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学者姓名:吴宏春

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< Page ,Total 35 >
Validation of SARAX for the China Fast Reactor with the extrapolated experimental data EI
期刊论文 | 2019 , 127 , 188-195 | Annals of Nuclear Energy
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Abstract :

The validation works have been implemented to a newly-developed code SARAX for China Fast Reactor (CFR) in this paper. Different with the conventional way to validate the code using the dedicated experimental data, a theoretical approach was proposed and implemented by using the existing similar experimental data. This theoretical approach is based on the technology of sensitivity/uncertainty analysis, similarity analysis and nuclear-data adjustment. In our previous works, detailed introduction towards sensitivity/uncertainty analysis and similarity analysis have been implemented to distinguish the existing experiments which are 'similar’ to the CFR's criticality characteristics. This paper focus on the method to extrapolate the 'similar’ experimental data to predict the best-estimate measurements of CFR with application of nuclear-data adjustment, providing reference values to validate the SARAX code. From the numerical results, it can be observed that through nuclear-data adjustment, the simulation results of the existing experiments can agree well with corresponding measurements, with the bias reduced notably to be within 25 pcm. Moreover, the nuclear-data adjustment can also improve the nuclear-data uncertainties notably, with the relative uncertainties of the reactor keff due to nuclear data having been reduced from the value exceeding 1.0% to the values about 0.15%. © 2018

Keyword :

Code validation Nuclear data Numerical results Reference values Relative uncertainty Sensitivity/uncertainty analysis Similarity analysis Theoretical approach

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GB/T 7714 Wan, Chenghui , Qiao, Liang , Zheng, Youqi et al. Validation of SARAX for the China Fast Reactor with the extrapolated experimental data [J]. | Annals of Nuclear Energy , 2019 , 127 : 188-195 .
MLA Wan, Chenghui et al. "Validation of SARAX for the China Fast Reactor with the extrapolated experimental data" . | Annals of Nuclear Energy 127 (2019) : 188-195 .
APA Wan, Chenghui , Qiao, Liang , Zheng, Youqi , Cao, Liangzhi , Wu, Hongchun . Validation of SARAX for the China Fast Reactor with the extrapolated experimental data . | Annals of Nuclear Energy , 2019 , 127 , 188-195 .
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Heterogeneous discontinuity factor treatment in Variational Nodal Method EI
期刊论文 | 2019 , 127 , 341-350 | Annals of Nuclear Energy
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Abstract :

To handle the control rod cusping effect in Pressurized Water Reactor (PWR) core calculation, the heterogeneous Variational Nodal Method (VNM) employed by the fuel management calculation code system NECP-Bamboo has been enhanced to treat heterogeneous discontinuity factor (DF) appearing on nodal interface. To solve the neutron-diffusion equations with heterogeneous DF, firstly, the functional in VNM is modified to contain the discontinuity of neutron flux in the surface integral term. Secondly, other than volumetric flux and surface partial currents, cross sections and surface DF are also expanded into the sum of orthogonal piece-wise polynomials to construct the nodal response matrixes. Four test problems in this paper including the CISE, Henry-Worley, HAFAS benchmark problems and the BEAVRS problems were employed to verify the method in treating heterogeneous DF. It has been demonstrated that the control rod cusping effect can be fully eliminated by the heterogeneous VNM with heterogeneous DF in terms of control rod differential worth and three-dimensional pin-power profile. © 2018 Elsevier Ltd

Keyword :

Bench-mark problems Core calculations Differential worth Discontinuity factor Neutron diffusion equations Partial currents Surface integrals Variational nodal method

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GB/T 7714 Li, Yunzhao , Liang, Boning , Wu, Hongchun et al. Heterogeneous discontinuity factor treatment in Variational Nodal Method [J]. | Annals of Nuclear Energy , 2019 , 127 : 341-350 .
MLA Li, Yunzhao et al. "Heterogeneous discontinuity factor treatment in Variational Nodal Method" . | Annals of Nuclear Energy 127 (2019) : 341-350 .
APA Li, Yunzhao , Liang, Boning , Wu, Hongchun , Li, Zhipeng , Yang, Jiewei . Heterogeneous discontinuity factor treatment in Variational Nodal Method . | Annals of Nuclear Energy , 2019 , 127 , 341-350 .
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The material-region-based 2D/1D transport method EI
期刊论文 | 2019 , 128 , 1-11 | Annals of Nuclear Energy
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The 2D/1D transport method is the dominant method for high-fidelity direct whole-core transport calculations, which attracts a lot of attention in recent decades. In the 2D/1D method, some sources of deviation are introduced, including spatial and angular approximation of leakage term and cross-section homogenization for 1D axial calculation. These approximations are analyzed and a material-region-based 2D/1D transport method with anisotropic leakage term, which avoids cross-section homogenization, as well as reduces spatially flat leakage approximation, are developed to improve accuracy at the expense of acceptable memory and efficiency loss. Finally, a BWR assembly case, the C5G7 benchmark and a rod-cluster assembly case are tested to verify the accuracy and performance of the material-region-based 2D/1D method. © 2018 Elsevier Ltd

Keyword :

Cluster assembly Efficiency loss High-fidelity Leakage terms Material region NECP-X Transport method Whole-core transport

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GB/T 7714 Liu, Zhouyu , Zhao, Chen , Cao, Lu et al. The material-region-based 2D/1D transport method [J]. | Annals of Nuclear Energy , 2019 , 128 : 1-11 .
MLA Liu, Zhouyu et al. "The material-region-based 2D/1D transport method" . | Annals of Nuclear Energy 128 (2019) : 1-11 .
APA Liu, Zhouyu , Zhao, Chen , Cao, Lu , Wu, Hongchun , Cao, Liangzhi . The material-region-based 2D/1D transport method . | Annals of Nuclear Energy , 2019 , 128 , 1-11 .
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Analysis and improvement of global-local self-shielding calculation scheme for aic control rods Scopus
会议论文 | 2018 , 3
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Abstract :

Copyright © 2018 ASME. Global-local self-shielding calculation scheme is a new high-fidelity resonance calculation model proposed by NECP laboratory of Xi’an Jiaotong University. Neutron Current Method (NCM) is utilized for resonance calculation in the global aspect to obtain Dancoff factors. Then each fuel pin is transformed into individual 1D cylindrical problems by conserving Dancoff factors. The Pseudo-Resonant-Nuclide Subgroup Method (PRNSM) is used to conduct resonance calculation in the local aspect for each 1D cylindrical pin. Global-local self-shielding calculation scheme has been successfully implemented in high-fidelity numerical nuclear reactor physics code NECP-X. Verification results of global-local self-shielding calculation scheme showed good accuracy for UO2 fuels. The maximum relative error of microscopic absorption cross sections (XSs) for 238U in resonance range was 1.5% compared with MCNP5 [1]. AIC control rods serve as strong absorbers in reactor. Strong self-shielding phenomenon occurs when AIC control rods are inserted. Analysis was performed to determine the effects of AIC control rods on the accuracy of global-local self-shielding calculation scheme and the sources of error. Evaluation results showed that the main part of error was introduced by NCM and radius searching. The relative errors were larger than 10% in several resonance groups. Therefore, a supercell model is proposed to couple with global-local self-shielding calculation scheme to treat resonance calculation for AIC control rods in this paper. Numerical results show that this model improves the accuracy of the global-local self-shielding calculation scheme. The relative errors of microscopic absorption XSs for AIC in most resonance groups were decreased to less than 2%.

Keyword :

AIC Cross section Resonance Supercell

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GB/T 7714 Li, Jikui , Zu, Tiejun , Cao, Liangzhi et al. Analysis and improvement of global-local self-shielding calculation scheme for aic control rods [C] . 2018 .
MLA Li, Jikui et al. "Analysis and improvement of global-local self-shielding calculation scheme for aic control rods" . (2018) .
APA Li, Jikui , Zu, Tiejun , Cao, Liangzhi , Wu, HongChun , He, Qingming . Analysis and improvement of global-local self-shielding calculation scheme for aic control rods . (2018) .
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Fast sub-grid scale finite element method for the first order neutron transport equation Scopus
会议论文 | 2018 , 3
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Copyright © 2018 ASME. This paper presents a fast sub-grid scale (SGS) finite element method for the first order neutron transport equation. The spherical harmonics method is adopted for the angular discretization. The sub-grid scale discretization embeds discontinuous component in each element to provide a stabilization term for the continuous finite element formulation. Traditional SGS method uses Riemann decomposition and vacuum boundary assumption to decouple the discontinuous component. Here we propose a new method to perform the decoupling based on the assumption that the convection term of the discontinuous component is proportional to the residual of angular flux in each element. The computing costs for the establishment of the coefficient matrix of discontinuous component are reduced to O(1) from O(n3). Further more, the computing costs for the inversion of the coefficient matrix are reduced to O(n) from O(n3) by applying mass lumping technique. Numerical results show that the new method is not only more efficient but also yields more accurate solution than traditional sub-grid scale method.

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GB/T 7714 Fang, Chao , Wu, Hongchun , Cao, Liangzhi et al. Fast sub-grid scale finite element method for the first order neutron transport equation [C] . 2018 .
MLA Fang, Chao et al. "Fast sub-grid scale finite element method for the first order neutron transport equation" . (2018) .
APA Fang, Chao , Wu, Hongchun , Cao, Liangzhi , Li, Yunzhao . Fast sub-grid scale finite element method for the first order neutron transport equation . (2018) .
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Development and verification of a new nuclear data processing system NECP-Atlas Scopus
会议论文 | 2018 , 3
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Copyright © 2018 ASME. To process the evaluated nuclear data file (ENDF) libraries and generate the cross section data library for neutronics calculations, a new nuclear data processing system NECP-Atlas was developed by Nuclear Engineering Computational Physics Lab. of Xi’an Jiaotong University. Meanwhile, some flaws of the current widely used nuclear data processing systems were made up. Some new methods and techniques were proposed and integrated into NECP-Atlas. NECP-Atlas could process ENDF and generate point-wise evaluated nuclear data file (PENDF) and the multigroup cross section data library in WIMS-D format. Verification of NECP-Atlas was carried out by comparing the keff values for WLUP benchmark cases and benchmark experiments in the ICSBEP handbook using cross section data libraries processed by NECP-Atlas with those by NJOY2016. The results showed that NECP-Atlas processes the ENDF correctly and generates more reliable cross section data libraries.

Keyword :

Benchmark Cross section data library Evaluated nuclear data file Nuclear data processing

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GB/T 7714 Xu, Jialong , Zu, Tiejun , Cao, Liangzhi et al. Development and verification of a new nuclear data processing system NECP-Atlas [C] . 2018 .
MLA Xu, Jialong et al. "Development and verification of a new nuclear data processing system NECP-Atlas" . (2018) .
APA Xu, Jialong , Zu, Tiejun , Cao, Liangzhi , Wu, Hongchun . Development and verification of a new nuclear data processing system NECP-Atlas . (2018) .
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Development of an optimized transport solver in SARAX for fast reactor analysis Scopus
会议论文 | 2018 , 9
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Copyright © 2018 ASME Although the neutron mean free path in typical fast reactors is long compared with that of typical thermal reactors, the heterogeneous flux distributions in angular domain and spatial domain are notable due to the strong elastic-scattering resonance and common core-level heterogeneous layout. Full-core multigroup transport analysis becomes inevitable for advanced fast reactor R&D. A fast reactor neutronics analysis system, SARAX is under development at XJTU of China. An existing in-house SN-nodal solver in triangular-Z geometry, DNTR was primarily chosen for its geometric adaptability. However, insufferable problems of computing time and storage were encountered when the solver was on active service. In this paper, the problems are firstly analysed and resolved from the theory and code levels. Then, widely used CMFD acceleration method with some stabilization techniques is implemented for the SN-nodal method in triangular-Z geometry, which can largely reduce the computing time. What's more, a new acceleration method TCD is proposed and can obtain superior speedups with reasonable accuracy sacrifice. The updated solver, DNTR 1.1 has been developed based on these improvements, which can obtain dozens of speedup with reduced storage compared with DNTR for typical fast reactor simulation on a desktop computer.

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GB/T 7714 Xu, Zhitao , Wu, Hongchun , Zheng, Youqi et al. Development of an optimized transport solver in SARAX for fast reactor analysis [C] . 2018 .
MLA Xu, Zhitao et al. "Development of an optimized transport solver in SARAX for fast reactor analysis" . (2018) .
APA Xu, Zhitao , Wu, Hongchun , Zheng, Youqi , He, Mingtao . Development of an optimized transport solver in SARAX for fast reactor analysis . (2018) .
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Multi-dimensional heterogeneous resonance integral tables generated for embedded self-shielding method towards irregular lattices Scopus
会议论文 | 2018 , 3
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Copyright © 2018 ASME. The concept of multi-dimensional heterogeneous resonance integral tables is proposed. The new type of resonance integral is designed for different fuel pins appearing in one lattice with two extra dimension of optical radius and number density ratio in the fuel. Numerical results show that this treatment improves the accuracy of embedded self-shielding method on irregular lattices.

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GB/T 7714 Zhang, Qian , Zhao, Qiang , Wu, Hongchun et al. Multi-dimensional heterogeneous resonance integral tables generated for embedded self-shielding method towards irregular lattices [C] . 2018 .
MLA Zhang, Qian et al. "Multi-dimensional heterogeneous resonance integral tables generated for embedded self-shielding method towards irregular lattices" . (2018) .
APA Zhang, Qian , Zhao, Qiang , Wu, Hongchun , Cao, Liangzhi , Zheng, Zheng . Multi-dimensional heterogeneous resonance integral tables generated for embedded self-shielding method towards irregular lattices . (2018) .
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Development and application of a 2D/1D fusion code with leakage reconstruction method Scopus
会议论文 | 2018 , 3
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Copyright © 2018 ASME. The 2D/1D fusion method (2D/1D method) is becoming a popular transport method for whole-core calculations, which reduces the group condense and assembly homogenization approximations in the conventional two-step reactor physics calculations. In most 2D/1D codes, a pin is chosen as a 1D calculation domain, which assumes that the axial leakage of the pin is flat on top/bottom surfaces. Similar to the axial leakage, the radial leakage of every 2D plane also introduces several approximations along axial direction for the 1D calculation. In this paper a 2D/1D fusion code is developed, while a leakage reconstruction method is proposed and applied. In this 2D/1D fusion code, MOC is applied to the radial 2D calculation and the Sn diamond difference method is used for the axial 1D calculation. Numerical results indicate that the 2D/1D fusion code developed in this paper is precise in three-dimensional transport calculation and show the performance of this leakage reconstruction method especially when the leakage term is significant.

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GB/T 7714 Liang, Liang , Liu, Zhouyu , Wu, Hongchun et al. Development and application of a 2D/1D fusion code with leakage reconstruction method [C] . 2018 .
MLA Liang, Liang et al. "Development and application of a 2D/1D fusion code with leakage reconstruction method" . (2018) .
APA Liang, Liang , Liu, Zhouyu , Wu, Hongchun , Wang, Sheng , Zhang, Qian , Zhao, Qiang . Development and application of a 2D/1D fusion code with leakage reconstruction method . (2018) .
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Transient analysis for liquid-fuel molten salt reactor based on MOREL2.0 code EI SCIE Scopus
期刊论文 | 2018 , 42 (1) , 261-275 | INTERNATIONAL JOURNAL OF ENERGY RESEARCH
SCOPUS Cited Count: 1
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Abstract :

A molten salt reactor (MSR) is characterized by simultaneously using liquid fuel salt as both the nuclear fuel and coolant. The redistribution of delayed neutron precursors (DNPs) makes the transient behavior of MSRs different from traditional solid-fuel reactors. In this study, a 3D coupled neutronics/thermal hydraulics code, MOREL2.0, was employed to analyze a liquid-fuel Thorium Molten Salt Reactor (TMSR-LF) under perturbations of fuel pump start-up and coast-down and by overheating and overcooling the inlet fuel temperature. Some transient processes were simulated to provide guidance for the future design and optimization of TMSR-LFs. In response to the perturbations, reactivity was lost and gained in the pump start-up and coast-down, respectively. Overheating the inlet fuel temperature introduced negative reactivity, and TMSR-LF maintained a safety state, while overcooling the inlet fuel temperature resulted in positive reactivity. Overcooling by 70 K produced a supercritical transient condition and a rapid increase in power within a short period, which was followed by a decrease in power due to negative temperature feedback. The transient results demonstrate that the negative temperature feedback coefficients guarantee TMSR-LF inherent safety and the variation range of temperature stay within the safety margin.

Keyword :

TMSR transient thermal-hydraulics neutronics molten salt reactor

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GB/T 7714 Cao, Liangzhi , Zhuang, Kun , Zheng, Youqi et al. Transient analysis for liquid-fuel molten salt reactor based on MOREL2.0 code [J]. | INTERNATIONAL JOURNAL OF ENERGY RESEARCH , 2018 , 42 (1) : 261-275 .
MLA Cao, Liangzhi et al. "Transient analysis for liquid-fuel molten salt reactor based on MOREL2.0 code" . | INTERNATIONAL JOURNAL OF ENERGY RESEARCH 42 . 1 (2018) : 261-275 .
APA Cao, Liangzhi , Zhuang, Kun , Zheng, Youqi , Hu, Tianliang , Wu, Hongchun . Transient analysis for liquid-fuel molten salt reactor based on MOREL2.0 code . | INTERNATIONAL JOURNAL OF ENERGY RESEARCH , 2018 , 42 (1) , 261-275 .
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