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学者姓名：吴宏春
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Abstract ：
The high-fidelity whole-core heterogeneous transport calculation attracts a lot of attention because of its high accuracy. However, too many energy-groups for the transport calculation lead to more computational time. This paper presents a new online energy group condensation (OGC) method to accelerate the whole-core heterogeneous transport calculation. Three different models are established for the fuel pin-cell, none-fuel pin-cell and reflector regions, to obtain the spectrum. For the fuel pin-cell regions, a one-dimension (1D) cylinder model is generated during the global-local resonance calculations, and the spectrum is calculated based on the equivalent 1D model and the macro cross-section is condensed based on this spectrum. For the none-fuel pin-cell regions, a super-cell model is established with this none-fuel pin-cell and its surrounding fuel, and the spectrum is calculated with the super-cell model. For the reflector regions, the fuel-lattice-reflector models are generated to calculate spectrum in reflector regions offline, and the spectrum is built-in for condensing the cross-sections for all reflector regions. The condensed energy group structure is studied with the particle swarm optimization (PSO) method to get a superior energy group structure with fewer group numbers but preserving accuracy. A set of benchmark problems are tested and the results show good performance of the new online group condensation method for the transport calculations. (C) 2021 Elsevier Ltd. All rights reserved.
Keyword ：
NECP-X Particle swarm optimization The online energy group condensation The whole-core heterogeneous transport calculations
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| GB/T 7714 | Zhou, Xinyu , Liu, Zhouyu , Cao, Liangzhi et al. A new online energy group condensation method for the high-fidelity neutronics code NECP-X [J]. | ANNALS OF NUCLEAR ENERGY , 2021 , 158 . |
| MLA | Zhou, Xinyu et al. "A new online energy group condensation method for the high-fidelity neutronics code NECP-X" . | ANNALS OF NUCLEAR ENERGY 158 (2021) . |
| APA | Zhou, Xinyu , Liu, Zhouyu , Cao, Liangzhi , Wu, Hongchun , Zhai, Zian . A new online energy group condensation method for the high-fidelity neutronics code NECP-X . | ANNALS OF NUCLEAR ENERGY , 2021 , 158 . |
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Abstract ：
The substantial time required for solving the neutron transport equation can be reduced through parallel calculation. The block-Jacobi algorithm is a common method for parallel neutron transport calculation in an unstructured mesh; however, iteration degradation is an inevitable problem that limits the application of this algorithm by reducing the parallel efficiency. In this study, we applied the block-Jacobi algorithm to the SN nodal method with a triangular-z mesh and proposed an improvement to achieve high parallel efficiency. The interface prediction (IP) method was developed to prevent iteration degradation. This method is based on extrapolating the interface information instead of using the information from the preceding iteration at the interfaces in sub-domains. Meanwhile, the prediction was used recursively to accelerate the self-group scattering source iteration in the entire space; this process is referred to as the inner iteration prediction method (IIP). These two methods effectively prevent iteration degradation and reduce time required for self-group scattering source iteration. A more stable and improved parallel performance was thus achieved. © 2021 Elsevier Ltd
Keyword ：
Efficiency Forecasting Iterative methods Mesh generation Neutron flux Transport properties
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| GB/T 7714 | Qiao, Liang , Zheng, Youqi , Wu, Hongchun et al. Improved block-Jacobi parallel algorithm for the SN nodal method with unstructured mesh [J]. | Progress in Nuclear Energy , 2021 , 133 . |
| MLA | Qiao, Liang et al. "Improved block-Jacobi parallel algorithm for the SN nodal method with unstructured mesh" . | Progress in Nuclear Energy 133 (2021) . |
| APA | Qiao, Liang , Zheng, Youqi , Wu, Hongchun , Wang, Yongping , Du, Xianan . Improved block-Jacobi parallel algorithm for the SN nodal method with unstructured mesh . | Progress in Nuclear Energy , 2021 , 133 . |
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Abstract ：
The thermal neutron scattering cross sections of zirconium hydride (ZrHx) is heavily affected by its lattice structure which can be represented by phonon density of states (DOS). In the past decades, many works tried to get parameterized phonon DOS of ZrHx, by fitting certain types of experimental results. In the present work, we adopt the first-principles calculation to obtain the phonon DOS of ZrH(1.5 )in delta phase and ZrH2 in epsilon phase. The theoretical phonon DOS is used to calculate the thermal scattering cross sections which are then used in the neutronics simulations of several TRIGA reactors. The numerical results show that the phonon DOS obtained by first-principles calculations can produce more accurate scattering cross sections, and improve the neutronics results of TRIGA reactors compared with the phonon DOS models applied in ENDF/B and JEFF evaluated nuclear data libraries. (C) 2021 Elsevier Ltd. All rights reserved.
Keyword ：
First-principles NECP-Atlas Thermal scattering cross section Thermal scattering law Zirconium hydride
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| GB/T 7714 | Zu, Tiejun , Tang, Yongqiang , Wang, Lipeng et al. Thermal scattering law data generation for hydrogen bound in zirconium hydride based on the phonon density of states from first-principles calculations [J]. | ANNALS OF NUCLEAR ENERGY , 2021 , 161 . |
| MLA | Zu, Tiejun et al. "Thermal scattering law data generation for hydrogen bound in zirconium hydride based on the phonon density of states from first-principles calculations" . | ANNALS OF NUCLEAR ENERGY 161 (2021) . |
| APA | Zu, Tiejun , Tang, Yongqiang , Wang, Lipeng , Cao, Liangzhi , Wu, Hongchun . Thermal scattering law data generation for hydrogen bound in zirconium hydride based on the phonon density of states from first-principles calculations . | ANNALS OF NUCLEAR ENERGY , 2021 , 161 . |
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Abstract ：
Numerical reactor based on high fidelity model and method is with the characteristics of high precision and high resolution, but the inherent uncertainty of nuclear data and other parameters will seriously affect the uncertainty of numerical reactor analysis results. Based on the review of the research progresses of numerical reactors and its uncertainty quantification in the world, this paper focuses on the research progresses in NECP Laboratory of Xi'an Jiaotong University in recent years, including the development of one-step high fidelity numerical reactor program NECP-X, the generation of nuclear data covariance database, the uncertainty propagation method based on deterministic method and sampling method, and the uncertainty quantification in transient calculation. An advanced sampling method COST is proposed. Based on the high fidelity numerical reactor program, the uncertainty propagation of covariance of various nuclear parameters in the steady-state and transient analysis of the reactor core is quantified for the first time, which is of great significance for engineering application of numerical reactors. © 2021, Editorial Board of Journal of Nuclear Power Engineering. All right reserved.
Keyword ：
Application programs Nuclear reactors Numerical methods Transient analysis Uncertainty analysis
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| GB/T 7714 | Cao, Liangzhi , Zou, Xiaoyang , Liu, Zhouyu et al. Research Progresses of Uncertainty Quantification Methods for High Fidelity Numerical Nuclear Reactor [J]. | Nuclear Power Engineering , 2021 , 42 (2) : 1-15 . |
| MLA | Cao, Liangzhi et al. "Research Progresses of Uncertainty Quantification Methods for High Fidelity Numerical Nuclear Reactor" . | Nuclear Power Engineering 42 . 2 (2021) : 1-15 . |
| APA | Cao, Liangzhi , Zou, Xiaoyang , Liu, Zhouyu , Wan, Chenghui , Wu, Hongchun . Research Progresses of Uncertainty Quantification Methods for High Fidelity Numerical Nuclear Reactor . | Nuclear Power Engineering , 2021 , 42 (2) , 1-15 . |
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Abstract ：
In this paper, the transient pin-power reconstruction based on source-expansion method has been proposed. The transient calculation is sufficient to simulate the reactor-physics process in second-time scale, such as the load-dump process and dynamic rod worth measurement. During the transient process, the pin-power distribution is essential for the determination of the power-peak factors, including Fq and FAH. However, with the widely-applied transverse-integrated nodal methods for transient simulation, the detailed pin-averaged results can't be provided directly. To address this problem, the transient pin power reconstruction was proposed. It was derived from the transient fixed-source problem assuming the total-source and the transient fixed-source terms could be respectively expanded by the fourth order Legendre polynomials and the biquadratic Legendre polynomials constructed from the nine-node problem. Moreover, to improve the accuracy of pin-averaged results, the corner-point condition and corner-discontinuity factors were employed in the reconstruction process. This proposed transient pin power reconstruction method has been implemented in our self-developed code named SPARK and verified with several benchmarks. Results of the verification indicated that this proposed method for transient pin-power reconstruction can provide a satisfactory accuracy in transient process. (C) 2021 Elsevier Ltd. All rights reserved.
Keyword ：
Source-expansion method The SPARK code Transient pin-power reconstruction Transient simulation
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| GB/T 7714 | Bai, Jiahe , Wan, Chenghui , Li, Yunzhao et al. A Source-expansion-based method for transient Pin-Power reconstruction [J]. | ANNALS OF NUCLEAR ENERGY , 2021 , 164 . |
| MLA | Bai, Jiahe et al. "A Source-expansion-based method for transient Pin-Power reconstruction" . | ANNALS OF NUCLEAR ENERGY 164 (2021) . |
| APA | Bai, Jiahe , Wan, Chenghui , Li, Yunzhao , Wu, Hongchun , Li, Fan . A Source-expansion-based method for transient Pin-Power reconstruction . | ANNALS OF NUCLEAR ENERGY , 2021 , 164 . |
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Abstract ：
In this paper, the ex-core source-range detector response was simulated for AP1000 cores in its fuel-loading process by using a Monte Carlo code. Firstly, the current sensitivity coefficients of the ex-core source-range detectors have been calibrated by utilizing the experiments with single fuel assembly and the primary source loaded in the core. Secondly, both relative and absolute calibration methods have been applied and analyzed for the sensitivity coefficients calibration. Thirdly, the simulation values for the ex-core source-range response with fully loaded fuel have been validated by using the actual measurement of the first AP1000 cores in the world. Fourthly, the contribution of the 238U spontaneous fission to the ex-core source-range detector response has been studied. Through the numerical simulation results, the following observations can be obtained: 1) The relative calibration method is better than the absolute calibration method, especially when the ex-core source-range response is small. 2) Compared with the measurements, the relative deviation between the simulated and the measured values is within 14% for the two AP1000 cores at Sanmen Nuclear Power Plant. 3) The intensity of 238U spontaneous fission source accounts for 15% of the total source strength in the AP1000 core Unit 1, but the maximum change of response is only 0.02 cps, which is still insignificant for the ex-core detector response simulation in PWR. © 2020 Elsevier B.V.
Keyword ：
Calibration Nuclear fuels Nuclear power plants Numerical methods Pressurized water reactors
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| GB/T 7714 | Sun, Bin , Li, Yunzhao , Cao, Liangzhi et al. Simulations of the source-range detector response for the fuel-loading process of the AP1000 cores [J]. | Nuclear Engineering and Design , 2021 , 372 . |
| MLA | Sun, Bin et al. "Simulations of the source-range detector response for the fuel-loading process of the AP1000 cores" . | Nuclear Engineering and Design 372 (2021) . |
| APA | Sun, Bin , Li, Yunzhao , Cao, Liangzhi , Li, Xuesong , Wu, Hongchun , Shen, Wei et al. Simulations of the source-range detector response for the fuel-loading process of the AP1000 cores . | Nuclear Engineering and Design , 2021 , 372 . |
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Abstract ：
In this paper, the method research and engineering application of the B10-abundance correction for PWR have been implemented. The boric acid is one of the main neutron-absorber materials, within which the B10 is the most significant isotope for absorbing the thermal neutron. Due to the depletion effect and boronation effect, the B10 abundance would variate with the time. These effects would result in the phenomenon that the calculation values of the critical boron concentrations (CBC) provided by nuclear-design code would be smaller than corresponding measurement values. Therefore, for reducing the calculation errors of CBC, the B10-abundance correction method has been proposed based on the Bamboo-C code. Applying the B10-abundance correction method, the engineering validation has been implemented, comparing the calculation values of CBC and B10 abundances with corresponding measurement values. The numerical results indicated that the proposed correction method for B10 abundance can notably improve the calculation accuracy.
Keyword ：
B10-abundance correction method Bamboo-C Boronation effect Depletion effect
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| GB/T 7714 | Wan, Chenghui , Bai, Jiahe , Liu, Yu et al. Method research and engineering application of the B10-abundance correction for PWR [J]. | NUCLEAR ENGINEERING AND DESIGN , 2021 , 378 . |
| MLA | Wan, Chenghui et al. "Method research and engineering application of the B10-abundance correction for PWR" . | NUCLEAR ENGINEERING AND DESIGN 378 (2021) . |
| APA | Wan, Chenghui , Bai, Jiahe , Liu, Yu , Huang, Xing , Wu, Hongchun . Method research and engineering application of the B10-abundance correction for PWR . | NUCLEAR ENGINEERING AND DESIGN , 2021 , 378 . |
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Abstract ：
A new module designed to generate thermal scattering law (TSL) data is developed in the nuclear data processing code NECP-Atlas, which is currently under development at Xi'an Jiaotong University. The module can calculate TSL data for solid materials of any crystal structure and liquid materials in the free gas or the Egelstaff and Schofield approximations. To acquire a precise description of the neutron thermalization, the module eliminates the cubic approximation and atom site approximation used in the conventional method. In addition, a technique based on the anisotropic displacement parameters method is implemented to consider the anisotropy of interatomic forces and the correlation of forces from different directions. Comparison with the LEAPR and the Debye–Waller matrix method used in ENDF/B-VIII.0 enabled verification of the TSL data generated by the module. The numerical results revealed the good performance of the module. © 2020 Elsevier Ltd
Keyword ：
Anisotropy Crystal structure Data handling Neutron scattering
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| GB/T 7714 | Tang, Yongqiang , Zu, Tiejun , Yi, Siyu et al. Development and verification of thermal neutron scattering law data calculation module in nuclear data processing code NECP-Atlas [J]. | Annals of Nuclear Energy , 2021 , 153 . |
| MLA | Tang, Yongqiang et al. "Development and verification of thermal neutron scattering law data calculation module in nuclear data processing code NECP-Atlas" . | Annals of Nuclear Energy 153 (2021) . |
| APA | Tang, Yongqiang , Zu, Tiejun , Yi, Siyu , Cao, Liangzhi , Wu, Hongchun . Development and verification of thermal neutron scattering law data calculation module in nuclear data processing code NECP-Atlas . | Annals of Nuclear Energy , 2021 , 153 . |
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Abstract ：
The environment effect arises when pin-cell homogenized parameters are generated with reflective boundary conditions. To treat it in whore-core pin-by-pin calculation, two works are summarized in this article. Firstly, by analyzing the relative errors of pin-cell homogenized group constants and the relative importance of pin-cell discontinuity factors (PDF) in each group, the importance of correcting the PDF of the thermal group is recognized. Secondly, the least-squares method for a multivariate polynomial is utilized to functionalize the relation of the thermal group PDF and the core parameters, including diffusion coefficient, removal cross-section, neutron source, and normalized surface flux. The C5G7 and KAIST benchmarks are employed to evaluate the performance of the PDF predication. Numerical results indicate its effectiveness in reducing the errors of eigenvalue and pin power, especially for the cases with the fuel pins located near the interface between different assemblies.
Keyword ：
environment effect functionalization least-squares method pin-by-pin PWR
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| GB/T 7714 | Zhang, Bin , Li, Yunzhao , Wu, Hongchun . Environment Effect Treatments in PWR Whole-Core Pin-by-Pin Calculation [J]. | FRONTIERS IN ENERGY RESEARCH , 2021 , 9 . |
| MLA | Zhang, Bin et al. "Environment Effect Treatments in PWR Whole-Core Pin-by-Pin Calculation" . | FRONTIERS IN ENERGY RESEARCH 9 (2021) . |
| APA | Zhang, Bin , Li, Yunzhao , Wu, Hongchun . Environment Effect Treatments in PWR Whole-Core Pin-by-Pin Calculation . | FRONTIERS IN ENERGY RESEARCH , 2021 , 9 . |
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Abstract ：
The latest CENDL and ENDF/B evaluated nuclear data libraries was released in 2020 and 2018, respectively. To apply CENDL-3.2 and ENDF/B-VIII.0 in the reactor physic simulations of pressurized water reactor (PWR), the CNP-1000 PWR, which is an improved GEN-II PWR and operated in Fuqing Nuclear Power Plant in China, is simulated using these two libraries. The key parameters during the startup physics tests and power operation in the first three fuel cycles of the CNP-1000 reactor have been simulated and compared with corresponding measurement values. The numerical results show that ENDF/B-VIII.0 performs better in several parameters of the startup physics tests than ENDF/B-VII.0; the cross-section data in CENDL-3.2 is competent in the engineering application of PWR. (C) 2021 Elsevier Ltd. All rights reserved.
Keyword ：
CENDL-3.2 CNP-1000 ENDF/B-VIII.0 LOCUST/SPARK NECP-Atlas
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| GB/T 7714 | Zu, Tiejun , Huang, Yihan , Teng, Qichen et al. Application of CENDL-3.2 and ENDF/B-VIII.0 on the reactor physics simulation of PWR [J]. | ANNALS OF NUCLEAR ENERGY , 2021 , 158 . |
| MLA | Zu, Tiejun et al. "Application of CENDL-3.2 and ENDF/B-VIII.0 on the reactor physics simulation of PWR" . | ANNALS OF NUCLEAR ENERGY 158 (2021) . |
| APA | Zu, Tiejun , Huang, Yihan , Teng, Qichen , Han, Fenglin , Huang, Xing , Wan, Chenghui et al. Application of CENDL-3.2 and ENDF/B-VIII.0 on the reactor physics simulation of PWR . | ANNALS OF NUCLEAR ENERGY , 2021 , 158 . |
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