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Some fundamental understandings of Zn-injection water chemistry on material corrosion in pressurized water reactor primary circuit

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Corrosion Communications 2022年第2期6卷 52-61页

作者： Xinqiang Wu Xiahe Liu Ziyu Zhang Jibo Tan En-Hou Han Wei Ke CAS Key Laboratory of Nuclear Materials and Safety Assessment Liaoning Key Laboratory for Safety and Assessment Technique of Nuclear Materials Institute of Metal Research Chinese Academy of Sciences

The optimization of water chemistry in pressurized water reactor(PWR) primary circuit is one of the most effective ways to achieve both safety and economy for operating PWR nuclear power plants(NPPs). Special attention has been paid to fundamental research and engineering application of Zn-injection water chemistry(ZWC) in PWR primary circuit in recent years in China due to the rapid development of nuclear power. The present paper mainly reports the status of PWR NPPs in China and some fundamental understandings of ZWC on material corrosion. Effects of temperature(T), pH at T value, Zn concentration, Zn-injection sequence and lasting time on electrochemical corrosion behavior, oxide film characteristic and stress corrosion cracking(SCC) susceptibility of nuclear-grade austenitic alloys were investigated. A modified point defect model was used to discuss the effects of ZWC on the oxide films in high-temperature pressurized water. Some water chemistry parameters in PWR primary coolant system are proposed for mitigating corrosion and SCC of nuclear-grade austenitic alloys.

关键词： pressurized water reactor water chemistry Zinc-injection Material corrosion Stress corrosion cracking

Development of CONTHAC-3D and hydrogen distribution analysis of HPR1000

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Nuclear Science and Techniques 2024年第2期35卷 210-221页

作者： Hui Wang Jing-Jing Li Yuan Chang Gong-Lin Li Ming Ding China Nuclear Power Engineering Co.LTD. Beijing 100084China Fundamental Science on Nuclear Safety and Simulation Technology Laboratory Harbin 100049China

An in-house code,CONTHAC-3D,was developed to calculate and analyze thermal-hydraulic phenomena in containments during severe ***-3D is a three-dimensional computational fluid dynamics code that can be applied to predict gas flow,diffusion,and steam condensation in a containment during a severe hypothetical accident,as well as to obtain an estimate of the local hydrogen concentration in various zones of the ***-3D was developed using multiple models to simulate the features of the proprietary systems and equipment of HPR1000 and ACP100,such as the passive cooling system,passive autocatalytic recombiners and the passive air cooling *** validate CONTHAC-3D,a GX6 test was performed at the Battelle Model Containment *** hydrogen concentration and temperature monitored by the GX6 test are accurately predicted by ***,the hydrogen distribution in the HPR1000 containment during a severe accident was *** results show that the hydrogen removal rates calculated using CONTHAC-3D for different types of PARs agree well with the theoretical values,with an error of less than 1%.As the accident progresses,the hydrogen concentration in the lower compartment becomes higher than that in the large space,which implies that the lower compartment has a higher hydrogen risk than the dome and large space at a later stage of the *** amount of hydrogen removed by the PARs placed on the floor of the compartment is small;therefore,raising the installation height of these recombiners appropriately is ***,we do not recommend installing all autocatalytic recombiners at high *** study findings in regard to the hydrogen distribution in the HPR1000 containment indicate that CONTHAC-3D can be applied to the study of hydrogen risk containment.

关键词： Hydrogen risk mitigation pressurized water reactor HPR1000 Thermal hydraulic CONTHAC-3D

Response characteristics of PWR primary circuit under SBLOCAs considering steam bypass discharging

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Nuclear Science and Techniques 2024年第6期35卷 189-201页

作者： Shuai Yang Xiang-Bin Li Yu-Sheng Liu Jia-Ning Xu De-Chen Zhang School of Nuclear Science and Engineering North China Electric Power UniversityBeijing 102206China Beijing Key Laboratory of Passive Safety Technology for Nuclear Energy Beijing 102206China Nuclear and Radiation Safety Center MEPBeijing 100084China

Small-break superposed station blackout(SBO)accidents are the basic design accidents of nuclear power *** the condition of a small break in the cold leg,SBO further increases the severity of the accident,and the steam bypass discharg-ing system(GCT)in the second circuit can play an important role in guaranteeing core *** explore the influence of the GCT on the thermal-hydraulic characteristics of the primary circuit,RELAP5 software was used to establish a numerical model based on a typical pressurized water reactor nuclear power *** different small breaks in the cold-leg super-posed SBO were selected,and the impact of the GCT operation on the transient response characteristics of the primary and secondary circuit systems was *** results show that the GCT plays an indispensable role in core heat removal during an accident;otherwise,core safety cannot be *** GCT was used in conjunction with the primary safety injection system during the placement *** the break diameter was greater than a certain critical value,the core cooling rate could not be guaranteed to be less than 100 K/h;however,the core remained in a safe state.

关键词： Steam bypass discharging pressurized water reactor SBLOCA Numerical simulation

Transient Behaviors of Thermo-Hydraulic and Thermal Stratification in the Pressurizer Surgeline for the Nuclear Power Plant

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Journal of Thermal Science 2022年第2期31卷 344-358页

作者： YU Huajie LI Lu TANG Qionghui PENG Yue LI Yinshi State Key Laboratory of Nuclear Power Safety Monitoring Technology and Equipment Shenzhen 518172China Key Laboratory of Thermo-Fluid Science and Engineering of Ministry of Education School of Energy and Power EngineeringXi’an Jiaotong UniversityXi’an 710049China School of Mechanics and Safety Engineering Zhengzhou UniversityZhengzhou 450001China

In pressurized water reactor(PWR)system,the surgeline plays an important role in bonding the pressurizer and the primary *** considerable problems,including the thermo-hydraulics,the thermal stratification and the accompanying thermal stress under transient conditions,pose risks to the surgeline ***,a fully-coupled flow-heat-thermo-elasticity model was developed to investigate the transient behavior of thermo-hydraulic parameters and the thermal stratification phenomenon in *** evaluate the nonuniformity of the stratified flow,a stratification degree indicatorζis *** is found that during the outsurge flow,the increase of temperature variation will enlarge the temperature gradient on the wall,corresponding to a more serious *** the cases of positive temperature variation,the higher temperature variation causes higher stratification degreeζ,and vice *** mass flow rate m and the stratification degree are in negative *** local deformation can reach 1.802 cm under a 50 K temperature variation,while its location varies from case to *** attention should be paid to the regulation between the highest deformation location and the surgeline thermo-hydraulic parameters.

关键词： pressurized water reactor surgeline transient thermo-hydraulic thermal stratification thermal stress-strain

Influence evaluation of loading conditions during pressurized thermal shock transients based on thermal-hydraulics and structural analyses

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Frontiers of Mechanical Engineering 2018年第4期13卷 563-570页

作者： Jinya KATSUYAMA Shumpei UNO Tadashi WATANABE Yinsheng LI Japan Atomic Energy Agency Ibaraki 319-1195 Japan Mizuho Information & Research Institute Inc. Tokyo 101-8443 Japan Fukui University Fukui 914-0055 Japan

The thermal hydraulic （TH） behavior of coo- lant water is a key factor in the structural integrity assessments on reactor pressure vessels （RPVs） of pressurized water reactors （PWRs） under pressurized thermal shock （PTS） events, because the TH behavior may affect the loading conditions in the assessment. From the viewpoint of TH behavior, configuration of plant equipment and their dimensions, and operator action time considerably influence various parameters, such as the temperature and flow rate of coolant water and inner pressure. In this study, to investigate the influence of the operator action time on TH behavior during a PTS event, we developed an analysis model for a typical Japanese PWR plant, including the RPV and the main components of both primary and secondary systems, and performed TH analyses by using a system analysis code called RELAP5. We applied two different operator action times based on the Japanese and the United States （US） rules： Operators may act after 10 min （Japanese rules） and 30 min （the US rules） after the occurrence of PTS events. Based on the results of TH analysis with different operator action times, we also performed structural analyses for evaluating thermal-stress distributions in the RPV during PTS events as loading conditions in the structural integrity assessment. From the analysis results, it was clarified that differences in operator action times significantly affect TH behavior and loading conditions, as the Japanese rule may lead to lower stresses than that under the US rule because an earlier operator action caused lower pressure in the RPV.

关键词： structural integrity reactor pressure vessel pressurized thermal shock thermal hydraulic analysis pressurized water reactor weld residual stress

Evaluation of Nonlinear Finite Element Module for the Simulation of Fuel Behavior

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Journal of Energy and Power Engineering 2013年第4期7卷 689-694页

作者： Hyo Chan Kim Yong Sik Yang Yang Hyun Koo Young Doo Kwon Light water reactor Fuel Technology Division Korea Atomic Energy Research Institute Daejeon 305-353 Korea Department of Mechanical Engineering Kyungpook National University Daegu 702-701 Korea

Because zirconium alloy cladding is the first containment barrier for fission products, its mechanical integrity is the most important concern. In view of the mechanical integrity, stress and strain are the main factors that affect the cladding performance during normal or off-normal operation, which induces force interaction between the pellet and cladding. In the case of a normal operation period, to estimate the cladding stress and strain, various models and codes have been developed using a simplified 1D （one-dimensional） assumption. However, in the case of a slow ramp during start-up and shut-down and a fast transient such as an AOO （anticipated operational occurrence）, it is difficult for a 1D model to simulate the cladding stress and strain accurately due to its modeling limitation. To model a large deformation along the radial and axial directions such as a ＂＇ballooning＂ phenomenon, FE （finite element） modeling, which can simulate a higher degree of freedom, is an indispensable requirement. In this work, an axisymmetric two-dimensional FE module, which will be integrated into the transient fuel performance code, has been developed. To solve the mechanical equilibrium of the pellet-cladding system, taking into account the geometrical and material non-linearities, the FE module employs an ESF （effective-stress-function） algorithm. Verifications of the FE module for the cases of thermal and elastic analyes were performed using the results of ANSYS 13.0.

关键词： pressurized water reactor fuel performance code finite element method thermo-mechanical analysis.

Combined Heat and Power Design Considerations for the APR1400

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Journal of Energy and Power Engineering 2017年第3期11卷 195-203页

作者： Michal Wierzchowski Robert M. Field KEPCO International Nuclear Graduate School (KINGS) 658-91 Haemaji-ro Seosaeng-myeon Ulju-gun Ulsan 689-882 Republic of Korea

To date, nuclear cogeneration applications have been limited, primarily to district heating in Eastern Europe and heavy water production in Canada. With the current global price for oil and energy, this technology is not economically viable for most countries. However, oil and fossil fuel prices are known to be highly volatile, and the Paris Agreement calls for a reduction in fossil fuel use. Under these circumstances, heat supplied by nuclear power may abruptly return to favor. To prepare for such a scenario, this study will investigate design considerations for a prototypical modem nuclear power plant, the Korean APR1400 （advanced power reactor 1400）（e.g., Shin Kori Units 3, 4, Shin Hanul 1, 2, Barakah Units 1, 2, 3, 4）. Nuclear cogeneration can impact balance of plant system and component design for the condensate, feedwater, extraction steam, and heater drain systems. The APR1400 turbine cycle will be reviewed for a parametric range of pressures and flow rates of the steam exported for cogeneration to identify major design challenges.

关键词： Cogeneration energy heat nuclear energy steam turbine design pressurized water reactor APR1400.

Thermal Economy Analysis Model with Matrix Method for the Secondary Loop of PWR Nuclear Power Station

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Thermal Economy Analysis Model with Matrix Method for the Se...

The General Design and Technology Innovations of CAP1400

2012年计算机应用与系统建模国际会议

作者： Li Yongling Huang Yu Ma Jin Wang Binshu North China Electric Power University North China Baoding Electric Power Voc. & Tech. College North China Electric Power University

According to characteristics of pressurized water reactor (PWR) nuclear power station, the two-stage moisture separator reheater (MSR) model was proposed with thermal equilibrium equation and mass conservation equation, finally the thermo-economy analysis model with matrix equation for secondary loop was put up. Based on the model, the thermal economy influence values of the auxiliary steam-water are calculated and compared with those of habitual thermal balance method (HTBM). The results indicate that the model presented in the article was correct;it is available and convenient for computer coding and online calculating. The matrix analysis method can provide theoretical basis for simulating of the secondary loop thermal system for PWR nuclear power station.

关键词： pressurized water reactor secondary loop thermal economy state equation Mechanism modeling

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Engineering 2016年第1期2卷 103-111页

作者：郑明光严锦泉申屠军田林王煦嘉邱忠明 Shanghai Nuclear Engineering Research and Design Institute

A historical review of in-vessel melt retention （IVR） is given, which is a severe accident mitigation mea- sure extensively applied in Generation III pressurized water reactors （PWRs）. The idea of IVR actually originated from the back-fitting of the Generation 11 reactor Loviisa WER-440 in order to cope with the core-melt risk. It was then employed in the new deigns such as Westinghouse APIO00, the Korean APR1400 as well as Chinese advanced PWR designs HPRIO00 and CAP1400. The most influential phe- nomena on the IVR strategy are in-vessel core melt evolution, the heat fluxes imposed on the vessel by the molten core, and the external cooling of the reactor pressure vessel （RPV）. For in-vessel melt evolution, past focus has only been placed on the melt pool convection in the lower plenum of the RPV; however, through our review and analysis, we believe that other in-vessel phenomena, including core degradation and relocation, debris formation, and coolability and melt pool formation, may all contrib- ute to the final state of the melt pool and its thermal loads on the lower head. By looking into previous research on relevant topics, we aim to identify the missing pieces in the picture. Based on the state of the art, we conclude by proposing future research needs.

关键词： pressurized water reactor Severe accident In-vessel melt retention Debris formationDebris remeltingMelt pool formationMelt pool thermal-hydraulicsCritical heat flux