Study on High-Fidelity Thermal-Neutronic Coupling Method Based on the Unified Geometry Modeling and its Application in Experimental Reactor Core Calculation for SPERT

Zhang Minwan; Liu Zhouyu; Wang Bo; Cao Lu; Zhao Chen; Cao Liangzhi

doi:10.13832/j.jnpe.2021.05.0042

Volume 42 Issue 5

Sep. 2021

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Article Navigation > Nuclear Power Engineering > 2021 > 42(5): 42-50

Zhang Minwan, Liu Zhouyu, Wang Bo, Cao Lu, Zhao Chen, Cao Liangzhi. Study on High-Fidelity Thermal-Neutronic Coupling Method Based on the Unified Geometry Modeling and its Application in Experimental Reactor Core Calculation for SPERT[J]. Nuclear Power Engineering, 2021, 42(5): 42-50. doi: 10.13832/j.jnpe.2021.05.0042

Citation:

Zhang Minwan, Liu Zhouyu, Wang Bo, Cao Lu, Zhao Chen, Cao Liangzhi. Study on High-Fidelity Thermal-Neutronic Coupling Method Based on the Unified Geometry Modeling and its Application in Experimental Reactor Core Calculation for SPERT[J]. Nuclear Power Engineering, 2021, 42(5): 42-50. doi: 10.13832/j.jnpe.2021.05.0042

Citation:

Zhang Minwan, Liu Zhouyu, Wang Bo, Cao Lu, Zhao Chen, Cao Liangzhi. Study on High-Fidelity Thermal-Neutronic Coupling Method Based on the Unified Geometry Modeling and its Application in Experimental Reactor Core Calculation for SPERT[J]. Nuclear Power Engineering, 2021, 42(5): 42-50. doi: 10.13832/j.jnpe.2021.05.0042

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Study on High-Fidelity Thermal-Neutronic Coupling Method Based on the Unified Geometry Modeling and its Application in Experimental Reactor Core Calculation for SPERT

doi: 10.13832/j.jnpe.2021.05.0042

1.
Xi’an Jiaotong University, Xi’an, 710049, China
2.
Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu, 610213, China

Received Date: 2020-08-25
Rev Recd Date: 2020-09-08
Publish Date: 2021-09-30

Abstract

Abstract

To solve the problem that the thermal-neutronic grid mapping relation is complicated and cannot be preset in a centralized manner due to the existence of irregular geometry in performing the thermal-neutronic coupled simulation calculation for various small power reactors and experimental reactors, this paper studies the thermal-neutronic coupling method based on the unified geometric modeling, using the high-fidelity numerical code for reactor neutronics calculation, NECP-X. This study establishes the mapping relation for the thermal-neutronic coupling on the basis of the neutronics model, and enables the direct transient calculation of the experimental reactor core for the special power excursion reactor test (SPERT) via combination with the transient calculation method in NECP-X. Then, this study calculates the steady-state case for the experimental reactors of SPERT, and compares the calculation results with the results gained from the Monte Carlo code. On this basis, this study conducts transient calculation and analysis for these experimental reactors and compares the corresponding results with the experimental results. The final results show that the eigenvalues from the neutronics calculation by the NECP-X and the rod power distribution calculation results are of high accuracy; that the grid mapping method based on the unified geometric modeling allows a easy and fast thermal-neutronic coupled calculation of the PWRs of complex geometry; and that compared with the experimental values, the curve of change in the total power and reactivity gained from transient calculation with time is more accurate and can provide refined distributions of power and temperature.
- Complex geometry,
- Thermal-Neutronic coupling,
- Experimental reactor for SPERT

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References(13)

References

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