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学者姓名:吴宏春

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< Page ,Total 55 >
Uncertainty analysis of infinite multiplication factor and nuclide number density based on the UAM-PWR benchmark with respect to cross sections, fission yields and decay half-life EI
期刊论文 | 2022 , 165 | Annals of Nuclear Energy
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Abstract :

Uncertainty analysis is performed to qualify the uncertainty of the infinite multiplication factor (kinf) and nuclide number density induced by cross sections, fission yields and decay half-life in ENDF/B-VIII.0. The SUNDEW code, initially developed for burnup sensitivity and uncertainty analysis with respect to cross sections based on the generalized perturbation theory, is extended to consider fission yields and decay half-life. The Bayesian/general least-squares method is adopted to generate the covariance matrix of fission yields. TMI-1 PWR pin cell given in the UAM project is calculated by SUNDEW. In addition to kinf, the uncertainty of the number densities of various nuclides important for different aspects are calculated and analyzed, including nuclides involved in burnup credit, nuclides used for burnup measurement, and 135Xe. Main conclusions are that the cross section covariances in ENDF/B-VIII.0 can induce larger uncertainty to kinf than ENDF/B-VII.1; the fission yields and decay half-life have obvious contributions for the uncertainty of nuclide number density in some cases. © 2021 Elsevier Ltd

Keyword :

Pressurized water reactors Uncertainty analysis Covariance matrix Perturbation techniques Least squares approximations Decay (organic) Factor analysis Isotopes

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GB/T 7714 Zu, Tiejun , Lu, Zerun , Han, Fenglin et al. Uncertainty analysis of infinite multiplication factor and nuclide number density based on the UAM-PWR benchmark with respect to cross sections, fission yields and decay half-life [J]. | Annals of Nuclear Energy , 2022 , 165 .
MLA Zu, Tiejun et al. "Uncertainty analysis of infinite multiplication factor and nuclide number density based on the UAM-PWR benchmark with respect to cross sections, fission yields and decay half-life" . | Annals of Nuclear Energy 165 (2022) .
APA Zu, Tiejun , Lu, Zerun , Han, Fenglin , Shu, Nengchuan , Liu, Zhouyu , Cao, Liangzhi et al. Uncertainty analysis of infinite multiplication factor and nuclide number density based on the UAM-PWR benchmark with respect to cross sections, fission yields and decay half-life . | Annals of Nuclear Energy , 2022 , 165 .
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Effective multi-group cross section calculation method for FCM fuel based on improved disadvantage factor method EI
期刊论文 | 2022 , 166 | Annals of Nuclear Energy
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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 :

Software testing Eigenvalues and eigenfunctions Homogenization method Fuels Shielding Resonance

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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 , 2022 , 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 (2022) .
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 , 2022 , 166 .
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An IFDF accelerated parallel nodal SN method for XYZ geometry in SARAX code system EI
期刊论文 | 2022 , 166 | Annals of Nuclear Energy
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Abstract :

There are some fast reactors designed with square assemblies requiring whole-core multigroup neutron transport solutions. A simplified method to calculate the high-order transverse leakage terms was proposed based on the NEFD nodal SN scheme in XYZ geometry. The IFDF acceleration method with adaptive diffusion coefficients and parallel algorithm with spatial domain decomposition and MPI communications were employed to solve the discretization system efficiently. A small fast reactor and the ZPPR-10B large fast reactor were calculated. Encouraging accuracy improvements were obtained for both the eigenvalue and the distribution quantities with only about 10% additional calculation cost for the simplified high-order transverse leakage terms. The 4-group small fast reactor was solved in about 2 s with 16 CPU cores and the 33-group ZPPR-10B large fast reactor was solved in about 19 s with 196 CPU cores. Speedup beyond 60 was obtained compared with the traditional serial fission-source iteration on a 16-CPU workstation. © 2021 Elsevier Ltd

Keyword :

Fast reactors Eigenvalues and eigenfunctions Domain decomposition methods Iterative methods

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GB/T 7714 Xu, Zhitao , Zheng, Youqi , Wu, Hongchun . An IFDF accelerated parallel nodal SN method for XYZ geometry in SARAX code system [J]. | Annals of Nuclear Energy , 2022 , 166 .
MLA Xu, Zhitao et al. "An IFDF accelerated parallel nodal SN method for XYZ geometry in SARAX code system" . | Annals of Nuclear Energy 166 (2022) .
APA Xu, Zhitao , Zheng, Youqi , Wu, Hongchun . An IFDF accelerated parallel nodal SN method for XYZ geometry in SARAX code system . | Annals of Nuclear Energy , 2022 , 166 .
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Neutronics/thermal-hydraulics/fuel-performance coupling for light water reactors and its application to accident tolerant fuel EI
期刊论文 | 2022 , 166 | Annals of Nuclear Energy
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To accurately assess the in-pile performance of the accident tolerant fuel (ATF) over the whole life for the light water reactors (LWR), a full coupling of neutronics/thermal-hydraulics/fuel-performance is needed. In this paper, a multiphysics coupling code system named NECP-CLAMPERL is developed based on the high-fidelity neutronics code NECP-X, the sub-channel thermal-hydraulics code CTF, the finite-element fuel performance code NECP-CALF and the multiphysics object-oriented simulation environment (MOOSE). The code system is then applied to evaluate the performance of U3Si2 fuel, one of the most promising ATFs, by simulating a typical fuel assembly in pressurized water-cooled reactors (PWR). The quantities of interest including the reactivity, power distribution, fuel temperature, clad stress, fission gas release, plenum pressure, fuel radial displacement, gap width and PCMI are presented and analyzed in comparison with the traditional UO2 fuel. Future works and other potential applications are also identified. © 2021 Elsevier Ltd

Keyword :

Uranium dioxide Codes (symbols) Light water reactors Pressurized water reactors Fission products Piles Fuels Accidents Silicon compounds

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GB/T 7714 Xu, Xiaobei , Liu, Zhouyu , Wu, Hongchun et al. Neutronics/thermal-hydraulics/fuel-performance coupling for light water reactors and its application to accident tolerant fuel [J]. | Annals of Nuclear Energy , 2022 , 166 .
MLA Xu, Xiaobei et al. "Neutronics/thermal-hydraulics/fuel-performance coupling for light water reactors and its application to accident tolerant fuel" . | Annals of Nuclear Energy 166 (2022) .
APA Xu, Xiaobei , Liu, Zhouyu , Wu, Hongchun , Cao, Liangzhi . Neutronics/thermal-hydraulics/fuel-performance coupling for light water reactors and its application to accident tolerant fuel . | Annals of Nuclear Energy , 2022 , 166 .
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Moment Matching: A New Optimization-Based Sampling Scheme for Uncertainty Quantification of Reactor-Physics Analysis SCIE
期刊论文 | 2021 | NUCLEAR SCIENCE AND ENGINEERING
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Because of the complexity of the nuclear reactor system, traditional statistical sampling methods, such as random sampling and Latin hypercube sampling, often lead to unstable uncertainty quantification results of the reactor physics analysis. In order to make the analysis results robust, traditional sampling methods require a large number of samples, which brings a huge computation cost. For this reason, this paper proposes a new sampling scheme based on the moment matching method to generate efficient samples for the uncertainty quantification of reactor physics calculations. A linear programming model is established to minimize the deviations of the first- and second-order moments. The generated samples can better reflect the statistical characteristics of the real distribution than classical sampling methods. A series of numerical experiments is carried out to demonstrate the superiority of the proposed moment matching sampling method, which can quickly provide more reliable uncertainty quantification results with a small sample size.

Keyword :

small sample size linear programming moment matching Reactor physics analysis uncertainty quantification

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GB/T 7714 Ji, Bingbing , Chen, Zhiping , Liu, Jia et al. Moment Matching: A New Optimization-Based Sampling Scheme for Uncertainty Quantification of Reactor-Physics Analysis [J]. | NUCLEAR SCIENCE AND ENGINEERING , 2021 .
MLA Ji, Bingbing et al. "Moment Matching: A New Optimization-Based Sampling Scheme for Uncertainty Quantification of Reactor-Physics Analysis" . | NUCLEAR SCIENCE AND ENGINEERING (2021) .
APA Ji, Bingbing , Chen, Zhiping , Liu, Jia , Cao, Liangzhi , Sui, Zhuojie , Wu, Hongchun . Moment Matching: A New Optimization-Based Sampling Scheme for Uncertainty Quantification of Reactor-Physics Analysis . | NUCLEAR SCIENCE AND ENGINEERING , 2021 .
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Resonance calculation based on the deep learning method for treating the non-uniform fuel temperature distribution in PWRs EI SCIE
期刊论文 | 2021 , 160 | ANNALS OF NUCLEAR ENERGY
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In the previous work, a global-local resonance method was developed for the PWR self-shielding calculations, but there are still some difficulties when dealing with the non-uniform temperature profile of the fuel rod. This work proposes a deep learning based resonance calculation scheme for treating the non-uniform fuel temperature distribution. In this scheme, the deep learning model is investigated to replace the ultra-fine group method for the local calculations of the global-local resonance method, because the ultra-fine group method is too slow for performing the whole core resonance calculations when considering the temperature profile. The generation of training data sets, deep learning models and some numerical results are introduced. The given tests demonstrate that the new method could give accurate self-shielding cross sections and significantly improve efficiency. (C) 2021 Elsevier Ltd. All rights reserved.

Keyword :

Non-uniform fuel temperature distribution Resonance calculation Deep learning NECP-X

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GB/T 7714 Cao, Lu , Liu, Zhouyu , Wen, Xingjian et al. Resonance calculation based on the deep learning method for treating the non-uniform fuel temperature distribution in PWRs [J]. | ANNALS OF NUCLEAR ENERGY , 2021 , 160 .
MLA Cao, Lu et al. "Resonance calculation based on the deep learning method for treating the non-uniform fuel temperature distribution in PWRs" . | ANNALS OF NUCLEAR ENERGY 160 (2021) .
APA Cao, Lu , Liu, Zhouyu , Wen, Xingjian , Cao, Liangzhi , Wu, Hongchun , Qin, Shuai . Resonance calculation based on the deep learning method for treating the non-uniform fuel temperature distribution in PWRs . | ANNALS OF NUCLEAR ENERGY , 2021 , 160 .
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Code Development and Engineering Validation of PWR Fuel Management Software Bamboo-C EI
期刊论文 | 2021 , 42 (5) , 15-22 | Nuclear Power Engineering
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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 :

Transient analysis C (programming language) Bamboo Pressurized water reactors Fuels

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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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Method research and engineering application of the B10-abundance correction for PWR EI SCIE
期刊论文 | 2021 , 378 | NUCLEAR ENGINEERING AND DESIGN
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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 Boronation effect Bamboo-C 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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Simulations of the source-range detector response for the fuel-loading process of the AP1000 cores EI SCIE
期刊论文 | 2021 , 372 | Nuclear Engineering and Design
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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 Pressurized water reactors Nuclear fuels Numerical methods Nuclear power plants

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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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Performance of CENDL-3.2 evaluated nuclear data library for the shielding benchmarks EI SCIE
期刊论文 | 2021 , 136 | PROGRESS IN NUCLEAR ENERGY
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The latest CENDL-3.2 evaluated nuclear data library was released in June 2020. To verify the capability of CENDL-3.2 in shielding benchmarks, a broad-group shielding library is produced using NECP-Atlas based on CENDL-3.2, following the generation method of BUGLE library. Other broad-group shielding libraries based on evaluated nuclear data libraries including ENDF/B-VII.0, ENDF/B-VII.1 and ENDF/B-VIII.0 are also generated for comparison. The broad-group libraries are firstly validated against several benchmarks, and then used for the calculation of shielding benchmarks including Iron-88, PCA-Replica and HBR-2. The numerical results indicate that the accuracy of the results obtained with CENDL-3.2 in these shielding benchmarks is acceptable. For the Sulphur dosimeter in Iron-88, CENDL-3.2 gives better performance, and through detailed analysis, it is found that the difference is mainly caused by the reevaluation of the inelastic and elastic scattering cross sections of 56Fe.

Keyword :

NECP-Atlas HBR-2 2 NECP-Hydra Iron-88 CENDL-3 PCA-Replica

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GB/T 7714 Shu, Wenyu , Zu, Tiejun , Cao, Liangzhi et al. Performance of CENDL-3.2 evaluated nuclear data library for the shielding benchmarks [J]. | PROGRESS IN NUCLEAR ENERGY , 2021 , 136 .
MLA Shu, Wenyu et al. "Performance of CENDL-3.2 evaluated nuclear data library for the shielding benchmarks" . | PROGRESS IN NUCLEAR ENERGY 136 (2021) .
APA Shu, Wenyu , Zu, Tiejun , Cao, Liangzhi , Wu, Hongchun . Performance of CENDL-3.2 evaluated nuclear data library for the shielding benchmarks . | PROGRESS IN NUCLEAR ENERGY , 2021 , 136 .
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