NECP-MCX: A hybrid Monte-Carlo-Deterministic particle-transport code for the simulation of deep-penetration problems

He, Qingming; Zheng, Qi; Li, Jie; Wu, Hongchun; Shen, Wei; Cao, Liangzhi; Liu, Zhouyu; Xu, Jialong

doi:10.1016/j.anucene.2020.107978

Cited by 56 publications

(7 citation statements)

References 18 publications

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“…NECP-MCX is developed based on a hybrid Monte Carlo deterministic method, where the deterministic method is utilized to generate consistent mesh-based weightwindow and source-biasing parameters for the Monte Carlo method to reduce variance. The code has been verified against with various benchmarks (He et al, 2021;Li et al, 2021).…”

Section: Results Of Critical Benchmarksmentioning

confidence: 99%

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Treatments of Thermal Neutron Scattering Data and Their Effect on Neutronics Calculations

Tang

Huang

et al. 2021

Front. Energy Res.

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In the conventional method to generate thermal scattering cross section of moderator materials, only one of the coherent elastic scattering and incoherent elastic scattering is considered in neutronics calculations. For the inelastic scattering, fixed incident energy grid is used in the nuclear data processing codes. The multipoint linearization method is used to refine the incident energy grid for inelastic scattering. We select ZrHx (zirconium hydride) as an example to analyze the effects of the above described treatments on the reactivity of several critical benchmarks. The numerical results show that the incident energy grid has an obvious effect on the effective multiplication factor (keff) of the analyzed reactors; simultaneously considering the coherent and incoherent elastic scattering also affects keff by tens of pcm.

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Section: Results Of Critical Benchmarksmentioning

confidence: 99%

“…In the present work, we will investigate the effects of the aforementioned treatments for thermal neutron scattering data on the neutronics calculations. The analysis is carried out with the Monte Caro code NECP-MCX (He et al, 2021).…”

Section: Introductionmentioning

confidence: 99%

Treatments of Thermal Neutron Scattering Data and Their Effect on Neutronics Calculations

Tang

Huang

et al. 2021

Front. Energy Res.

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“…III lists the building and running environment of OpenMC, in order to redo or evaluate the results. For industrial operations, it suggests to use programs such as JMCT [13], RMC [14], MCX [15], cosRMC [16] and etc. C. On-the-fly Doppler Broadening…”

Section: B Reference Codementioning

confidence: 99%

Monte Carlo neutron transport using low power mobile GPU devices

Liu

2022

Preprint

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Abstract-The using of GPU for Monte Carlo particle transport is lacking of fair comparisons. This work performs simulations on both CPU and GPU in the same package under the same manufacturing process of low power mobile devices. The experiment with simple pincell benchmark problems with fresh fuel gives consistent results between CPU and GPU. In the meanwhile, it finds that the Apple M1 GPU is as twice capable as M1 CPU, while entitled with a 5 times advantage in power consumption. The particle sorting algorithm optimized for GPU improves computing efficiency by 28%, while prominently reducing GPU power consumption. Such advantage of sorting algorithm is expected to be greater for depleted fuel problems than fresh fuel problem. The kernel reconstruction Doppler broadening algorithm designed for continuously varying materials is demonstrated to produce consistent Doppler coefficients with the reference code and the algorithm can be efficiently implemented on GPU. Compared with the reference code with double precision floating point numbers, the testing codes with single precision floating point numbers could underestimate the K-effective values by about 500 pcm, and the Doppler coefficients of the fuel are well reproduced though. The conclusion may strengthen the argument that it is helpful for high performance computer to adopt GPU in order to reduce gross power consumption. This work is an English translated version of an article submitted to the CORPHY 2022 conference(http://corphy2022.org.cn). Original text is in Chinese, and there may be minor difference in the contents. This work is also an extension to the published article. https://www.sciencedirect.com/science/article/pii/S0306454922001852

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“…Neutronic analysis are simulated by the NECP‐MCX, 22 which is a hybrid Monte‐Carlo‐Deterministic particle‐transport code developed by Nuclear Engineering Computational Physics (NECP) laboratory of Xi'an Jiaotong University. The neutronic analysis is based on the geometry shown in Figure 1 and the reactor core design parameters listed in Table 2.…”

Section: Nuster‐100 Core Performance Analysesmentioning

confidence: 99%

Preliminary conceptual design and analysis of a 100 kW _e level Nuclear Silent Thermal‐Electrical Reactor ( NUSTER ‐100)

Huang

Wang

Tian

et al. 2022

Intl J of Energy Research

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Summary To meet the requirements of the unmanned underwater vehicles (UUVs) for the power source, a conceptual design of a 100‐kWe level Nuclear Silent Thermal‐Electrical Reactor (NUSTER‐100) is proposed. The NUSTER‐100 is a 1‐MWt reactor that has been designed to couple with a 100‐kWe thermoelectric generator system. The high temperature sodium heat pipes are adopted to cool the reactor core. The cascaded wide temperature range thermoelectric generation (TEG) system is used to achieve thermoelectric conversion, which ensures high power generation efficiency and low operating noise. The neutronic and thermal‐mechanical analysis results are provided for the design. The core power distribution and energy spectrum are obtained. The neutronic analysis results are imported into ANSYS as input for thermal‐mechanical analysis. In this paper, the fuel‐HPs channel and the 1/8 symmetrical core are selected as the analysis objects, and the temperature distribution and statics parameters under the normal operation condition are obtained. The analysis results show that under normal operation conditions, the core parameters meet the design requirements. These design parameters are all lower than the safety limit. This work provides a certain reference for the design and application of the heat pipe reactor power system.

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NECP-MCX: A hybrid Monte-Carlo-Deterministic particle-transport code for the simulation of deep-penetration problems

Cited by 56 publications

References 18 publications

Treatments of Thermal Neutron Scattering Data and Their Effect on Neutronics Calculations

Treatments of Thermal Neutron Scattering Data and Their Effect on Neutronics Calculations

Monte Carlo neutron transport using low power mobile GPU devices

Preliminary conceptual design and analysis of a 100 kW _e level Nuclear Silent Thermal‐Electrical Reactor ( NUSTER ‐100)

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NECP-MCX: A hybrid Monte-Carlo-Deterministic particle-transport code for the simulation of deep-penetration problems

Cited by 56 publications

References 18 publications

Treatments of Thermal Neutron Scattering Data and Their Effect on Neutronics Calculations

Treatments of Thermal Neutron Scattering Data and Their Effect on Neutronics Calculations

Monte Carlo neutron transport using low power mobile GPU devices

Preliminary conceptual design and analysis of a 100 kW e level Nuclear Silent Thermal‐Electrical Reactor ( NUSTER ‐100)

Contact Info

Product

Resources

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Preliminary conceptual design and analysis of a 100 kW _e level Nuclear Silent Thermal‐Electrical Reactor ( NUSTER ‐100)