In vessel corium propagation sensitivity study of reactor pressure vessel rupture time with PROCOR platform

• Autorzy: Skrzypek E. , Skrzypek M. , Saas L. , Letellier R.
• Języki publikacji: EN

Abstrakty

EN

The problem of corium propagation for PWRs in the Reactor Pressure Vessel (RPV) and the timing of RPV failure is one of the main issues of study in the area of severe accidents. The PROCOR numerical platform created by the CEA severe accident laboratory is modelling corium propagation for LWRs, its relocation to the Lower Plenum and RPV failure. The idea behind the platform was to provide a tool that is fast enough to be able to perform numerous calculations within a reasonable time frame in order to deliver a statistical study. Work on the development of models that describe in-vessel issues is being pursued through simplified phenomena modelling, their verification and sensitivity studies. Recent activities related to PROCOR development involved cooperation between French CEA experts and Polish PhD students, who were engaged in the topics of core support plate modelling and analysis of the phenomena occurring in a thin metallic layer on top of the corium pool. Those issues were identified as strongly influencing the course of severe accidents and the timing of RPV failure. In some sensitivity studies performed on a given generic high power Light Water Reactor with heavy reflector, two groups of RPV ruptures were distinguished related to the two issues, which provided motivation for further work on these topics. The paper will present a sensitivity study of corium propagation in order to identify the relevance of those two issues for the timing of RPV rupture.

Słowa kluczowe

EN

sensitivity study; PROCOR platform; IVMR strategy

PL

badanie wrażliwości; platforma PROCOR; strategia IVMR

• Wydawca: Institute of Heat Engineering, Warsaw University of Technology
• Czasopismo: Journal of Power Technologies
• Rocznik: 2017
• Tom: Vol. 97, nr 2
• Strony: 110--116
• Opis fizyczny: Bibliogr. 19 poz., rys., tab., wykr.

Twórcy

autor

Skrzypek E.

• National Center for Nuclear Research,7 Andrzeja Soltana Street, 05-400 Otwock, Poland
• Warsaw University of Technology, 21/25 Nowowiejska Street, 00-665 Warsaw, Poland

autor

Skrzypek M.

• National Center for Nuclear Research,7 Andrzeja Soltana Street, 05-400 Otwock, Poland
• Warsaw University of Technology, 21/25 Nowowiejska Street, 00-665 Warsaw, Poland

autor

Saas L.

• CEA Cadarache, DEN/DTN/SMTA/LPMA, 13108 Saint-Paul-Lez-Durance, France

autor

Letellier R.

• CEA Cadarache, DEN/DTN/SMTA/LPMA, 13108 Saint-Paul-Lez-Durance, France

Bibliografia

• [1] F. Fichot, J.-M. Bonnet, B.Chaumont, Irsn views and perspectives on in-vessel melt retention strategy for severe accident mitigation, in: EUROSAFE Forum 2015, Brussels, Belgium, 2015.
• [2] B.R.Sehgal, Nuclear safety in light water reactors: Severe accident phenomenolog, Elsevier, (2012) 550–551.
• [3] M. Sangiorgi, In-vessel melt retention (ivmr) analysis of a vver-1000 npp, in: 6th ASTEC user’s club / 2nd CESAM workshop, 2015.
• [4] R. LeTellier, L. Saas, F. Payot (Eds.), Phenomenological analyses of corium propagation in LWRs: the PROCOR software platform, Marseille, France, 2015, eRMSAR 2015.
• [5] L. Saas, R. L. Tellier, S. Bajard, A simplified geometrical model for transient corium propagation in core for an lwr with heavy reflector, in: International Congress on Advances in Nuclear Power Plants, 2015.
• [6] R. LeTellier, L. Saas, S. Bajard, Transient stratification modelling of a corium pool in a lwr vessel lower head, Nuclear Engineering and Design 287 (2015) 68–77.
• [7] F. Gaudier, Uranie : The cea/den uncertainty and sensitivity platform, Procedia Social and Behavioral Sciences 2, Elsevier Ltd. (2010) 7660–7661.
• [8] D. Skrien, Object-Oriented Design Using Java (Jan. 2008).
• [9] ROOT Data Analysis Framework, User’s Guide (May 2014).
• [10] H. Loeffler, J. Peschke, M. Sonnenkalb, Classical event tree analysis and dynamic event tree analysis for high pressure core melt accidents in a german pwr, in: OECD International Workshop on Level 2 PSA and Severe Accident Management, Koeln, Germany, 2004.
• [11] P.Darnowski, E.Skrzypek, P. Mazgaj, K. Swirski, P. Gandrille, Total loss of ac power analysis for epr reactor, Nuclear Engineering and Design 289 (2015) 8–18.
• [12] A. Bonelli, O. Mazzantini, M. Sonnenkalb, Station black-out analysis with melcor 1.8.6 code for atucha 2 nuclear power plant, Science and Technology of Nuclear Installations 2012.
• [13] J. S. et al., Equations for solidification of corium without sparging gas - scaling criteria, in: OECD workshop on ex-vessel debris coolability, Karlsruhe, Germany, 1999.
• [14] I.Lindholm, A review of dryout heat fluxes and coolabiliy of particle beds, Tech. Rep. APRI 4, Stage 2 Report, VTT Energy, Finland (Apr. 2002).
• [15] H. Esmaili, M. Khatib-Rahbar, Analysis of in-vessel retention and exvessel fuel coolant interaction for ap1000, Tech. Rep. NUREG/CR- 6849 ERI/NRC04-201, U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research (2004).
• [16] S. Globe, D. Dropkin, Natural convection heat transfer in liquids confined by two horizontal plates and heated from below, Journal of Heat Transfer 81 (1959) 24–28.
• [17] S. Churchill, H. Chu, Correlating equations of laminar rand turbulent free convection from a vertical plate, International Journal of Heat and Mass Transfer 18 (1975) 1323–1329.
• [18] T. Chawla, S. Chan, Heat transfer from vertical/inclined boundaries of heat-generating boiling pools, Journal of Heat Transfer 104 (1982) 465–473.
• [19] J. Bonnet, J. Seiler, Thermohydraulic phenomena in corium pool: the bali experiment, in: ICONE 7, Tokyo, Japan, 1999.

Uwagi

PL

Opracowanie ze środków MNiSW w ramach umowy 812/P-DUN/2016 na działalność upowszechniającą naukę (zadania 2017).