A new high-fidelity neutronics code NECP-X

Chen, Jun; Liu, Zhouyu; Zhao, Chen; He, Qingming; Zu, Tiejun; Cao, Liangzhi; Wu, Hongchun

doi:10.1016/j.anucene.2018.02.049

Cited by 125 publications

(9 citation statements)

References 20 publications

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“…In this section, the nuclear-data uncertainties are generally characterized by the multigroup cross-section covariance libraries, which were generated by NJOY [8] based on ENDF/B-VII.1. Moreover, our home-developed high-fidelity PWR code named as NECP-X [9] was used for the transient modeling and simulation for the C5G7 benchmark. As both the multigroup microscopic cross-section library and continue-energy cross-section library were required in NECP-X for the modeling and simulation, hence the continue-energy cross-section perturbation model EPJ Web of Conferences 247, 15007 (2021) PHYSOR2020 https://doi.org/10.1051/epjconf/202124715007 has been established in UNICORN to generate the cross-section samples based on the corresponding covariance libraries.…”

Section: Numerical Results and Analysismentioning

confidence: 99%

Uncertainty Analysis to C5g7-Td Benchmark Based on the Cost Method

et al. 2021

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A method of Covariance-Oriented Sample Transformation (COST) has been proposed in our previous work to provide the converged uncertainty analysis results with a minimal sample size. The transient calculation of nuclear reactor is a key part of the reactor-physics simulation, so the accuracy and confidence of the neutron kinetics results have attracted much attention. In this paper, the Uncertainty Quantification (UQ) function of the high fidelity neutronics code NECP-X has been developed based on our home-developed uncertainty analysis code UNICORN, building a platform for the UQ of the transient calculation. Furthermore, the well-known space-time heterogeneous neutron kinetics benchmark C5G7 and its uncertainty propagation from the nuclear data to the interested key parameters of the core have been investigated. To address the problem of “the curse of dimensionality” caused by the large number of input parameters, the COST method has been applied to generate multivariate normal-distribution samples in uncertainty analysis. As a result, the law of the assembly/pin normalized power and their uncertainty with respect to time after introducing an instantaneous perturbation has been obtained. From the numerical results, it can be observed that the maximum relative uncertainties for the assembly normalized power can up to be about 1.65% and the value for the pin-wise power distributions can be about 2.71%.

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Section: Numerical Results and Analysismentioning

confidence: 99%

Uncertainty Analysis to C5g7-Td Benchmark Based on the Cost Method

et al. 2021

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“…Assembly module ray tracing method (Hong and Cho, 1998) is applied in ray tracing to save the memory cost. Up to now, rectangle and hexagon modular ray tracing has been developed in SHARK with the established method (Cho et al, 2008;Chen et al, 2018). In the future, the long characteristic method (Suslov, 2001) will be added for complicated geometry cases without regular assembly structure.…”

Section: Ray Tracingmentioning

confidence: 99%

“…The research of the direct transport method has become a hotspot in numerical nuclear reactor physics calculation. Several direct transport codes have been developed for numerical nuclear reactor physics calculation, including MPACT (Kelley and Larsen, 2013), nTRACER (Jung, 2013), DeCART (Hursin, 2010), STREAM (Choi et al, 2019), NECP-X (Chen et al, 2018), and PANX (Zhang et al, 2017a;Zhang et al, 2017b). PWR whole-core multi-physics coupling direct calculation has been conducted, and numerical results show good accuracy.…”

Section: Introductionmentioning

confidence: 99%

A New Numerical Nuclear Reactor Neutronics Code SHARK

et al. 2021

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In order to establish the next-generation reactor physics calculation method based on the numerical nuclear reactor technology and realize high-fidelity modeling and calculation, a new numerical nuclear reactor neutronics code SHARK is developed. The code is based on the direct transport method with construct solid geometry (CSG) method, advanced subgroup resonance method, direct transport MOC method in rectangle and hexagonal geometry, large-scale parallel, and CMFD acceleration method. The C5G7, macro BEAVRS and VERA benchmarks are verified to show the accuracy of the code and method. Numerical results show good accuracy and calculation performance of SHARK, and the direct transport method can be adopted on numerical nuclear reactor calculation.

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“…Then the consistent subgroup fixed-source equations are solved to obtain the self-shielded XSs. These methods are implemented in the self-shielding module of NECP-X [7], which is a high-fidelity deterministic neutronics code, like DeCART, MPACT and nTRACER.…”

Section: Introductionmentioning

confidence: 99%

Methods for Calculation of Self-Shielded Cross Sections of Fully Ceramic Micro-Encapsulated Fuel Double Heterogeneous System

Wen

Liu

et al. 2021

EPJ Web Conf.

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Fully ceramic micro-encapsulated (FCM) fuel is a kind of fuel that employs tri-structural isotropic (TRISO) particles to enhance safety. The FCM fuel assembly is a double heterogeneous system. The conventional self-shielding calculation methods cannot treat the DH effect. In this paper, three methods based on equivalent homogenization of the TRISO particle and the matrix are studied and compared: the hyper-fine energy group cross sections (XSs) homogenization based hyper-fine energy group method (HHM), the hyper-fine energy group XSs homogenization based subgroup method (HSM) and the subgroup XSs homogenization based subgroup method (SSM). These methods are implemented in a high-fidelity neutronics code NECP-X. Numerical results show that these methods are able to treat the double heterogeneity of the FCM fuel. The precision of the HHM and HSM is higher than that of the SSM.

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A new high-fidelity neutronics code NECP-X

Cited by 125 publications

References 20 publications

Uncertainty Analysis to C5g7-Td Benchmark Based on the Cost Method

Uncertainty Analysis to C5g7-Td Benchmark Based on the Cost Method

A New Numerical Nuclear Reactor Neutronics Code SHARK

Methods for Calculation of Self-Shielded Cross Sections of Fully Ceramic Micro-Encapsulated Fuel Double Heterogeneous System

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