Creep rupture assessment for Level-2 PSA of a 2-loop PWR: Accounting for phenomenological uncertainties

Faramarz Yousefpour; Seyed Mohsen Hoseyni; SeyedMojtaba Hoseyni; SeyedAli Hashemi Olia; Kaveh Karimi

doi:10.1007/s41365-017-0269-9

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Creep rupture assessment for Level-2 PSA of a 2-loop PWR: Accounting for phenomenological uncertainties

NUCLEAR ENERGY SCIENCE AND ENGINEERING | Updated：2021-02-23

- Creep rupture assessment for Level-2 PSA of a 2-loop PWR: Accounting for phenomenological uncertainties
- Nuclear Science and Techniques Vol. 28, Issue 8, Article number: 107(2017)
- Affiliations：
  
  1.Nuclear Science and Technology Research Institute, Tehran, Iran
  2.Young Researchers and Elite Club, East Tehran branch, Islamic Azad University, Tehran, Iran
- Author bio：
  
  Corresponding author E-mail address: m.hoseini@srbiau.ac.ir
- Funds：
- DOI：10.1007/s41365-017-0269-9
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Faramarz Yousefpour, Seyed Mohsen Hoseyni, SeyedMojtaba Hoseyni, et al. Creep rupture assessment for Level-2 PSA of a 2-loop PWR: Accounting for phenomenological uncertainties. [J]. Nuclear Science and Techniques 28(8):107(2017)
DOI：

Faramarz Yousefpour, Seyed Mohsen Hoseyni, SeyedMojtaba Hoseyni, et al. Creep rupture assessment for Level-2 PSA of a 2-loop PWR: Accounting for phenomenological uncertainties. [J]. Nuclear Science and Techniques 28(8):107(2017) DOI： 10.1007/s41365-017-0269-9.

摘要

Abstract

The Level-2 Probabilistic Safety Assessment (PSA) of pressurized water reactors studies the possibility of creep rupture for major reactor coolant system components during the course of high pressure severe accident sequences. The present paper covers this technical issue and tries to quantify its associated phenomenological uncertainties for the development of Level-2 PSA. A framework is proposed for the formal quantification of uncertainties in the Level-2 PSA model of a PWR type nuclear power plant using an integrated deterministic and probabilistic safety assessment approach. This is demonstrated for estimation of creep rupture failure probability in station blackout severe accident of a 2-loop PWR, which is the representative case for high pressure sequences. MELCOR 1.8.6 code is employed here as the deterministic tool for the assessment of physical phenomena in the course of accident. In addition, a MATLAB code is developed for quantification of the probabilistic part by treating the uncertainties through separation of aleatory and epistemic sources of uncertainty. Theprobability for steam generator tube creep rupture is estimated at 0.17.

关键词

Keywords

IDPSACreep rupturesevere accidentstation blackoutTISGTRPSA level-2.

references

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Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China

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Nuclear Physics and Biophysics Research Division, Physics Department, Faculty of Mathematics and Natural Science, Institut Teknologi Bandung, Jalan Ganesha 10

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