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Optimization of Manufacturing Process and Experimental Verification of Vulnerable Part in reactor

IOP Conference Series: Earth and Environmental science 2020年第1期585卷

作者： Yang Bo Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institution of China Sichuan ChengDu 610041

The technical requirements of the control rod driving mechanism of the third generation PWR nuclear power plant are higher. The latch parts are designed with double tooth structure and welded on the wear-resistant surface by cobalt-based alloy surfacing. According to the structural characteristics and manufacturing process difficulties, a special welding device is developed. According to the test and finite element simulation, the related process parameters are optimized. The trial production of the parts was completed, and the parts were tested by metallography, liquid permeation test, hardness test and thermal life test. The results show that the latch parts have high hardness and wear resistance, and meet the requirements of the driving mechanism operation life of the third generation PWR nuclear power plant.

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Comparisons of Subgrid-Scale Models for OpenFoam Large- Eddy Simulation

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Journal of Physics: Conference Series 2021年第4期1802卷

作者： Zhipeng Feng Huanhuan Qi Xuan Huang Shuai Liu Jian Liu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China

OpenFOAM is a free, open-source software package that can be used for the solutions of computational fluid dynamics and simulation of various fluid flow processes. Nevertheless, OpenFOAM still lacks default settings and a large number of different numerical schemes and turbulent models should be validated. In this paper, the unsteady flow around a cylinder (Re=3900) is calculated by the large eddy simulation of OpenFOAM. The predictions include the drag and lift coefficient, the pressure distribution around the cylinder, the velocity distribution and Reynolds stress distribution in the wake region, as well as the prediction of the recirculation length and separation angle. Thanks to several simulations, these five subgrid-scale (SGS) models are compared and studied: The Smagorinsky SGS model, wall adaptive local eddy visibility SGS model, dynamic SGS model with Lagrangian averaging, dynamic one equation eddy visibility model, one equation eddy visibility model. The numerical results are verified with the published experimental data.

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Fatigue and fracture mechanical behavior for Chinese A508-3 steel at room temperature

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IOP Conference Series: Materials science and Engineering 2018年第1期372卷

作者： K K Shi H Xie B Zheng X L Fu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu 610213 China

Material, A508-3 steel, has been used in nuclear reactor vessels. In the present study, fatigue and fracture mechanical behavior of Chinese A5083 steel at room temperature are studied by mechanical material testing machine (MTS). Test data of material's mechanical behavior including uniaxial tension, low cycle fatigue (LCF), threshold value of stress intensity factor (SIF) range, fatigue crack growth (FCG), and fracture toughness is generated and given for further study. It is worth noting that the model in predicting FCG of material from LCF parameters is verified and discussed.

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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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Effective Strengthening and Toughening in High Entropy-Alloy by Combining Extrusion Machining and Heat Treatment

SSRN

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SSRN 2022年

作者： Pu, Zhuo Cai, Songlin L. Dai, Lan-Hong Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610200 China State Key Laboratory of Nonlinear Mechanics Institute of Mechanics Chinese Academy of Sciences Beijing100190 China School of Engineering Science University of Chinese Academy of Sciences Beijing100049 China CAS Center for Excellence in Complex System Mechanics Beijing100190 China China Electric Power Research Institute State Grid Corporation of China Beijing100192 China

This study presents an effective strengthening and toughening approach to improve tensile properties of CrMnFeCoNi high entropy-alloy (HEA) by combining extrusion machining and short-time heat treatment. After such process, preliminary structural heterogeneity has been achieved in this alloy, of which the strength is significantly enhanced while good elongation is retained compared to its homogeneous counterparts. Microstructural characterization reveals varying degrees of partial recrystallization with more or less dislocation density for HEAs under different treatment. It is such structural inhomogeneity that imparts this alloy with good strain hardening ability at high flow stress levels. © 2022, The Authors. All rights reserved.

关键词： Strain hardening

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Thermal-Hydraulic Characteristics of Helical Cruciform Single Rod Based on Cfd Investigation

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SSRN 2024年

作者： Chen, Yiwen Zhang, Dalin Jiang, Dianqiang Deng, Jian Zhang, Xisi Wang, Xinan Tian, Wenxi Qiu, Suizheng Su, Guanghui State Key Laboratory of Multiphase Flow in Power Engineering Shaanxi Key Laboratory of Advanced Nuclear Energy and Technology School of Nuclear Science and Technology Xi’an Jiaotong University Xi’an710049 China Nuclear Power Institute of China 328 Changshun Avenue Chengdu610213 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China China Institute of Atomic Energy Beijing611731 China

To increase power density in nuclear reactors, the helical cruciform fuel (HCF) is designed as a fundamental element. HCF takes advantages of strong mixing effects, self-supporting features, and strengthened heat transfer, potential to increase the fuel performance in nuclear reactors. In this paper, friction factor and heat transfer coefficient empirical correlations of HCF single rod cooled by water were proposed using a geometric dimensionless number based on numerical simulation. The velocity field, and temperature distribution were obtained to discuss influences of geometric sizes on the fuel performance. Finally, correlations with a geometric dimensionless number were validated with various cases, whose application conditions were also proposed and discussed. This paper provides a guide for study of heat transfer and fluid flow in helical structures, and a reference for establishment of reactor system model using helical cruciform fuels. © 2024, The Authors. All rights reserved.

关键词： Heat transfer coefficients

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Research on Optimization of Key Parts of Nuclear reactor Based on Finite Element Simulation

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IOP Conference Series: Earth and Environmental science 2021年第3期632卷

作者： Bo Yang Tianda Yu Qiang Deng Cong Chen Fengshou Zhang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institution of China Sichuan ChengDu 610041 China

In view of the higher technical requirements of the third generation PWR for the control rod drive mechanism (CRDM), China's third generation nuclear reactor CRDM hook parts are designed to multi-tooth hook with wear-resistant surfacing. The surfacing welding process is simulated and optimized based on finite element simulation, and reasonable welding process and parameters are determined by simulation results in this *** to the optimized welding process, the parts are trial-produced, and the parts of the hook are subjected to metallographic inspection, hardness test and thermal life test. The results show that the optimized hook parts have high hardness and wear resistance, and meet the requirements of the third generation PWR nuclear power plant.

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A series of power reconstruction methods based on solving diffusion equation approximately in node

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Hedongli Gongcheng/Nuclear Power Engineering 2010年第SUPPL. 2期31卷 63-67页

作者： Zhao, Wen-Bo Hu, Yong-Ming Wang, Kan Yao, Dong Zhang, Guo-Shu Feng, Kai-Ming Department of Engineering Physics Tsinghua University Beijing 100084 China Institute of Nuclear and New Energy Technology Tsinghua University Beijing 100084 China Science and Technology on Reactor System Design Technology Laboratory Chengdu 610041 China

Based on approximately solving diffusion equation in node, a series of reconstruction methods are developed for NGFMN code. These methods are categorized by hyperbolic functions, boundary conditions and the order of Legendre polynomials utilized in equation solving. 2D IAEA and 2D BIBLIS benchmarks have been solved. The numerical results are compared with those from CITATION, and the reconstruction method with smaller errors is selected.

关键词： Hyperbolic functions

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Characterization of the Defects in Oxide Dispersion-Strengthened Iron Alloy after Helium Ion Irradiation

Characterization of the Defects in Oxide Dispersion-Strength...

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IEEE Symposium on Fusion Engineering

作者： P. L. Gao X. Ju Y. Xin Q. S. Cao Z. M. Guo L. P. Guo B. Y. Wang China Academy of Engineering Physics Institute of Nuclear Physics and Chemistry Mianyang China Department of Physics University of Science and Technology Beijing Beijing China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute China Chengdu China Institute of Power Metallurgy University of Science and Technology Beijing Beijing China Accelerator Laboratories School of Physics Wuhan University Wuhan China Key Laboratory of Nuclear Analysis Techniques Institute of High Energy Physics Chinese Academy of Sciences Beijing China

Vacancy-type defects in Fe-2%A1_2O_3 (ODS-Fe) and pure Fe produced by helium ion irradiation at room temperature were investigated using positron beam Doppler broadening energy spectra (DBES). Defect profiles of the S parameter derived from DBES as a function of positron incident energy up to 20 KeV were analyzed. The S parameter values of the damaged layer for the ODS-Fe are smaller than those of pure Fe for irradiation dose of le+16/cm~2. This finding indicates that ODS-Fe has a higher irradiation resistance than pure Fe. It also suggests that helium irradiation induced point defects in ODS-Fe were trapped and annihilated at the interfaces between the Fe matrix and the A1_2O_3 nanoparticles. The S-W curves indicate that only one type of defect was formed in the post-irradiated ODS-Fe. Vacancy clusters and helium-vacancy complexes were identified as the main defects. The defects are complicated for the irradiated pure Fe, for which the surface defects are different from the ones inside bulk.

关键词： oxide dispersion-strengthened alloy positron annihilation spectroscopy ion irradiation defects

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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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