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Development of modular dynamic code for steam power conversion system in sodium-cooled fast reactor

PROGRESS IN NUCLEAR ENERGY 2024年 173卷

作者： Liu, Shuo Jin, Zhao Wang, Chenglong Zhang, Dalin Tian, Wenxi Qiu, Suizheng Yang, Jun Xi An Jiao Tong Univ Sch Nucl Sci & Technol State Key Lab Multiphase Flow Power Engn Shaanxi Key Lab Adv Nucl Energy & Technol Xian 710049 Shaanxi Peoples R China China Inst Atom Energy Beijing 102413 Peoples R China

Sodium -cooled fast reactor (SFR) is one of the most promising Generation IV reactor types. To improve operational safety and economy, SFR requires accurate prediction of the steam power conversion system. Therefore, a one-dimensional thermal -hydraulic analysis code of the steam power conversion system of SFR is developed in this study where each equipment is packaged as a separate module with modular modeling method. JFNK is adopted to solve ODE in this code, which expand the universality of the code. The model is validated by comparing with design results from the Daya Bay nuclear power plant and the BN-600 steam generator. The maximum relative errors for pressure and temperature are 5.44% and 2.40%, respectively. After validation of the code, the steady-state analysis of the SFR steam power conversion circuit is carried out where the flow and heat transfer process of key equipment can be accurately simulated considering a complete steam power conversion cycle. This work provides reference value for the thermal -hydraulic characteristics of steam power conversion system in SFR.

关键词： Sodium -cooled fast reactor (SFR) Steam power conversion system Modular modeling safety analysis code Thermal -hydraulic

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Verification of a self-developed CFD-based multi-physics coupled code MPC-LBE for LBE-cooled reactor

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Nuclear Science and Techniques 2021年第5期32卷 84-100页

作者： Zhi-Xing Gu Qing-Xian Zhang Yi Gu Liang-Quan Ge Guo-Qiang Zeng Mu-Hao Zhang Bao-Jie Nie College of Nuclear Technology and Automation Engineering Chengdu University of TechnologyChengdu 610059China Sino-French Institute of Nuclear Engineering and Technology Sun Yat-Sen UniversityZhuhai 519082China

To perform an integral simulation of a pool-type reactor using CFD code,a multi-physics coupled code MPC-LBE for an LBE-cooled reactor was proposed by integrating a point kinetics model and a fuel pin heat transfer model into self-developed CFD *** code verification,a code-to-code comparison was employed to validate the CFD ***,a typical BT transient benchmark on the LBE-cooled XADS reactor was selected for verification in terms of the integral or system *** on the verification results,it was demonstrated that the MPC-LBE coupled code can perform thermal-hydraulics or safety analyses for analysis for processes involved in LBE-cooled pool-type reactors.

关键词： LBE-cooled pool-type reactor Computational fluid dynamics Multi-physics coupling code safety analysis code Verification

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Simulations of unprotected loss of heat sink and combination of events accidents for a molten salt reactor

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ANNALS OF NUCLEAR ENERGY 2013年第Mar.期53卷 309-319页

作者： Guo, Zhangpeng Zhang, Dalin Xiao, Yao Tian, Wenxi Su, Guanghui Qiu, Suizheng Xi An Jiao Tong Univ State Key Lab Multiphase Flow Power Engn Xian 710049 Peoples R China Xi An Jiao Tong Univ Dept Nucl Sci Technol Xian 710049 Peoples R China

The molten salt reactor (MSR) is one of the Generation IV reactors. The fuel is dissolved in the carrier salt and circulates in the loop. The technologies are different from that in the solid-fuel reactors. In this work, the attention is focused on development of safety analysis tool for the MSRs. A single channel model, a heat transfer model, a heat sink model and a liquid-fuel point kinetic model that takes into account the effect of circulation are employed. The validation of this code is done with the experimental data of pump coastdown and startup in MSRE. The unprotected loss of heat sink (ULOHS), combination of ULOHS and unprotected loss of flow (ULOF) are performed on the Molten Salt Actinide Recycler and Transmuter (MOSART). This work aims to study temperature fluctuation, corresponding power change, effect of flow delayed neutron precursors and temperature reactivity feedback in transient accident to examine the inherent safety design of MOSART. The transient results reveal that the large negative temperature feedback coefficients guarantee MOSART inherent safety and the range of temperature is within the safety margin in case of combination of accidental events. (C) 2012 Elsevier Ltd. All rights reserved.

关键词： Molten salt reactor safety analysis code Unprotected loss of heat sink Combination of events

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KOREAN DEVELOPMENT OF ADVANCED THERMAL-HYDRAULIC codeS FOR WATER REACTORS AND HTGRs: SPACE AND GAMMA

KOREAN DEVELOPMENT OF ADVANCED THERMAL-HYDRAULIC CODES FOR W...

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14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH)

作者： No, Hee Cheon Ha, Sang Jun Kim, Kyung Doo Lim, Hong Sik Lee, Eo Hwak Jin, Hyung Gon Korea Adv Inst Sci & Technol Dept Nucl & Quantum Engn Taejon 305701 South Korea Khalifa Univ Sci Technol & Res KUSTAR Dept Nucl Engn Abu Dhabi U Arab Emirates KHNP Cent Res Inst Safety Technol Off Taejon 305343 South Korea Korea Atom Energy Res Inst Taejon 305353 South Korea

The Korea nuclear industry has been developing the thermal-hydraulic system analysis safety and Performance analysis code (SPACE) and the GAs Multicomponent Mixture analysis (GAMMA) code for safety analysis of pressurized water reactors (PWRs) and high-temperature gas-cooled reactors (HTGRs), respectively. SPACE will replace outdated vendor-supplied codes and will be used for the safety analysis of operating PWRs and for the design of an advanced PWR. SPACE consists of up-to-date physical models of two-phase flow dealing with multidimensional two-fluid, three-fieldflow GAMMA consists of multidimensional governing equations consisting of the basic equations for continuity, momentum conservation, energy conservation of the gas mixture, and mass conservation of n species. GAMMA is based on a porous media model so that thermofluid and chemical reaction behaviors in a multicomponent mixture system and heat transfer within solid components, free and forced convection between a solid and a fluid, and radiative heat transfer between solid surfaces can be dealt with. GAMMA has a two-dimensional helium turbine model based on the throughflow calculation and a coupled neutronics thermal-hydraulic model. Extensive code assessment has been performed for the verification and validation of SPACE and GAMMA.

关键词： SPACE GAMMA safety analysis code

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KOREAN DEVELOPMENT OF ADVANCED THERMAL-HYDRAULIC codeS FOR WATER REACTORS AND HTGRs: SPACE AND GAMMA

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NUCLEAR TECHNOLOGY 2013年第1期181卷 24-43页

关键词： SPACE GAMMA safety analysis code

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Establishment of freezing model for reactor safety analysis

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JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 2006年第10期43卷 1206-1217页

作者： Kamiyama, Kenji Brear, David J. Tobita, Yoshiharu Kondo, Satoru Japan Atom Energy Agcy Oarai Ibaraki 3111392 Japan

A mechanistic simulation of molten core-material relocation is required to reasonably assess consequences of postulated core disruptive accidents (CDAs) in fast reactors (FRs). The dynamics of molten core-material freezing when it is driven into the channels surrounding the core region plays an important role since this affects fuel removal from the core region. Therefore, a mechanistic model for freezing behavior was developed and introduced into the FR safety analysis code, SIMMER-III, in this study. Based on the micro-physics of crystallization, two key assumptions, supercooling of melt in the vicinity of the wall and melt-wall contact resistance due to imperfect contact, were introduced. As a result, encouraging agreement both with measured melt-penetration lengths and freezing modes of UO2 and metals was obtained. Furthermore, in order to reinforce the developed model, a semi-empirical correlation to predict the supercooling temperature was found. The developed model with the new correlation reproduced both stainless steel freezing and alumina freezing.

关键词： core disruptive accident freezing solidification supercooling contact resistance safety analysis code bulk freezing conduction freezing

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Fusion safety research and development in JAERI

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JOURNAL OF FUSION ENERGY 1997年第1-2期16卷 181-187页

作者： Kunugi, T Seki, Y JAPAN ATOM ENERGY RES INST DEPT FUS ENGNNAKAIBARAKI 31101JAPAN

This paper describes the current status and future plans of the fusion safety research and development regarding to the developments of the dust removal system and safety analysis code and the thermofluid experiments in the Japan Atomic Energy Research Institute (JAERI) for a fusion experimental reactor. The containment of the radioactive material is the key to achieve fusion safety. In the event of accidents, the source terms need to be evaluated with sufficient accuracy. Therefore, in JAERI, the dust characterization have been investigated and the dust removal system using electric force has been developed and tested. A safety analysis code including both thermal and plasma transient analyses under the various event sequences has been developed. Moreover, the preliminary experiments of thermofluid transients in the vacuum vessel such as Ingress of Coolant Event (ICE) and Loss of Vacuum Event (LOVA) have been started and the experimental results using preliminary LOVA/ICE apparatus during 1995-1996 are summarized in this paper.

关键词： fusion safety R&D dust removal system safety analysis code ingress of coolant event (ICE) loss of vacuum event (LOVA)

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