.S.) Department of Energy (DOE) at several sites in the U.S. Much of this fuel develops a corrosion layer with adsorbed water either through pre-corrosion, in-reactor, or in wet storage depending on the source. Gamma radiation from the fuel after removal from the reactor can cause evolution of hydrogen from this corrosion layer. For long-term safety it is desirable to remove the adsorbed water from the fuel surface to limit the production of hydrogen gases when the fuel is placed in road-ready sealed canisters for eventual final disposition. Methods This study presents a series of experiments showing the effectiveness of both forced helium drying and vacuum drying methods for the removal of water from surrogate fuel assemblies. A computational fluid dynamics (CFD) model is constructed to replicate the drying processes and simulate the experiments. Results The model suggests that increasing the inlet temperature is a more efficient way of improving the efficacy of a drying process. An increase in inlet temperature leads to a reduction in drying time 28–55% greater than the same relative increase in flow rate. One of the assemblies inside the canister consistently experiences the least amount of turbulent flow and require the longest time to dry. Therefore, the assembly with the highest decay heat should be placed in this slot in the basket in order to facilitate the drying of this assembly. Conclusions The validated CFD model can be used in the future as a design and analysis tool for a full-scale drying setup with actual fuel assemblies.","PeriodicalId":516082,"journal":{"name":"Nuclear Science and Technology Open Research","volume":"50 s250","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-06-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141377393","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

Decision-making based on Markov decision process in integrated artificial reasoning framework—Part 2: Applications 综合人工推理框架中基于马尔可夫决策过程的决策制定--第 2 部分：应用

Nuclear Science and Technology Open Research Pub Date : 2024-06-06 DOI: 10.12688/nuclscitechnolopenres.17490.1

Kyle Warns, Junyung Kim, Xinyan Wang, Xingang Zhao, Birdy Phathanapirom, Michael W. Golay, Hyun Kang

{"title":"Decision-making based on Markov decision process in integrated artificial reasoning framework—Part 2: Applications","authors":"Kyle Warns, Junyung Kim, Xinyan Wang, Xingang Zhao, Birdy Phathanapirom, Michael W. Golay, Hyun Kang","doi":"10.12688/nuclscitechnolopenres.17490.1","DOIUrl":"https://doi.org/10.12688/nuclscitechnolopenres.17490.1","url":null,"abstract":"To address the challenge of interpreting the decisions made by machine learning-based control algorithms, the Integrated Artificial Reasoning Framework (IARF) introduced in Part I of this work was developed. Further implementing this framework within a Markov Decision Process (MDP) yielded a decision-making system which produces optimal decisions whose logic can be easily traced and understood. The IARF with MDP formulated in Part I of this study was applied to two case studies of relevance to the nuclear power community: 1) operational decision-making while experiencing the degradation of key components, and 2) optimal control of a nuclear power and hydrogen co-generation facility. This work details the steps of implementing and analyzing these two studies, beginning with the creation of simulation models using the Modelica programming language, covering the formulation of the MDP using the IARF, and concluding with the execution and analysis of the scenarios. Technical challenges related to implementation, such as maintaining physically understandable traceability and efficiently producing large amount of data for calculating state transition probabilities, are addressed and their resolutions are suggested. Emphasis is placed on the benefits of relying on a replaceable modeling framework for simulating system behavior. Results indicate that optimal operational trajectories were successfully found by the MDP, and through utilization of the IARF, these results are easily interpretable and understandable to any user.","PeriodicalId":516082,"journal":{"name":"Nuclear Science and Technology Open Research","volume":"10 2","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-06-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141378465","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

A general-purpose code for correlated sampling using batch statistics with MCNP6 for fixed-source problems 利用 MCNP6 对固定源问题进行批量统计的相关采样通用代码

Nuclear Science and Technology Open Research Pub Date : 2024-03-27 DOI: 10.12688/nuclscitechnolopenres.17496.1

J. Favorite

引用次数: 0

The Bimodal Epithermal Astronuclear Reactor (BEAR): A LONG-LIVED, Universal Space Nuclear Powerplant 双模外热天体核反应堆 (BEAR)：长寿命通用空间核动力装置

Nuclear Science and Technology Open Research Pub Date : 2024-02-01 DOI: 10.12688/nuclscitechnolopenres.17481.1

Teyen Widdicombe, J. Herring, Brad Kirkwood

引用次数: 0