MATERIAL CONSTITUTIVE MODELING OF NUCLEAR PRESSURE VESSEL STEEL FOR IVR-ERVC SIMULATION

Eui Kyun Park, Yukio Takahashi, Ji Su Kim, Jun Won Park, Yun Jae Kim

Research output: Chapter in Book/Report/Conference proceedingConference contribution

Abstract

The lower head of the nuclear power plant reactor pressure vessel (RPV) can be subjected to severe thermal and pressure loads during the event of a core meltdown accident. As effective accident management strategy, IVR-ERVC (In-Vessel Retention of molten corium through External Reactor Vessel Cooling) strategy is introduced to reduce the possibility of a reactor containment failure by terminating the severe accident progress inside a reactor. In this case, the mechanical behavior of the reactor vessel lower head is of importance both in severe accident assessment and the assessment of accident mitigation strategies. This paper proposed material constitutive model, which is extended from constitutive model of Takahashi [1]. The proposed model can predict stress under large range of temperature and strain rate, which in turn can be used to predict material deformation under IVR-ERVC strategy.

Original languageEnglish
Title of host publicationProceedings of the 30th International Conference on Nuclear Engineering "Nuclear, Thermal, and Renewables
Subtitle of host publicationUnited to Provide Carbon Neutral Power", ICONE 2023
PublisherAmerican Society of Mechanical Engineers (ASME)
ISBN (Print)9784888982566
Publication statusPublished - 2023
Event30th International Conference on Nuclear Engineering, ICONE 2023 - Kyoto, Japan
Duration: 2023 May 212023 May 26

Publication series

NameInternational Conference on Nuclear Engineering, Proceedings, ICONE
Volume2023-May

Conference

Conference30th International Conference on Nuclear Engineering, ICONE 2023
Country/TerritoryJapan
CityKyoto
Period23/5/2123/5/26

Bibliographical note

Publisher Copyright:
© 2023 by JSME.

Keywords

  • IVR-ERVC
  • Nuclear pressure vessel
  • Temperature and strain rate dependent constitutive model
  • Visco-plastic model

ASJC Scopus subject areas

  • Nuclear Energy and Engineering

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