International Journal of Advanced Nuclear Reactor Design and Technology最新文献_第7页

Phenomena identification Ranking Table (PIRT) study for suppression containment of small modular reactor using new methodology 用新方法研究小型模块化反应堆抑制安全壳的现象识别排序表

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2023-06-01 DOI: 10.1016/j.jandt.2023.08.002

Yili Yang , Yuxin Wang , Yusheng Liu , Shengfei Wang

{"title":"Phenomena identification Ranking Table (PIRT) study for suppression containment of small modular reactor using new methodology","authors":"Yili Yang , Yuxin Wang , Yusheng Liu , Shengfei Wang","doi":"10.1016/j.jandt.2023.08.002","DOIUrl":"https://doi.org/10.1016/j.jandt.2023.08.002","url":null,"abstract":"<div><p>The Phenomena Identification and Ranking Table (PIRT) is a significant method for analyzing the safety of nuclear reactors. It helps researchers identify important phenomena within the reactor, enabling a focused and appropriate simplification of accident scenarios during the study. However, traditional PIRT methods often rely on experts' subjective opinions to rank phenomena’ importance and knowledge level, potentially distorting the PIRT results. This paper proposes a new PIRT method inspired by literature evaluation techniques used in the medical and healthcare field, which can be more objective. This new method utilizes a literature evaluation framework instead of relying solely on expert judgments, resulting in a more objective assessment of the phenomena’ importance and knowledge level. This study applies the new method to a simplified small modular reactor with a suppression containment system. Following a Loss of Coolant Accident (LOCA), the suppression containment can effectively suppress temperature and pressure increases, ensuring containment integrity. Relevant PIRT tables and a knowledge-level structure are obtained using the new method.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"5 2","pages":"Pages 104-113"},"PeriodicalIF":0.0,"publicationDate":"2023-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"49748769","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

The modeling of passive containment cooling system and sensitivity analysis using the SPRUCE code 基于SPRUCE规范的被动安全壳冷却系统建模及敏感性分析

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2023-06-01 DOI: 10.1016/j.jandt.2023.06.001

Zehao Zhang, Zichen Zhao, Fangqing Yang, Chao Guo, Peng Chen

{"title":"The modeling of passive containment cooling system and sensitivity analysis using the SPRUCE code","authors":"Zehao Zhang, Zichen Zhao, Fangqing Yang, Chao Guo, Peng Chen","doi":"10.1016/j.jandt.2023.06.001","DOIUrl":"https://doi.org/10.1016/j.jandt.2023.06.001","url":null,"abstract":"<div><p>Passive Containment Cooling Systems (PCCSs) have attracted wide attention to improve the inherent safety of nuclear power plants. This study summarizes a system analysis method of a closed-loop natural circulation of the PCCS via separately evaporation and condensation. A lumped-parameter model is built using SPRUCE, a code of system simulation of severe accidents, to investigate the natural circulation characteristics and heat-removal efficiency of the PCCS. The analysis of heat-removal capacity of the PCCS is carried out under typically thermal conditions of the containment, such as pressure from 0.28 to 0.4 MPa, and the results are validated by comparison with the experimental data obtained from a single-loop testing facility of the PCCS in full scale. The SPRUCE model agrees with the data, despite slightly underestimating the rate of condensation heat transfer. It conservatively satisfies the requirement of design capacity. To further verify the effectiveness of mitigation on containment overpressure, a representative scenario of LBLOCA (Large Break Loss of Coolant Accident) is implemented using the code. Meanwhile, the effects of non-condensable gas are also presented, as a reference for optimization of PCCS design.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"5 2","pages":"Pages 72-76"},"PeriodicalIF":0.0,"publicationDate":"2023-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"49758811","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 iterative algorithm of full plate cross-flow and counter-flow printed circuit heat exchanger for fluoride-salt-cooled high-temperature advanced reactor 氟盐冷却高温先进反应堆全板横流和逆流印刷电路换热器迭代算法研究

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2023-06-01 DOI: 10.1016/j.jandt.2023.07.002

Xinze Li, Dalin Zhang, Xinyu Li, Wenxi Tian, Suizheng Qiu, G.H. Su

{"title":"Study on iterative algorithm of full plate cross-flow and counter-flow printed circuit heat exchanger for fluoride-salt-cooled high-temperature advanced reactor","authors":"Xinze Li, Dalin Zhang, Xinyu Li, Wenxi Tian, Suizheng Qiu, G.H. Su","doi":"10.1016/j.jandt.2023.07.002","DOIUrl":"https://doi.org/10.1016/j.jandt.2023.07.002","url":null,"abstract":"<div><p>In the concept of Fluoride-Salt-cooled high-Temperature Advanced Reactor (FuSTAR), the Printed Circuit Heat Exchanger (PCHE) is mainly considered in its supercritical carbon dioxide (S–CO2) Brayton cycle secondary loop. The design of the PCHE is based on the property of working fluid provided by the Brayton cycle system design. Therefore, a fast and iterative PCHE design method is required. In this study, a mathematical model based on Picard's iterative method is proposed. The model can solve full plate temperature field with complex boundary conditions and initial values including cross-flow and counter-flow PCHE in parallel. Based on FuSTAR Brayton cycle design conditions, three sizes of cross-flow and counter-flow PCHEs are simulated respectively by the mathematical model. Compared with Computational Fluid Dynamics (CFD) models, a one and three-dimensional coupling model and a three-dimensional model, the mathematical model reduced the computing time from 1 to 10 h to 1–10 s, and the maximum temperature mean deviation is 18.6% for the smallest size case because of the edge effect and local disturbance, but 8.4% for other size. Taking secondary loop recuperator design as example, the computing speed and accuracy of the mathematical model can meet the design requirements of FuSTAR PCHE.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"5 2","pages":"Pages 86-96"},"PeriodicalIF":0.0,"publicationDate":"2023-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"49758812","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: 1

Advanced nuclear power engine: A brief overview of gas core reactor for space exploration 先进核动力发动机：用于太空探索的气芯反应堆概述

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2023-06-01 DOI: 10.1016/j.jandt.2023.05.001

Yebing Zhang

{"title":"Advanced nuclear power engine: A brief overview of gas core reactor for space exploration","authors":"Yebing Zhang","doi":"10.1016/j.jandt.2023.05.001","DOIUrl":"https://doi.org/10.1016/j.jandt.2023.05.001","url":null,"abstract":"<div><p>Space travel requires propulsion with high specific impulse. Gas core reactors using plasma nuclear fuel operate at high temperature (<span><math><mrow><msup><mn>10</mn><mn>4</mn></msup></mrow></math></span>-<span><math><mrow><msup><mn>10</mn><mn>5</mn></msup><mspace></mspace><mi>K</mi></mrow></math></span>) and theoretically produce higher specific impulses (<span><math><mrow><mn>2500</mn></mrow></math></span>-<span><math><mrow><mn>7000</mn><mspace></mspace><mi>s</mi></mrow></math></span>) than solid core nuclear thermal rockets. Ground-based reactors have higher total energy conversion efficiency (<span><math><mrow><mn>70</mn><mo>%</mo></mrow></math></span>). It also can exhaust all actinides and have negative density coefficients of reactivity. This paper reviews the evolution of gas core reactors for space exploration in the United States and the Soviet Union, including nuclear light bulb, open-cycle gas core nuclear rockets and gas core reactors for power generation. In terms of reactor physics, the selected materials and selection criteria for fuel and moderator-reflector are compiled, and the available neutron analysis methods are summarized. In terms of reactor physics, fluid thermal properties, radiation heat transfer models and thermal protection methods are briefly introduced. Fuel loss as the representative challenge is analyzed for the main causes of generation. In this paper, the principles, related studies, and advantages and disadvantages of four fuel confinements are reviewed. Finally, the principles and start-up process of the two start-up methods are discussed. Start-up is another challenge for this reactor design. Early this century, gas core reactor research stagnated due to the lack of thermal property data and the ability of high-temperature hydrodynamics simulation. Nowadays, with the increase of computing power and the breakthrough of computational fluid dynamics, these challenges are expected to be overcome.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"5 2","pages":"Pages 53-71"},"PeriodicalIF":0.0,"publicationDate":"2023-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"49748066","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: 1

Analysis of containment pressure control strategy in HPR1000 NPP under severe accidents HPR1000核电厂严重事故下安全壳压力控制策略分析

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2023-06-01 DOI: 10.1016/j.jandt.2023.08.001

Pingting Jiang , Yunna An , Wenxi Tian , Peng Chen , Dongyu He , Pingwen Ou , Deyang Xu

{"title":"Analysis of containment pressure control strategy in HPR1000 NPP under severe accidents","authors":"Pingting Jiang , Yunna An , Wenxi Tian , Peng Chen , Dongyu He , Pingwen Ou , Deyang Xu","doi":"10.1016/j.jandt.2023.08.001","DOIUrl":"https://doi.org/10.1016/j.jandt.2023.08.001","url":null,"abstract":"<div><p>Containment is the last barrier of preventing the release of radioactive fission products in a nuclear power plant (NPP). It has the top priority of its strategy in a severe accident (SA) to ensure the integrity of containment. Generally, there are two ways for the containment heat removal. One is to set exchangers or sprays to cool the atmosphere in the containment like CPR1000. The other is to set sprays out of the steel containment to remove heat like AP1000. After Fukushima Daiichi nuclear accident, mitigation strategies after severe accidents are focused and specific systems of dealing with containment failure threat are required to design in new built NPPs. HPR1000 is a generation-Ⅲ PWR in China, which deployed the dedicated severe accident (SA) system of containment spray to address the above conditions. Containment spray in HPR1000 has two identical trains isolated physically, and each train is capable to reduce containment pressure after severe accidents. The containment spray system cannot start automatically, but only be started by operator during severe accidents. According to the lessons from Fukushima accident, it is hard for the operator to make the right choice in such a high-pressure environment during severe accidents, so the proper start-up time is better given in advance as possible. This paper assesses the effectiveness of the containment spray, conducts sensitive calculations of different start-up time, and discuss the negative effects of containment spray. Based on the calculation results, insights of containment spray strategy are gained for HPR1000 NPP and the proper start-up time for the strategy of containment spray in SAMG are put forward.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"5 2","pages":"Pages 97-103"},"PeriodicalIF":0.0,"publicationDate":"2023-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"49748516","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

Prediction of heat transfer coefficients for steam condensation in the presence of air based on ANN method 基于神经网络的空气存在下蒸汽冷凝换热系数预测

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2023-06-01 DOI: 10.1016/j.jandt.2023.07.001

Haoran Cao , Boyang Cao , Congyi Xia , Zhaoming Meng , Haozhi Bian , Ming Ding

{"title":"Prediction of heat transfer coefficients for steam condensation in the presence of air based on ANN method","authors":"Haoran Cao , Boyang Cao , Congyi Xia , Zhaoming Meng , Haozhi Bian , Ming Ding","doi":"10.1016/j.jandt.2023.07.001","DOIUrl":"https://doi.org/10.1016/j.jandt.2023.07.001","url":null,"abstract":"<div><p>Artificial neural network (ANN) methods have been gradually used in the field of nuclear reactor thermal-hydraulics as new methods to improve accuracy or fast prediction. This study establishes a back propagation (BP) neural network model based on the ANN methods to predict the steam condensation heat transfer coefficient outside a heat tube in the presence of air. The main factors affecting condensing heat transfer, such as pressure, air mass fraction, subcooling, and tube diameter, are used as input quantities, and the condensation heat transfer coefficient is used as output quantity. A complete set of neural networks for predicting the heat transfer coefficient for steam condensation in the presence of air is established based on the relevant experimental data collected over the world. The results predicted by the ANN model are compared with the experimental data and those of traditional correlation methods. The data from 2276 experiments are distributed within a ±10% error band at the 95% confidence level. This means that the prediction accuracy of the ANN model is higher than that of the traditional experimental correlation. Therefore, the neural network model developed in this study can be used for the prediction of the heat transfer coefficient for steam condensation in the presence of air.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"5 2","pages":"Pages 77-85"},"PeriodicalIF":0.0,"publicationDate":"2023-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"49748591","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

The evaporation of nanoscale sodium liquid film on the non-ideal nanostructure surface: A molecular dynamics study 纳米钠液膜在非理想纳米结构表面的蒸发：分子动力学研究

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2023-03-01 DOI: 10.1016/j.jandt.2023.02.002

Zetao Wang, Tianzhou Ye, Kailun Guo, Wenxi Tian, Suizheng Qiu, Guanghui Su

{"title":"The evaporation of nanoscale sodium liquid film on the non-ideal nanostructure surface: A molecular dynamics study","authors":"Zetao Wang, Tianzhou Ye, Kailun Guo, Wenxi Tian, Suizheng Qiu, Guanghui Su","doi":"10.1016/j.jandt.2023.02.002","DOIUrl":"https://doi.org/10.1016/j.jandt.2023.02.002","url":null,"abstract":"<div><p>The nanoscale liquid sodium film inside the microporous wick structure is of great importance to understanding the evaporation mechanism of the sodium heat pipe. The novel optimized wick structure is made of several layers of special screen. The surface of each screen exhibits a nanostructure type. Some non-ideal nanostructures may result from experimental faults or limits. And they will have an effect on the evaporation of film. In the present study, molecular dynamics is adopted to investigate this effect. The simulation system consists of the liquid sodium film and the solid surface. The flat surface is set as the reference. Based on the three non-ideal shapes of deposition, the sinusoidal nanostructures, conical nanostructures, and spherical nanostructures are built. The results indicate that the evaporation is suppressed by the above nanostructure surfaces. The weakening effect is through three forms: the potential gradient of the liquid film is intensified and the evaporation difficulty is increased; the heat transfer in the solid-liquid contact region is impeded; the collision heat transfer inside the liquid film is affected due to the delay of the aggregation variation between liquid atoms.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"5 1","pages":"Pages 1-8"},"PeriodicalIF":0.0,"publicationDate":"2023-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"49748644","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: 2

Analysis of radiation shielding effectiveness of hydride and borohydride metals for nuclear industry 核工业用氢化物和硼氢化物金属的辐射屏蔽效能分析

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2023-03-01 DOI: 10.1016/j.jandt.2023.04.001

A.M. Osman

{"title":"Analysis of radiation shielding effectiveness of hydride and borohydride metals for nuclear industry","authors":"A.M. Osman","doi":"10.1016/j.jandt.2023.04.001","DOIUrl":"https://doi.org/10.1016/j.jandt.2023.04.001","url":null,"abstract":"<div><p>Hydrogenous materials are of great interest in nuclear industry because of the ability of hydrogen to slow-down neutrons. In comparison with common hydrogenous materials, metal hydrides highly enriched with hydrogen at high temperatures and good mechanical properties. This study concerns on evaluation the requirements for photons, charged particles and fast neutrons shielding parameters for several types of hydride and borohydride metals. Firstly, photons shielding ability of these materials were examined using Phy-X/PSD program. Assigning the attenuation factors of photons were evaluated for an energy range (1 keV–100 GeV). Secondly, the charged particle (electron, proton and alpha particle) interactions were examined by using ESTAR (Stopping Powers and Ranges for Electrons) and SRIM (The Stopping and Range of Ions in Matter) programs. Finally, fast neutron attenuation was tested by calculating the removal cross-sections (Σ<sub>R</sub>) for neutron with energy 4.5 MeV. The dependence of photons, charged particles and fast neutrons parameters on the constituent of tested samples was given and discussed. The obtained results through this work show that, the shielding proficiency parameters depend on the energy of the incident radiation and chemical constituents of the examined materials. As well as, these results showed that BaH<sub>2</sub>, ZrH<sub>2</sub>, VH<sub>2</sub> and TiH<sub>2</sub> samples own a good shielding performance compared to the other investigated samples.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"5 1","pages":"Pages 30-43"},"PeriodicalIF":0.0,"publicationDate":"2023-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"49748064","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 criticality assessment through user's routines in FLUKA 通过用户在FLUKA中的例程进行关键性评估

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2023-03-01 DOI: 10.1016/j.jandt.2023.02.001

F. Vanoni, E. Padovani, A. Porta, F. Campi, R. Chebac

引用次数: 0

Numerical investigation on transverse flow of helical cruciform fuel rod assembly in a lead-bismuth cooled fast reactor 铅铋冷快堆螺旋十字形燃料棒组件横向流动数值研究

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2023-03-01 DOI: 10.1016/j.jandt.2023.03.001

Yue Zeng, Huandong Chen, Pingjian Ming

引用次数: 0