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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 ：
Traditional nuclear fuel performance codes usually employ the operator split methods for multiphysics coupling and the quasi-two-dimensional approach (1.5D approach) for geometry modelling. However, the multiphysics behavior of the fuel element is often tightly coupled, and there exist some kinds of non-axisymmetric fuel elements in reactors. In this paper, a new fuel performance code named NECP-CALF has been developed based on the Multiphysics Object-Oriented Simulation Environment (MOOSE). It solves the multiphysics coupled equations using the JFNK method. A new geometrical approach called mixed dimensional approach is implemented in the new code, which allows users to model fuel elements with more flexibility. A UO2-Zr fuel rod is simulated and the results are compared with BISON and CAMPUS to establish the proof-of-concept of this code. Then a small-scale UO2-Zr fuel rod with azimuthally asymmetric cladding temperature is designed and simulated with the mixed dimensional approach to verify the accuracy and demonstrate the performance. Finally, the code is applied to the fuel performance simulation of a supercritical water-cooled fuel rod with flow blockage, and the results show the impact of the flow blockage on the fuel rod performance. © 2021 Elsevier B.V.
Keyword ：
Confined flow Fuels Nuclear fuel elements Oxide minerals Uranium dioxide
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| GB/T 7714 | Liu, Zhouyu , Xu, Xiaobei , Wu, Hongchun et al. Multidimensional multiphysics simulations of the supercritical water-cooled fuel rod behaviors based on a new fuel performance code developed on the MOOSE platform [J]. | Nuclear Engineering and Design , 2021 , 375 . |
| MLA | Liu, Zhouyu et al. "Multidimensional multiphysics simulations of the supercritical water-cooled fuel rod behaviors based on a new fuel performance code developed on the MOOSE platform" . | Nuclear Engineering and Design 375 (2021) . |
| APA | Liu, Zhouyu , Xu, Xiaobei , Wu, Hongchun , Cao, Liangzhi . Multidimensional multiphysics simulations of the supercritical water-cooled fuel rod behaviors based on a new fuel performance code developed on the MOOSE platform . | Nuclear Engineering and Design , 2021 , 375 . |
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Abstract ：
Based on the conventional 'two-step' scheme for the PWR fuel management, an advanced PWR fuel management software Bamboo-C has been developed by the most advanced methodologies in the reactor-physics field. Bamboo-C consists of three main functional codes: LOCUST code for the heterogeneous modeling and simulation and homogenization calculation of 2D assemblies; SPARK code for 3D core steady-state and transient analysis; and LtoS code for assembly homogenization parameter function, which links LOCUST and SPARK. Bamboo-C has all the necessary analysis functions for the fuel management and nuclear design of PWRs, mainly including the start-up physics tests, calculations of the neutron-kinetics parameters, differential and integral worth of rod cluster control assemblies (RCCAs), and power-operation following simulation. Finally, the engineering validations of Bamboo-C have been completed according to the operation data from the reactors CNP300, CNP650 and CNP1000 designed by China independently. The validation results show that the errors between the values of such key parameters of cores as critical boron concentration, temperature coefficient, RCCA worth, and power distributions, calculated by Bamboo-C, and their measured values satisfy the corresponding engineering criterion limits. © 2021, Editorial Board of Journal of Nuclear Power Engineering. All right reserved.
Keyword ：
Bamboo C (programming language) Fuels Pressurized water reactors Transient analysis
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| GB/T 7714 | Wan, Chenghui , Li, Yunzhao , Zheng, Youqi et al. Code Development and Engineering Validation of PWR Fuel Management Software Bamboo-C [J]. | Nuclear Power Engineering , 2021 , 42 (5) : 15-22 . |
| MLA | Wan, Chenghui et al. "Code Development and Engineering Validation of PWR Fuel Management Software Bamboo-C" . | Nuclear Power Engineering 42 . 5 (2021) : 15-22 . |
| APA | Wan, Chenghui , Li, Yunzhao , Zheng, Youqi , Liu, Zhouyu , Zu, Tiejun , Cao, Liangzhi et al. Code Development and Engineering Validation of PWR Fuel Management Software Bamboo-C . | Nuclear Power Engineering , 2021 , 42 (5) , 15-22 . |
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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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Abstract ：
PWR-core fuel management code system NECP-Bamboo has been validated by about one hundred commercial PWR-core cycles including both the Second and Third Generations such as Chinese Nuclear Power CNP300 and CNP600, Chinese Pressurized Reactor CPR1000 and ACPR1000, France Framatome M310 and European Pressurized Reactor (EPR), USA Westinghouse AP1000, Chinese CAP1400 and Hualong-1, etc. Among those, the last four belong to the Third Generation PWRs. Besides, BEAVRS is the first worldwide well-known international benchmark that provides a complete two-cycle description of a commercial PWR-core, including detailed operating data and measurements, while the Unit 1 of Sanmen Nuclear Power Plant (NPP) in mainland China is the worldwide first AP1000 core. Thus, in this paper, measurements of these two PWR-cores were chosen to validate the code NECP-Bamboo, including critical boron concentration (CBC), control rod worth, temperature coefficient, detector response and boron letdown curve. Satisfying agreements have been found between the simulation with NECP-Bamboo and the corresponding measurements. © 2021 Elsevier B.V.
Keyword ：
Bamboo Boron Nuclear energy Nuclear fuels Nuclear power plants Pressurized water reactors
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| GB/T 7714 | Yang, Jiewei , Wu, Hongchun , Guo, Lin et al. Validation of NECP-Bamboo with BEAVRS and AP1000 measurements [J]. | Nuclear Engineering and Design , 2021 , 376 . |
| MLA | Yang, Jiewei et al. "Validation of NECP-Bamboo with BEAVRS and AP1000 measurements" . | Nuclear Engineering and Design 376 (2021) . |
| APA | Yang, Jiewei , Wu, Hongchun , Guo, Lin , Li, Yunzhao , Li, Xuesong , Wang, Sicheng et al. Validation of NECP-Bamboo with BEAVRS and AP1000 measurements . | Nuclear Engineering and Design , 2021 , 376 . |
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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 ：
Fully Ceramic Micro-encapsulated (FCM) fuel is an important candidate for the accident tolerant fuel (ATF). Compared with traditional fuel, the double heterogeneity of FCM fuel makes the effective multi-group cross section calculation more challenging. In this paper, an improved disadvantage factor method is proposed to deal with the self-shielding effect of FCM fuel in the resonance energy range and non-resonance energy range, to achieve the equivalent homogenization of the FCM fuel. A new equivalent particle–matrix model was constructed by using the particle Dancoff factor to overcome the problem that the traditional volume-weight equivalent model could not consider the macro heterogeneity between fuel rods. Based on the new one-dimensional equivalent sphere model, the ultrafine group slowing down equation is solved to obtain the ultrafine group disadvantage factor in the resonance energy region. In the non-resonance energy region, the multi-group disadvantage factors of the fast group and thermal group are obtained by using the eigenvalue calculation in the new equivalent particle–matrix model. The proposed method has been implemented in the high fidelity neutronics program NECP-X and tested with a set of cases. The results show its good agreement with the Monte Carlo reference for both the reactivity and self-shielding cross sections. © 2021 Elsevier Ltd
Keyword ：
Eigenvalues and eigenfunctions Fuels Homogenization method Resonance Shielding Software testing
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| GB/T 7714 | Yi, Siyu , Liu, Zhouyu , He, Qingming et al. Effective multi-group cross section calculation method for FCM fuel based on improved disadvantage factor method [J]. | Annals of Nuclear Energy , 2021 , 166 . |
| MLA | Yi, Siyu et al. "Effective multi-group cross section calculation method for FCM fuel based on improved disadvantage factor method" . | Annals of Nuclear Energy 166 (2021) . |
| APA | Yi, Siyu , Liu, Zhouyu , He, Qingming , Zu, Tiejun , Cao, Liangzhi , Wu, Hongchun et al. Effective multi-group cross section calculation method for FCM fuel based on improved disadvantage factor method . | Annals of Nuclear Energy , 2021 , 166 . |
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