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2020 International Conference on Nuclear Engineering, ICONE 2020, collocated with the ASME 2020 Power Conference

作者： Lei, Zhong Deng, Jian Li, Wei Wu, Xiaoli Deng, Chunrui Nuclear Power Institute of China Chengdu Sichuan China Science and Technology on Reactor System Design Technology Laboratory Chengdu Sichuan China

ISBN: (纸本)9780791883785

Core melting and molten migration behavior are hot and difficult issues in the field of nuclear reactor severe accident research. The Moving Particle Semi-implicit (MPS) meshless method has potential to simulate free-surface and multiphase flows. In this study, the MPS method was utilized to simulate the melting process of UO2-Zr rod-type fuel elements. The models of heat conduction with phase change, simplified UO2-Zr eutectic reaction, viscous flow and surface tension were implemented with the framework of standard MPS method. Then, the improved MPS code was used to simulate and analyze the process of high-temperature melting and characteristics of molten migration and solidification in the coolant channel, aiming at revealing the severe accidents for light water reactors (LWR), particularly the early core damage. The results showed that compared with the case of higher initial temperature, when the initial temperature of molten UO2 is lower, more molten UO2 will solidify on the surface of rod cluster, and the blockage of upper flow channel caused by molten UO2 is more serious. In addition, this study also demonstrated the potential of the MPS method for the study of complicated severe accident phenomena in not only traditional LWR but also advanced nuclear reactors in the future. © 2020 American Society of Mechanical Engineers (ASME). All rights reserved.

关键词： Solidification

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Study on the liquid seal discharge process in an over-pressurized accident

Study on the liquid seal discharge process in an over-pressu...

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2020 International Conference on Nuclear Engineering, ICONE 2020, collocated with the ASME 2020 Power Conference

作者： Wu, Dan Deng, Jian Du, Sijia Qian, Libo Nuclear Power Institute of China Chengdu Sichuan China Science and Technology on Reactor System Design Technology Laboratory Chengdu Sichuan China

ISBN: (纸本)9780791883785

In an over pressure accident, one or more pressurizer safety(or relief) valves will open due to the rapid pressure rise process. Once the safety(or relief) valves are open, the liquid seal will be discharged, and this will generate great discharge force to the downstream pipes. Multi-level protection is chosen using pressurizer safety(or relief) valves with different setpoint in most of Nuclear Power Plant, especially in the self-designed Generation-III Nuclear Power Plants. As the over pressure accident progresses, one or more safety(or relief) valves will be open. The downstream pipes will experience one or more times of impacts, which will influence the arrangement of the pipes. The whole discharge process is very complex, and the key influence factors are the pressure rise rate, safety(or relief) valve opening time, liquid seal temperature and volume, and the arrangement of the downstream discharge pipes. In present paper, liquid seal discharge process in an over pressure accident is studied. The pressure rise rate is so fast that three safety(or relief) valves will open one after another, which will generate three impacts on the downstream discharge pipes. It is found that for a specific design of Nuclear Power Plant, well design of the safety(or relief) valve setpoint is very important to the discharge force analysis results. © 2020 American Society of Mechanical Engineers (ASME). All rights reserved.

关键词： Liquids

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KYLIN-V2.0 code calculation ability verification based on VERA benchmark

KYLIN-V2.0 code calculation ability verification based on VE...

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2020 International Conference on Physics of reactors: Transition to a Scalable Nuclear Future, PHYSOR 2020

作者： Wang, Dongyong Yu, Yingrui Peng, Xingjie Wang, Chenlin Liu, Kun Chai, Xiaoming Science and Technology on Reactor System Design Technology Laboratory Chengdu610213 China Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9781713827245

Virtual Environmental for reactor Analysis (VERA) benchmark was released by the Consortium for Advanced Simulation of Light water reactors (CASL) project in 2012. VERA benchmark includes more than ten problems at different levels, from 2D fuel pin case to 2D fuel assembly case to 3D core refuelling case, in addition, reference results and experimental measured data of some problems were provided by CASL. Fuel assemblies in VERA benchmark are various, including control rod assemblies, Pyrex assembly, IFBA assembly, WABA assembly and gadolinium poison assembly, and so on. In this paper, various fuel assembly models in the VERA benchmark have been built by using KYIIN-V2.0 code to verify its calculation ability from 2D fuel pin case to 2D fuel assembly case to 2D 3x3 fuel assembly case, and making a comparative analysis on the reference results in VERA benchmark, as well as the calculation results of the Monte Carlo code RMC. KYLIN-V2.0 is an advanced neutron transport lattice code developed by Nuclear Power Institute of China (NPIC). The subgroup resonance calculation method is used in KYIIN-V2.0 to obtain effective resonance self-shielding cross section, method of modular characteristics (MOC) is adopted to solve the neutron transport equation, and CRAM method and PPC method is adopted to solve the depletion equation. The numerical results show that KYLIN-V2.0 code has the reliable capability of direct heterogeneous calculation of 2D fuel assembly, and the effective multiplication factor, assembly power distribution, rod power distribution and control rod reactivity worths of various fuel assemblies that are calculated by KYLIN-V2.0 are in better agreement with the reference. © The Authors, published by EDP sciences. This is an open access article distributed under the terms of the Creative Commons Attribution License 4.0 (http://***/licenses/by/4.0/).

关键词： Control rods

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The Grid Mapping for Pwr Core Pin-by-Pin Wise Coupling Calculation system

SSRN

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SSRN 2024年

作者： Li, Zhigang Pan, Junjie Xia, Bangyang Qiang, Shenglong Zeng, Wei Lu, Wei Feng, Ziliang Nuclear Power Institute of China 328 Huayang Changshun Avenue Section 1 Shuangliu County Chengdu China Science and Technology on Reactor System Design Technology Laboratory 328 Huayang Changshun Avenue Section 1 Shuangliu County Chengdu China Sichuan University No.24 South Section 1 Yihuan Road Chengdu China

Due to the possible cross between the fields during the neutronic/thermal-hydraulic coupling calculation at the pin-by-pin wise of reactor core, there is an urgent need for efficient mapping methods to improve the efficiency of large-scale grid mapping. This article combines the geometric characteristics of the pins in the reactor and proposes four acceleration strategies such as two-level grid search and OpenMP+MPI parallel based on the volume weight method. A mapping tool named MAP suitable for straight hexahedral grids has been developed. The evaluation based on 6 cases and 4 types mapping processes has been conducted, and compared with the open-source MEDCoupling. The results show that the four strategies proposed have achieved good acceleration effects. Compared with MEDCoupling, the average acceleration ratio of MPI-B reaches 21332.56, with an average relative deviation range of [0,1.57e-04]. The mapping time between millionth scale and million scale using MPI is less than 1.0s. © 2024, The Authors. All rights reserved.

关键词： Mapping

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Research on analysis and modeling methods for lbloca in nuclear power plants based on autonomous loca analysis platform arsac

Research on analysis and modeling methods for lbloca in nucl...

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2020 International Conference on Nuclear Engineering, ICONE 2020, collocated with the ASME 2020 Power Conference

作者： Zhou, Jiayue Wu, Dan Ding, Shuhua Jiang, Guangming Nuclear Power Institute of China Chengdu Sichuan China Science and Technology on Reactor System Design Technology Laboratory Chengdu Sichuan China

ISBN: (纸本)9780791883785

In order to meet the demand of continuous innovation of technologies and the general trend of autonomous nuclear power plants design and export of nuclear power plants, it is necessary to develop an autonomous LOCA analysis platform and corresponding analysis methods for the most complex design basis accidents. In this paper, the characteristics of LOCA analysis platform ARSAC, designed by Nuclear Power Institute of China, and the code ARSAC-K which meets the requirements of the US Federal Code 10 CFR 50.46 Appendix K model are introduced as well as a set of LOCA analysis methods and modeling methods. Based on the international advanced LOCA analysis code development concept, the code ARSAC has made new breakthroughs in matrix algorithms, key thermal hydraulic models and so on. Validation work has also been carried out in-depth. A set of advanced LOCA analysis methods has been developed using code ARSAC-K and advanced power plant parameter sampling methods. Analysis on LBLOCA of nuclear power plants with code ARSAC-K was performed, and the impact of different modeling methods on the LOCA analysis results was studied. To ensure the rationality and conservativeness of the analysis results, a set of reasonable and conservative modeling methods is fixed on the basis of a large number of sensitivity analyses for subsequent analysis and calculation. In the future, a lot of optimization work will be done to improve the LOCA code and corresponding methods. © 2020 American Society of Mechanical Engineers (ASME). All rights reserved.

关键词： Nuclear power plants

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A Metamorphic Relation Identification Method based on GEP 9

A Metamorphic Relation Identification Method based on GEP

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9th International Conference on Dependable systems and Their Applications, DSA 2022

作者： Hong, Jie Zhang, Jie Qiu, Qi Ma, Angang Li, Meng Yan, Shiyu Gong, Helin School of Computing University of South China Hunan Hengyang China Eng. and Technol. Res. Ctr. of Software Eval. and Test. for Intellectual Equipment of Hunan Province Hunan Hengyang China Cnnc Key Laboratory on High Trusted Computing Hunan Hengyang China Nuclear Power Institute of China Science and Technology on Reactor System Design Technology Laboratory Chengdu China

ISBN: (数字)9781665488778

ISBN: (纸本)9781665488778

Metamorphic testing is one of the effective methods to alleviate the test oracle problem. Metamorphic relation is the core of metamorphic testing, and there is no effective automatic identification technology. This paper transforms the metamorphic relation recognition issue into a symbolic expression regression problem. The test input pairs are generated using the preset input pattern, and the output pattern is mined by gene expression programming. After reduction and verification, a valid metamorphic relation is obtained. Experiments show that this method can identify the relation obtained by static analysis and a more significant number. At the same time, compared with existing technologies, it has apparent advantages in effectiveness, reliability, and performance. © 2022 IEEE.

关键词： Static analysis

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Sensitivity Study of Thermal Hydraulics to Corrosion of Heat Exchange Tubes in Steam Generator

SSRN

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SSRN 2022年

作者： Wang, Weibing Cheng, Kun Wang, Bo Zhang, Meng Tian, Ruifeng College of Nuclear Science and Technology Harbin Engineering University Harbin150001 China Heilongjiang Provincial Key Laboratory of Nuclear Power System & Equipment Harbin Engineering University Harbin150001 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

A leakage accident occurs in the circulating water system in PWRs, impurity ions enter the SG with the secondary circuit feed water and cause electrochemical corrosion of the heat transfer tubes(HTTs). Based on the hide out and return mathematical model in the SG, a coupled mathematical model of mass transfer-electromigration-hydrodynamics of impurity ions in the secondary boiling channel of SG is deduced. The accuracy of the mathematical model is verified by experiments, and the effects of heat flux density ( Pe q ), flow velocity ( Pe u ), impurity ion concentration ( Pe c ) and relative solubility of impurity ions ( K p ) on impurity ions migration and electrochemical corrosion are analyzed and discussed. The results show that P e q mainly controls the diffusion of impurity ions and thus affects the migration rate of impurity ions, and the mass transfer of impurity ions driven by Pe q can enhance the reduction of H + , thereby accelerating the corrosion rate of HTTs. The synergistic effect of convective mass transfer and electromigration during impurity ions migration is mainly affected by Pe u , and the electrochemical corrosion of the surface of HTTs is mainly controlled by charge transfer with the increase of Pe u . The Pe c on the surface of HTTs changes linearly with time, and the corrosion rate increases with the increase of Pe c . The corrosion rate of HTTs is proportional to K p , and K p is a main factor that causes different corrosion rates in different location of HTTs. In this paper, a coupled model for the sensitivity of thermal-hydraulic parameters ( Pe q , Pe u , Pe c and K p ) in SG to electrochemical corrosion of HTTs is proposed for the first time. The research results can provide important support for the development of corresponding response strategies, schemes for the life prediction of HTTs and the safety maintenance of nuclear power plants. © 2022, The Authors. All rights reserved.

关键词： Steam generators

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Effects of rolling on characteristics of system under forced circulation and natural circulation

Effects of rolling on characteristics of system under forced...

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2020 International Conference on Nuclear Engineering, ICONE 2020, collocated with the ASME 2020 Power Conference

作者： Cai, Rong Yue, Nina Fang, Hongyu Chen, Baowen Liu, Lili Ma, Zehua Science and Technology on Reactor System Design Technology Laboratory Chengdu China Cnnc Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology Chengdu China Xi'an Jiaotong University Xi'an Shaanxi China

ISBN: (纸本)9780791883785

The marine nuclear power plant operating in the marine environment has complicated motion under the influence of wind and waves. The movement of marine nuclear power plant will affect the thermal-hydraulic characteristics of its nuclear reactor system. Compared with other typical motion conditions, the effects of rolling conditions on the thermal-hydraulic characteristics of the nuclear reactor system are the most complex. In order to study the effects of rolling conditions on the thermal-hydraulic characteristics of the nuclear reactor system, a thermal-hydraulic system code for motion conditions named STAC was developed. The STAC code was verified by the experiments conducted in Japan. The effects of rolling conditions on the thermal-hydraulic characteristics of the nuclear reactor system under forced circulation and natural circulation are studied with the STAC code. The simulation results show that the thermal parameters of the reactor system under rolling condition fluctuate periodically. The fluctuation period of the thermal parameters of the core is half of the rolling period, and the fluctuation periods of other thermal parameters are the same as the rolling period. The effect of rolling condition on thermalhydraulic parameters under forced circulation is smaller than that under natural circulation. The fluctuation amplitudes of the thermal parameters increase with the angle amplitude of the rolling condition. There is a rolling period with the smallest fluctuation amplitude. Under the rolling condition with short period, the fluctuation amplitudes of the thermal parameters increase and their average values change rapidly as the rolling period decreases. Under the rolling condition with large period, the fluctuation amplitudes of the thermal parameters increase with the rolling period, and they tend to fixed values. © 2020 American Society of Mechanical Engineers (ASME). All rights reserved.

关键词： Nuclear fuels

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The Calculation Method of Neutron Transport Equation and its Conjugate Equation Based on Finite Deconvolution Neural Network

SSRN

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SSRN 2022年

作者： Li, Xiangyu Cheng, Kun Huang, Tao Tan, Sichao Heilongjiang Provincial Key Laboratory of Nuclear Power System & Equipment Harbin Engineering University Harbin15001 China College of Nuclear Science and Technology Harbin Engineering University Harbin150001 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

Aiming at the problem that the current neutron transport equation solving method, which is mainly numerical calculation, takes too long to calculate the inverse of the stiffness matrix, this paper designs a shallow neural network algorithm based on finite deconvolution neural network, which does not need to calculate the inverse matrix for the sti ness matrix or use other iterative algorithms. The algorithm uses a deconvolution neural network algorithm as the basic structure, replaces the deconvolution kernel with the finite volume method, and uses the calculation equation of reactor power as the loss function. The gradient descent method and the power residual are used to update the algorithm parameters to ensure algorithm convergence. The algorithm can use the neutron transport equation and its conjugate equation to calculate the distribution of neutron flux density and neutron importance in the space. © 2022, The Authors. All rights reserved.

关键词： Numerical methods

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Probabilistic model for state assessment of offshore pipeline with anti-corrosion coating based on Wiener process and Bayesian network

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Ocean Engineering 2025年 337卷

作者： Xiwen Xie Haiying Li Changzheng Yin Changhong Peng School of Nuclear Science and Technology University of Science and Technology of China Hefei China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

The demand for fresh water resources and sustainable energy is continuously growing, leading to the rapid development of nuclear energy systems. Maintaining a stable supply of electricity requires the essential service water system (SEC) to effectively cool the power generation equipment. As a key structure connecting the SEC and seawater, the integrity of pipeline affects the safety and economy of the power production. Reasonably assessing the corrosion state of pipelines in the complex marine environment is crucial for maintenance decision-making. By integrating stochastic process theory and probabilistic methods, this paper proposes an effective framework for evaluating the corrosion depth of pipelines with anti-corrosion coatings. The proposed framework leverages Wiener processes to describe the degradation behavior of anti-corrosion coatings and establishes availability models of repairable epoxy coatings with two feasible maintenance policies: periodic maintenance and state-based maintenance. Furthermore, to better align the model with real-world engineering conditions, the influence of maintenance quality on coating availability is considered. A Bayesian network is utilized to infer the probabilistic relationships between environmental variables and corrosion model parameters, with the posterior distribution means being used for corrosion depth predictions. The proposed method is practically implemented through a detailed case study of the SEC pipelines at a Chinese offshore nuclear power facility. The results illustrate the necessity of regular inspection and maintenance of pipelines to ensure their long-term service life and provide recommendations for the optimization of maintenance decisions. This framework accounts for various environmental factors and inherent uncertainties, providing a robust and comprehensive approach for estimating the long-term corrosion state of coated pipelines in marine environments.

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