Assessment of reactor vessel integrity (ARVI)

B.R. Sehgal (Corresponding Author), A. Theerthan, A. Giri, A. Karbojian, H.G. Willschütz, O. Kymäläinen, S. Vandroux, J.M. Bonnet, J.M. Seiler, Kari Ikonen, Risto Sairanen, S. Bhandari, M. Bürger, M. Buck, W. Widmann, J. Dienstbier, Z. Techy, P. Kostka, R. Taubner, T. Theofanous & 1 others T.N. Dinh

Research output: Contribution to journalArticleScientificpeer-review

52 Citations (Scopus)

Abstract

The cost-shared project ARVI (assessment of reactor vessel integrity) involves a total of nine organisations from Europe and USA. The objective of the ARVI Project is to resolve the safety issues that remain unresolved for the melt vessel interaction phase of the in-vessel progression of a severe accident. The work consists of experiments and analysis development. Four tests were performed in the EC-FOREVER Programme, in which failure was achieved in-vessels employing the French pressure vessel steel. The tests were analysed with the commercial code ANSYS-Multiphysics, and the codes SYSTUS+ and PASULA, and quite good agreement was achieved for the failure location. Natural convection experiments in stratified pools have been performed in the SIMECO and the COPO facilities, which showed that much greater heat is transferred downwards for immiscible layers or before layers mix. A model for gap cooling and a set of simplified models for the system codes have been developed. MVITA code calculations have been performed for the Czech and Hungarian VVERs, towards evaluation of the in-vessel melt retention accident management scheme.

Tests have been performed at the ULPU facility with organised flow for vessel external cooling. Considerable enhancement of the critical heat flux (CHF) was obtained. The ARVI Project has reached the halfway stage. This paper presents the results obtained thus far from the project.
Original languageEnglish
Pages (from-to)23-53
Number of pages31
JournalNuclear Engineering and Design
Volume221
Issue number1-3
DOIs
Publication statusPublished - 2003
MoE publication typeA1 Journal article-refereed
EventMid-Term Symposium on Shared-Cost and Concerted Actions -
Duration: 1 Jan 2003 → …

Fingerprint

integrity
vessels
Accidents
vessel
reactors
Cooling
Steel structures
Natural convection
Heat flux
Experiments
accidents
Costs
accident
cooling
melt
project assessment
pressure vessels
reactor
free convection
progressions

Keywords

  • critical heat flux
  • nuclear reactor accidents
  • nuclear reactor safety
  • nuclear reactors

Cite this

Sehgal, B. R., Theerthan, A., Giri, A., Karbojian, A., Willschütz, H. G., Kymäläinen, O., ... Dinh, T. N. (2003). Assessment of reactor vessel integrity (ARVI). Nuclear Engineering and Design, 221(1-3), 23-53. https://doi.org/10.1016/S0029-5493(02)00343-6
Sehgal, B.R. ; Theerthan, A. ; Giri, A. ; Karbojian, A. ; Willschütz, H.G. ; Kymäläinen, O. ; Vandroux, S. ; Bonnet, J.M. ; Seiler, J.M. ; Ikonen, Kari ; Sairanen, Risto ; Bhandari, S. ; Bürger, M. ; Buck, M. ; Widmann, W. ; Dienstbier, J. ; Techy, Z. ; Kostka, P. ; Taubner, R. ; Theofanous, T. ; Dinh, T.N. / Assessment of reactor vessel integrity (ARVI). In: Nuclear Engineering and Design. 2003 ; Vol. 221, No. 1-3. pp. 23-53.
@article{1feed1a980c346ecb4ccd73e0f2d1a83,
title = "Assessment of reactor vessel integrity (ARVI)",
abstract = "The cost-shared project ARVI (assessment of reactor vessel integrity) involves a total of nine organisations from Europe and USA. The objective of the ARVI Project is to resolve the safety issues that remain unresolved for the melt vessel interaction phase of the in-vessel progression of a severe accident. The work consists of experiments and analysis development. Four tests were performed in the EC-FOREVER Programme, in which failure was achieved in-vessels employing the French pressure vessel steel. The tests were analysed with the commercial code ANSYS-Multiphysics, and the codes SYSTUS+ and PASULA, and quite good agreement was achieved for the failure location. Natural convection experiments in stratified pools have been performed in the SIMECO and the COPO facilities, which showed that much greater heat is transferred downwards for immiscible layers or before layers mix. A model for gap cooling and a set of simplified models for the system codes have been developed. MVITA code calculations have been performed for the Czech and Hungarian VVERs, towards evaluation of the in-vessel melt retention accident management scheme.Tests have been performed at the ULPU facility with organised flow for vessel external cooling. Considerable enhancement of the critical heat flux (CHF) was obtained. The ARVI Project has reached the halfway stage. This paper presents the results obtained thus far from the project.",
keywords = "critical heat flux, nuclear reactor accidents, nuclear reactor safety, nuclear reactors",
author = "B.R. Sehgal and A. Theerthan and A. Giri and A. Karbojian and H.G. Willsch{\"u}tz and O. Kym{\"a}l{\"a}inen and S. Vandroux and J.M. Bonnet and J.M. Seiler and Kari Ikonen and Risto Sairanen and S. Bhandari and M. B{\"u}rger and M. Buck and W. Widmann and J. Dienstbier and Z. Techy and P. Kostka and R. Taubner and T. Theofanous and T.N. Dinh",
year = "2003",
doi = "10.1016/S0029-5493(02)00343-6",
language = "English",
volume = "221",
pages = "23--53",
journal = "Nuclear Engineering and Design",
issn = "0029-5493",
publisher = "Elsevier",
number = "1-3",

}

Sehgal, BR, Theerthan, A, Giri, A, Karbojian, A, Willschütz, HG, Kymäläinen, O, Vandroux, S, Bonnet, JM, Seiler, JM, Ikonen, K, Sairanen, R, Bhandari, S, Bürger, M, Buck, M, Widmann, W, Dienstbier, J, Techy, Z, Kostka, P, Taubner, R, Theofanous, T & Dinh, TN 2003, 'Assessment of reactor vessel integrity (ARVI)', Nuclear Engineering and Design, vol. 221, no. 1-3, pp. 23-53. https://doi.org/10.1016/S0029-5493(02)00343-6

Assessment of reactor vessel integrity (ARVI). / Sehgal, B.R. (Corresponding Author); Theerthan, A.; Giri, A.; Karbojian, A.; Willschütz, H.G.; Kymäläinen, O.; Vandroux, S.; Bonnet, J.M.; Seiler, J.M.; Ikonen, Kari; Sairanen, Risto; Bhandari, S.; Bürger, M.; Buck, M.; Widmann, W.; Dienstbier, J.; Techy, Z.; Kostka, P.; Taubner, R.; Theofanous, T.; Dinh, T.N.

In: Nuclear Engineering and Design, Vol. 221, No. 1-3, 2003, p. 23-53.

Research output: Contribution to journalArticleScientificpeer-review

TY - JOUR

T1 - Assessment of reactor vessel integrity (ARVI)

AU - Sehgal, B.R.

AU - Theerthan, A.

AU - Giri, A.

AU - Karbojian, A.

AU - Willschütz, H.G.

AU - Kymäläinen, O.

AU - Vandroux, S.

AU - Bonnet, J.M.

AU - Seiler, J.M.

AU - Ikonen, Kari

AU - Sairanen, Risto

AU - Bhandari, S.

AU - Bürger, M.

AU - Buck, M.

AU - Widmann, W.

AU - Dienstbier, J.

AU - Techy, Z.

AU - Kostka, P.

AU - Taubner, R.

AU - Theofanous, T.

AU - Dinh, T.N.

PY - 2003

Y1 - 2003

N2 - The cost-shared project ARVI (assessment of reactor vessel integrity) involves a total of nine organisations from Europe and USA. The objective of the ARVI Project is to resolve the safety issues that remain unresolved for the melt vessel interaction phase of the in-vessel progression of a severe accident. The work consists of experiments and analysis development. Four tests were performed in the EC-FOREVER Programme, in which failure was achieved in-vessels employing the French pressure vessel steel. The tests were analysed with the commercial code ANSYS-Multiphysics, and the codes SYSTUS+ and PASULA, and quite good agreement was achieved for the failure location. Natural convection experiments in stratified pools have been performed in the SIMECO and the COPO facilities, which showed that much greater heat is transferred downwards for immiscible layers or before layers mix. A model for gap cooling and a set of simplified models for the system codes have been developed. MVITA code calculations have been performed for the Czech and Hungarian VVERs, towards evaluation of the in-vessel melt retention accident management scheme.Tests have been performed at the ULPU facility with organised flow for vessel external cooling. Considerable enhancement of the critical heat flux (CHF) was obtained. The ARVI Project has reached the halfway stage. This paper presents the results obtained thus far from the project.

AB - The cost-shared project ARVI (assessment of reactor vessel integrity) involves a total of nine organisations from Europe and USA. The objective of the ARVI Project is to resolve the safety issues that remain unresolved for the melt vessel interaction phase of the in-vessel progression of a severe accident. The work consists of experiments and analysis development. Four tests were performed in the EC-FOREVER Programme, in which failure was achieved in-vessels employing the French pressure vessel steel. The tests were analysed with the commercial code ANSYS-Multiphysics, and the codes SYSTUS+ and PASULA, and quite good agreement was achieved for the failure location. Natural convection experiments in stratified pools have been performed in the SIMECO and the COPO facilities, which showed that much greater heat is transferred downwards for immiscible layers or before layers mix. A model for gap cooling and a set of simplified models for the system codes have been developed. MVITA code calculations have been performed for the Czech and Hungarian VVERs, towards evaluation of the in-vessel melt retention accident management scheme.Tests have been performed at the ULPU facility with organised flow for vessel external cooling. Considerable enhancement of the critical heat flux (CHF) was obtained. The ARVI Project has reached the halfway stage. This paper presents the results obtained thus far from the project.

KW - critical heat flux

KW - nuclear reactor accidents

KW - nuclear reactor safety

KW - nuclear reactors

U2 - 10.1016/S0029-5493(02)00343-6

DO - 10.1016/S0029-5493(02)00343-6

M3 - Article

VL - 221

SP - 23

EP - 53

JO - Nuclear Engineering and Design

JF - Nuclear Engineering and Design

SN - 0029-5493

IS - 1-3

ER -

Sehgal BR, Theerthan A, Giri A, Karbojian A, Willschütz HG, Kymäläinen O et al. Assessment of reactor vessel integrity (ARVI). Nuclear Engineering and Design. 2003;221(1-3):23-53. https://doi.org/10.1016/S0029-5493(02)00343-6