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Non-integer Order Control Scheme for pressurized water reactor Core Power

Computers, Materials & Continua 2022年第7期72卷 651-662页

作者： Ibrahim M.Mehedi Maher H.AL-Sereihy Asmaa Ubaid Al-Saggaf Ubaid M.Al-Saggaf Department of Electrical and Computer Engineering(ECE) King Abdulaziz UniversityJeddah21589Saudi Arabia Center of Excellence in Intelligent Engineering Systems(CEIES) King Abdulaziz UniversityJeddah21589Saudi Arabia

Tracking load changes in a pressurized water reactor(PWR)with the help of an efficient core power control scheme in a nuclear power station is very important.The reason is that it is challenging to maintain a stable core power according to the reference value within an acceptable tolerance for the safety of PWR.To overcome the uncertainties,a non-integer-based fractional order control method is demonstrated to control the core power of PWR.The available dynamic model of the reactor core is used in this analysis.Core power is controlled using a modified state feedback approach with a non-integer integral scheme through two different approximations,CRONE(Commande Robuste d’Ordre Non Entier,meaning Non-integer orderRobust Control)and FOMCON(non-integer order modeling and control).Simulation results are produced using MATLAB■program.Both non-integer results are compared with an integer order PI(Proportional Integral)algorithm to justify the effectiveness of the proposed scheme.Sate-spacemodel Core power control Non-integer control pressurized water reactor PI controller CRONE FOMCON.

关键词： Sate-space model core power control non-integer control pressurized water reactor PI controller CRONE FOMCON

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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 accidents.CONTHAC-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 containment.CONTHAC-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 system.To validate CONTHAC-3D,a GX6 test was performed at the Battelle Model Containment facility.The hydrogen concentration and temperature monitored by the GX6 test are accurately predicted by CONTHAC-3D.Subsequently,the hydrogen distribution in the HPR1000 containment during a severe accident was studied.The 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 accident.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 recommended.However,we do not recommend installing all autocatalytic recombiners at high positions.The 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

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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 circle.Some considerable problems,including the thermo-hydraulics,the thermal stratification and the accompanying thermal stress under transient conditions,pose risks to the surgeline integrity.Herein,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 PWR.To evaluate the nonuniformity of the stratified flow,a stratification degree indicatorζis introduced.It 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 deformation.In the cases of positive temperature variation,the higher temperature variation causes higher stratification degreeζ,and vice versa.The mass flow rate m and the stratification degree are in negative correlation.The local deformation can reach 1.802 cm under a 50 K temperature variation,while its location varies from case to case.More 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

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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

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Analysis of Severe Accident Management of Maanshan Nuclear Power Station Using MAAP4 Code

Analysis of Severe Accident Management of Maanshan Nuclear P...

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第13届太平洋地区核能大会

作者： Shih-Jen Wang Institute of Nuclear Energy Research, Taiwan, China Kwang-Sheng Chiang Radiation Protection Association, Taiwan, China Ching-Chuan Yao Show-Chyuan Chiang Jer-Cherng Kang Taiwan Power Company, Taiwan, China

＜正＞Maanshan is a Westinghouse pressurized water reactor nuclear power station（NPS） located in south Taiwan. The severe accident management guideline （SAMG） ofMaanshan NPS is developed based on the Westinghouse Owners Group （WOG） SAMGA station blackout （SBO） accident for Maanshan NPS is cited as a reference case. A SBO accident is simulated first. Mitigation actions are used to demonstrate theeffect of the SAMG A RCS depressurization, RCS injection, and reduce containmenthydrogen strategies are used to restore the system as the vessel water level decreases tothe bottom of the active fuel （BAF）. The effect of the RCS depressurization, RCSinjection, and reduce containment hydrogen strategies will be analyzed with MAAP4code.

关键词： Severe Accident Management pressurized water reactor MAAP4 SBO

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

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

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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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