Assessment of reactor vessel integrity (ARVI)

B.R. Sehgal (Corresponding Author), A. Karbojian, A. Giri, O. Kymäläinen, J.M. Bonnet, Kari Ikonen, R. Sairanen, S. Bhandari, M. Buerger, J. Dienstbier, Z. Techy, T. Theofanous

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    37 Citations (Scopus)


    The assessment of reactor vessel integrity (ARVI) project involved a total of nine organizations from Europe and USA. The work consisted of experiments and analysis development. The modeling activities in the area of structural analyses were focused on the support of EC-FOREVER experiments as well as on the exploitation of the data obtained from those experiments for modeling of creep deformation and the validation of the industry structural codes. Work was also performed for extension of melt natural convection analyses to consideration of stratification, and mixing (in the CFD codes). Other modeling activities were for (1) gap cooling CHF and (2) developing simple models for system code. Finally, the methodology and data was applied for the design of IVMR severe accident management scheme for VVER-440/213 plants. The work was broken up into five packages. They were divided into tasks, which were performed by different partners. The major experimental project continued was EC-FOREVER in which data was obtained on in-vessel melt pool coolability. In previous EC-FOREVER experiments data was obtained on melt pool natural convection and lower head creep failure and rupture. Those results obtained were related to the following issues: (1) multiaxial creep laws for different vessel steels, (2) effects of penetrations, and (3) mode and location of lower head failure. The two EC-FOREVER tests reported here are related to (a) the effectiveness of gap cooling and (b) water ingression for in vessel melt coolability. Two other experimental projects were also conducted. One was the COPO experiments, which was concerned with the effects of stratification and metal layer on the thermal loads on the lower head wall during melt pool convection. The second experimental project was conducted at ULPU facility, which provided data and correlations of CHF due to the external cooling of the lower head.

    Original languageEnglish
    Pages (from-to)213 - 232
    Number of pages20
    JournalNuclear Engineering and Design
    Issue number2-4
    Publication statusPublished - 2005
    MoE publication typeA1 Journal article-refereed


    • nuclear power plants
    • nuclear reactor safety
    • nuclear safety
    • validation
    • code validation
    • light water reactors
    • pressurized water reactors


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