Study on Activation-Induced Dose Calculation of Reactor Structural Materials

Yuan Xudong; Ma Huiqiang; Chen Zhenping; Guo Shuwei; Xie Jinsen; Yang Chao; Yu Tao

doi:10.13832/j.jnpe.2021.05.0103

Volume 42 Issue 5

Sep. 2021

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Article Navigation > Nuclear Power Engineering > 2021 > 42(5): 103-109

Yuan Xudong, Ma Huiqiang, Chen Zhenping, Guo Shuwei, Xie Jinsen, Yang Chao, Yu Tao. Study on Activation-Induced Dose Calculation of Reactor Structural Materials[J]. Nuclear Power Engineering, 2021, 42(5): 103-109. doi: 10.13832/j.jnpe.2021.05.0103

Citation:

Yuan Xudong, Ma Huiqiang, Chen Zhenping, Guo Shuwei, Xie Jinsen, Yang Chao, Yu Tao. Study on Activation-Induced Dose Calculation of Reactor Structural Materials[J]. Nuclear Power Engineering, 2021, 42(5): 103-109. doi: 10.13832/j.jnpe.2021.05.0103

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Study on Activation-Induced Dose Calculation of Reactor Structural Materials

doi: 10.13832/j.jnpe.2021.05.0103

Yuan Xudong^{1, 2
,},
Ma Huiqiang^{1, 2},
Chen Zhenping^{1, 2
,
,},
Guo Shuwei^{1, 2},
Xie Jinsen^{1, 2},
Yang Chao^{1, 2},
Yu Tao^{1, 2}

1.
School of Nuclear Science and Technology, University of South China, Hengyang, Hunan, 421001, China
2.
Hunan Engineering & Technology Research Center for Virtual Nuclear Reactor, Hengyang, Hunan, 421001, China

Received Date: 2020-08-11
Rev Recd Date: 2021-05-06
Publish Date: 2021-09-30

Abstract

Abstract

Numerous radioactive nuclides are produced from the neutron activation reaction of reactor structural materials under the core neutron irradiation. The decay photon from activated reactor structural materials is an important source of occupational exposure dose for workers during reactor shutdown & maintenance, refueling or decommissioning. This paper studies the cell activation calculation method for reactor structural materials based on Rigorous two step (R2S) method, and develops a reactor structural material activation dose calculation program, MOCA, based on the Monte Carlo particle transport code MCNP and point activation calculation code ORIGEN. Then, this study develops the functional interface and data interface codes to realize the automatic coupling of the transport code and activation calculation code thus to realize fully automated “neutron transport-activation analysis - dose calculation” coupled analysis. Finally, the M5 fuel cladding activation calculation model, stainless steel activation calculation model and NUREG/CR-6115 PWR model are used to perform benchmark verification of the MOCA, which proves that the MOCA is adequate and reliable.
- Structural material,
- R2S method,
- Activation calculation,
- Dose calculation

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References(10)

References

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