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SSRN 2023年

作者： Chen, Junyi Liu, Guodong Huang, Shanfang He, Ying Qiu, Meiming Wang, Kan Lu, Chuan Li, Zhigang Quan, Guoping Wang, Xing Department of Engineering Physics Tsinghua University Haidian district Beijing100084 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Shuangliu district Chengdu610213 China State Power Investment Corporation Research Institute Changping district Beijing102209 China

A full scale high-fidelity neutronic/thermal-hydraulic coupled analysis method for plate-type reactor is proposed based on the open-source Multiphysics Object-Oriented Simulation Environment (MOOSE) and Monte Carlo Code cosRMC. The two-dimensional heat transfer model and one-dimensional coolant subchannel model are established, and integrated into the thermalhydraulic analysis code SPhase. The code-to-code validations of SPhase show good agreement with that of CTF and Fluent. The detailed neutronics model and parameters schemes are verified through cosRMC to ensure accuracy of physics calculation. The criticality search is implented, and the full scale JRR-3M critical coupled simulation under steady state is realized through celllevel resolution. The thermal feedback introduces negative activity of about 400 pcm. The insertion of absorbers renders the complexity of power distribution. The reflection effect results in peak thermal neutrons flux and temperature at the edge of the core. This work provides reference for correlational studies on neutronic/thermal-hydraulic characteristics of plate-type reactor. © 2023, The Authors. All rights reserved.

关键词： Multiphysics

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Study on irradiation damage of rpv steels based on coupling cluster dynamics and crystal plasticity finite element method

Study on irradiation damage of rpv steels based on coupling ...

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

作者： Wang, Xiaotong Luo, Ying Dong, Yuanyuan Yao, Weihua Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791883785

Irradiation hardening is one of the most important aging effects of reactor pressure vessel (RPV) steel in long-term service. A number of studies have indicated that irradiation hardening is mainly caused by irradiation induced defects, such as dislocation loops and precipitates. In this paper, we have simulated the irradiation damage of low-copper reactor pressure vessel (RPV) steel. The generation of Mn-Ni-Si precipitates is simulated by the cluster dynamics model based on rate theory. On this basis, the crystal plasticity finite element method based on modified crystal plasticity model is used to simulate the effect of Mn-Ni-Si precipitates on the mechanical properties of RPV steels. The simulated results has been compared with the experimental results from the literature. By coupling the cluster dynamics and the crystal plastic finite element method, we suggest a multi-scale simulation method to simulate and predict irradiation damage of RPV steel. © 2020 American Society of Mechanical Engineers (ASME). All rights reserved.

关键词： Finite element method

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Burnup performance analysis for candidate materials of the LWR accident tolerant control rod 14

Burnup performance analysis for candidate materials of the L...

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14th International Nuclear Fuel Cycle Conference, GLOBAL 2019 and Light Water reactor Fuel Performance Conference, TOP FUEL 2019

作者： Liu, Yong Li, Mancang Yu, Yingrui Xiao, Peng Wang, Xingbo Zhang, Bin Wang, Dongyong Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China 328# Section 1 Changshun Avenue Chengdu China

The accident tolerant fuel (ATF) is considered to be one of the most advanced concepts to improve the safety of light water reactors (LWRs). The accident tolerant control rod (ATCR) is a significant aspect in the ATF concept. The melting temperature of traditional LWR control rod (CR) materials, such as silver-indium-cadmium (AIC) alloy, is too low to endure the high temperature under several accident conditions. Consequently, varieties of candidate materials for control rod have been proposed in the previous researches because of the outstanding characters of high-temperature behavior, as well as reactivity worth, stability in coolant water and so on. In this paper, the burnup performance of the sesquioxides of the rare-earth elements (RE2O3) and the hafnia (Hf) are analyzed and compared to the traditional AIC control rod. Two parameters are chosen to reveal the burnup performance of the candidate control rod materials, the initial control rod reactivity worth and the stability of the control rod reactivity worth along with the assembly burnup. The candidate-material control rods and the AIC control rod are placed in the ATF assembly with the same conditions, and the assembly is depleted to sufficient burnup. Numerical results indicate that Eu2O3, Dy2O3, Sm2O3 and Hf are excellent ATCR materials compared to AIC. The initial control rod reactivity worth of these materials is not less than that of AIC, and the burnup stability is outstanding, especially Eu2O3 and Dy2O3 Copyright © GLOBAL 2019 - International Nuclear Fuel Cycle Conference and TOP FUEL 2019 - Light Water reactor Fuel Performance *** rights reserved.

关键词： Hafnium oxides

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DC-Side Decoupling Method of DC-link Capacitor Reduction for Single Phase PWM Rectifier

DC-Side Decoupling Method of DC-link Capacitor Reduction for...

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2020 IEEE Sustainable Power and Energy Conference, iSPEC 2020

作者： Wang, Jinxin Renyong, Peng Dai, Kailei Zhang, Zhiqiang Zhang, Ting Zhao, Junjie Chen, Qian Zhang, Jianjian Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9781728191645

The integrated charging system of Electric vehicle consists of a single-phase PWM rectifier and a Buck converter. When the single-phase PWM rectifier works with unit power factor, the dc-link voltage contains a large voltage harmonic of twice the grid frequency. Although the output voltage ripple can be controlled within the required range using the proper control of the Buck converter, the second-order harmonic on the dc-link capacitor is inevitable. This paper proposes a de-side decoupling method for the second-order harmonic problem of the dc-link voltage of the integrated charging system. A de-side decoupling half bridge is connected in parallel to absorb the ripple power of twice the grid frequency. This paper analyzes the operating principle of the de-side decoupling half bridge. Based on a theoretical analysis of the topology and the design procedure for controlling system, the proposed converter is designed and simulated. Finally, simulation results are given to verify the method. © 2020 IEEE.

关键词： Pulse width modulation

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Study on the thermal-mechanical performance of SiC composites cladding under multiple conditions

Study on the thermal-mechanical performance of SiC composite...

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

作者： Yin, Chunyu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China No.328 section 1 changshun avenue Chengdu China

ISBN: (纸本)9780791883761

SiC has become a candidate cladding material of Accident Tolerant Fuels (ATF) due to its excellent irradiation stability and corrosion resistance. However, because SiC is a ceramic material with low toughness, brittle failure is a significant concern. In order to improve the toughness, SiC fiber is required to manufacture multi-layer SiC composites. But the current performance model or analysis tool is not available for SiC composites cladding due to its obviously difference with Zr alloy cladding. On one side, Finite element method was used in this paper to analyze the performance of SiC composites cladding under operation conditions which include normal, transient conditions and LOCA conditions;on the other side, this paper gives the performance of the SiC composites with two layers under multiple operating conditions. The result showed that the temperature was stable and the maximum hoop stress was reached at about 70d under normal condition. The power ramp can increase the cladding temperature and has visible influence on the stress distribution. The hoop stress of the cladding reversed under LOCA condition. The tensile hoop stress on the outer surface significantly increased, which caused the obvious increase of failure probability of monolithic SiC, and the failure probability of SiC layer is significantly increased. The conclusion of the analysis has guiding significance for the theoretical design of SiC composites. Copyright © 2020 ASME.

关键词： Silicon carbide

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Effect of residual pore on the performance of UN kernel TRISO particle fuel 14

Effect of residual pore on the performance of UN kernel TRIS...

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14th International Nuclear Fuel Cycle Conference, GLOBAL 2019 and Light Water reactor Fuel Performance Conference, TOP FUEL 2019

作者： Liu, Shichao Chen, Ping Zhou, Yi Pang, Hua Xi, Yong Tang, Changbing Zhao, Yanli Qiu, Xi Guo, Zixuan Nuclear power institute of China Science and Technology on Reactor System Design Technology Laboratory Changshun Stress No.1 Chengdu Sichuan610200 China

The thermal-mechanical performance of uranium nitride (UN) tri-isotropic (TRISO) particle with residual pore used in Light Water reactor (LWR) was investigated. The effect of pore size and location on the performance of inner dense pyrolytic carbon (IPyC) layer, SiC layer and outer dense pyrolytic carbon (OPyC) layer was studied. The dimensional changes, fission gas release and internal pressure of the fuel particle layers were calculated. Residual pore size can affect the stress distribution of coated layers obviously. The residual pore especially the big ones may be the source of cracks on the SiC layer, which can break or loss the capacity of SiC layer. Residual pore location has little influence on the stress distribution of the coated layer. In order to maintain the fission retention capacity of TRISO particle, the big residual pore should be avoided during fabrication process. Copyright © GLOBAL 2019 - International Nuclear Fuel Cycle Conference and TOP FUEL 2019 - Light Water reactor Fuel Performance *** rights reserved.

关键词： Silicon carbide

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Simulation of Nuclear reactor Accident Scenarios using Physics-Informed Neural Networks and Transfer-Learning

Simulation of Nuclear Reactor Accident Scenarios using Physi...

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Electronic Information Engineering and Computer science (EIECS), 2021 International Conference on

作者： Yang Yu Yufei Xie Wenlin Wang Guohua Wu Haitao Lan Enbo Lin Ping An Zibin Sun Haichuan Zhang Yixian Wu National Key Laboratory of Parallel and Distributed Computing National University of Defense Technology Changsha China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China School of Automation Wuhan University of Technology Wuhan China Sino-German College of Intelligent Manufacturing Shenzhen Technology University Shenzhen China Institute of Automotive Engineers Hubei University of Automotive Technology Shiyan China Nuclear Power Institute of China Chengdu China

ISBN: (数字)9798331531409

ISBN: (纸本)9798331531416

In a Loss of Coolant Accident (LOCA), reactor core temperatures can rise rapidly, leading to potential fuel damage and radioactive material release. This research presents a groundbreaking method that combines the power of Monte Carlo Sampling and Physics-Informed Neural Networks (PINNs) to simulate and effectively address the challenging Loss of Coolant Accidents (LOCA) in nuclear reactors. In the event of a LOCA, reactor core temperatures can soar rapidly, posing a significant threat to fuel integrity and potentially leading to the release of radioactive materials. By leveraging the strengths of both Monte Carlo Sampling and PINNs, this approach aims to provide a comprehensive and accurate simulation framework for assessing and mitigating the consequences of such accidents. The method yields high prediction accuracy (MAE: 0.033, RMSE: 0.098, R2: 0.814) and demonstrates robustness through transfer learning, maintaining strong performance (MAE: 0.064, RMSE: 0.163, R 2 : 0.735).

关键词： Radioactive materials Accuracy Monte Carlo methods Transfer learning Neural networks Robustness Data models Fuels Inductors Accidents

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Facility Layout in Horizontal Cylindrical Space Based on Adaptive Gradient Descent Algorithm

Facility Layout in Horizontal Cylindrical Space Based on Ada...

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第40届中国控制会议

作者： Peng Yuan Yiran Miao Deqing Huang Junkai Hao Na Qin School of Electrical Engineering Southwest Jiaotong University Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China

FLP（Facility layout problem） is a common problem in engineering field. The past research tend to focus on the twodimensional space and a small amount of three-dimensional box space as well as vertical cylinder space. The FLP in horizontal cylinder space is less researched. Lots of FLP of box space and vertical cylinder space can transform to FLP of two-dimensional space easily, however, the FLP of horizontal cylinder space need design special calculation method of overlap constraint and boundary constraint. This paper designed the calculation method of overlap constraint and boundary constraint, used adaptive gradient descent algorithm with heuristic strategy to optimize the facility layout in horizontal cylinder space. On this basis,added symmetry constraint, absolute height constraint and relative distance constraint to close to the real engineering *** experiment results confirm the feasible of proposed scheme.

关键词： facility layout problem horizontal cylinder space gradient descent algorithm

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Corrosion Behavior of Micro-Arc Oxidation Coating on Zirconium Alloy in Superheated Steam Condition

SSRN

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SSRN 2022年

作者： Li, Zheng-Yang Zheng, Mei-Yin Yang, Zhong-Bo Ren, Quan-Yao Cai, Zhen-Bing Jiao, Yong-Jun Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China Tribology Research Institute Key Lab of Advanced Technologies of Materials Southwest Jiaotong University Chengdu610031 China

The corrosion behavior of MAO coating on Zr alloy was investigated in 400 ºC/10.3 MPa condition. The surface morphology, phase composition, and element distribution of MAO coating were analyzed. The results indicate that the oxide layer on MAO coating presents fewer cracks and pores than Zr alloy, because the high compressive stress can stabilize the t-ZrO2 and restrain the phase transformation, thus reducing the volume expansion and future decreasing defect. Based on the XRD, Raman mapping, SEM, XPS, and TEM results, the formation mechanism of t-ZrO2 and phase transformation process of MAO coating in superheated steam condition was summarized. © 2022, The Authors. All rights reserved.

关键词： Phase transitions

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Neutronic and thermal-mechanical coupling schemes for heat pipe cooled reactor designs

Neutronic and thermal-mechanical coupling schemes for heat p...

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

作者： Ma, Yugao Liu, Minyun Han, Wenbin Xie, Biheng Chai, Xiaoming Huang, Shanfang Yu, Hongxing Department of Engineering Physics Tsinghua University Beijing China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791883785

Space fission power systems can enable ambitious solarsystem and deep-space science missions. The heat pipe cooled reactor is one of the most potential candidates for near-term space power supply, featured with safety, simplicity, reliability, and modularity. Heat pipe cooled reactors are solid-state and high temperature (up to 1500 K) reactors, where the thermal expansion is remarkable and the mechanical response significantly influences the neutronics and thermal analyses. Due to the considerable difference between heat pipe cooled reactors and traditional water reactors in the structure and design concept, the coupling solutions for light water reactors cannot be directly applied to heat pipe cooled reactor analyses. Therefore, new coupling framework and program need to consider the coupling effects among neutronics, heat transfer as well as mechanics. Based on the Monte Carlo program RMC and commercial finite element program ANSYS Mechanical APDL, this work introduces the three coupling fields of neutronics (N), thermal (T), and mechanics (M) for heat pipe cooled reactors. The neutronic and thermal-mechanical (N/T-M) coupling strategy is developed theoretically, focusing on the formulation of the nonlinear problem, iteration schemes, and relaxation methods. Besides, the finite element method and the Monte Carlo program use different meshes and geometry construction methods. The spatial mapping and geometry reconstruction are also essential for the N/T-M coupling, which is discussed and established in detail. Furthermore, the N/T-M coupling methods are applied to the preliminary self-designed 10 kWe space heat pipe cooled reactor. Coupling shows that the thermal-mechanical feedback in the solid-state reactor has negative reactivity feedback (-2007 pcm) while it has a deterioration in heat transfer due to the expansion in the gas gap. © 2020 American Society of Mechanical Engineers (ASME). All rights reserved.

关键词： Heat pipes

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