Validation of the heat and mass transfer models within a pressurized water reactor containment using the International Standard Problem No. 47 data

• Autorzy: Bury T. , Składzień J. , Fic A.
• Języki publikacji: EN

Abstrakty

EN

A lumped parameter type code, called HEPCAL, has been worked out in the Institute of Thermal Technology of the Silesian University of Technology for simulations of a pressurized water reactor containment transient response to a loss-of-coolant accident. The HEPCAL code has been already verified and validated against available experimental data, which in fact have been taken from separate effect tests mainly. This work is devoted to validation of the latest version of the HEPCAL code against experimental data from more complex tests. These experiments have been performed on three different test rigs (called TOSQAN,MISTRA and ThAI) and a part of them became the basis of the International Standard Problem No. 47 (ISP-47) dedicated to containment thermal-hydraulics. Selected experiments realized within the framework of the ISP-47 project have been simulated using the HEPCAL-AD code. The obtained results allowed for drawing of some important conclusions concerning heat and mass transfer models (especially steam condensation), two-phase flow model and buoyancy effects.

Słowa kluczowe

EN

containment; ISP-47; lumped parameter code; PWR; thermal-hydraulic analysis

PL

analiza cieplno-przepływowa; powstrzymywanie

• Wydawca: Wydawnictwo Instytutu Maszyn Przepływowych PAN ; Komitet Termodynamiki i Spalania PAN
• Czasopismo: Archives of Thermodynamics
• Rocznik: 2012
• Tom: Vol. 33, no. 2
• Strony: 23--46
• Opis fizyczny: Bibliogr. 28 poz.,

Twórcy

autor

Bury T.

autor

Składzień J.

autor

Fic A.

• Silesian University of Technology, Institute of Thermal Technology, Konarskiego 22, 44-100 Gliwice, Poland,

Bibliografia

• [1] IAEA: Power Reactor Information System — PRIS http://www.iaea.org/programmes/a2/ – access on 12.01.2012.
• [2] IAEA: Development and Application of Level 1. Probabilistic Safety Assessment for Nuclear Power Plants. Specific Safety Guide No. SSG-3, International Atomic Energy Agency, Vienna 2010.
• [3] IAEA: Development and Application of Level 2. Probabilistic Safety Assessment for Nuclear Power Plants Specific Safety Guide No. SSG-4, International Atomic Energy Agency, Vienna 2010.
• [4] IAEA: Deterministic Safety Analysis for Nuclear Power Plants. Specific Safety Guide No. SSG-2, International Atomic Energy Agency, Vienna 2009.
• [5] NEA: Validation Matrix for the Assessment of Thermal-Hydraulic Codes for VVER LOCA and Transients. Nuclear Energy Agency Report No. NEA/CSNI/R(2001)4, Paris 2001.
• [6] NEA: State Of the Art Report on Containment Thermalhydraulics and Hydrogen Distribution. Nuclear Energy Agency Report No. NEA/CSNI/R(99)16, Paris 1999.
• [7] Sehgal B.R.: Accomplishments and challenges of the severe accident research. Nuclear Engineering and Design 210(2001), 79–94.
• [8] Yadigaroglu G., Andreani M., Dreier, J., Coddington P.: Trends and needs in experimentation and numerical simulation for lwr safety. Nuclear Engineering and Design 221(2003), 205–223.
• [9] IAEA: Use of Computational Fluid Dynamics Codes for Safety Analysis of Nuclear Reactor Systems. International Atomic Energy Agency, IAEA-TECDOC-1379, Vienna 2003.
• [10] NEA: CSNI International Standard Problems (ISP). Nuclear Energy Agency Report No. NEA/CSNI/R(2000)5, Paris 2000.
• [11] NEA: OECD Standard Problem ISP23: Rupture of a Large Diameter Pipe Within the HDR Containment – Specification Nuclear Energy Agency Report No. NEA/CSNI Report, Paris 1988.
• [12] NEA: OECD Standard Problem ISP29: Distribution of Hydrogen Within the HDR Containment under Severe Accident Conditions. Nuclear Energy Agency Report No. NEA/CSNI, Paris 1993.
• [13] NEA: Final comparison Report on ISP35 NUPEC Hydrogen Mixing and Distribution Test (Test M-7-1). Nuclear Energy Agency Report No. NEA/CSNI/R, Paris 1994.
• [14] NEA: OECD Standard Problem ISP37: VANAM-M3 — A Multi-Compartment Aerosol Depletion Test with Hygroscopic Aerosol Material: Comparison Report. Nuclear Energy Agency Report No. NEA/CSNI/R(1996)26, Paris 1996.
• [15] Fic A., Skorek J.: Mathematical Model of Transient Thermal and Flow Processes in Containment of a PWR Nuclear Reactor. Archives of Energy (1993), 22, 1-2, 19–32.
• [16] Skorek J., Składzień J.: Thermal Analysis of the Loss-of-Coolant Accident Within the Containment of the WWER-440 and WWER-1000 Nuclear Reactors. Computer Assisted Mechanics and Engineering Sciences 1(1994), 217–231.
• [17] Bury T.: Analysis of Thermal and Flow Processes within Containments of Water Nuclear Reactors During Loss-of-Coolant Accidents. PhD thesis, Silesian University of Technology, Gliwice 2005 (in Polish).
• [18] Hobler T.: Heat Transport and Heat Exchangers. WNT, Warsaw 1971 (in Polish).
• [19] Marshall J., Holland P.G., Woodman W.: OECD/CSNI Containment Analysis Standard Problem No. 3. Experimental Results. CASP3-1, OECD, Paris 1981.
• [20] Tagami T.: Interim Report on Safety Assessment and Facilities, Establishment Project in Japan for Period Ending June 1965 (No. 1). Japanese Atomic Energy Research Institute, Tokio 1965.
• [21] Hobler T.: Diffusive Mass Transport and Absorbers. WNT, Warsaw 1976 (in Polish).
• [22] Welty J.R., Wicks C.E., Wilson R.E., Rorrer G.: Fundamentals of Momentum, Heat and Mass Transfer. John Wiley&Sons Inc., New York 2001.
• [23] Stempniewicz M.M.: Simulation of Containment Transient Response During Accidents in Advanced Reactor Types — the Computer Code SPECTRA. PhD thesis., Silesian University of Technology, Gliwice, NRG Arnhem, Holland, 2000.
• [24] NEA: International Standard Problem ISP-47 on Containment Thermal Hydraulics. Nuclear Energy Agency Report No. NEA/CSNI/R(2007)10, Paris 2007.
• [25] Vendel J., Cornet P., Malet J., Porcheron E., Paillere H., Carlon-Charles M.L., Studer E., Fischer K., Allelein H.J.: ISP 47 on Containment Thermal-Hydraulics. Computer Codes Exercise Based on TOSQAN, MISTRA and ThAI Experiments. In: Proc. of the Eurosafe Forum, Berlin 2002, http://www.eurosafe-forum.org/eurosafe-2002
• [26] Bury T.: Evaluation of the in-containment heat transfer coefficient determination methods for loss-of-coolant accidents simulation purposes. In: Proc. 7thWorld Conference on Experimental Heat Transfer, Fluid Flow and Thermodynamics, Kraków 2009.
• [27] Bury T.: Validation of the heat and mass transfer models within containments of pressurized water reactors. In: Proc. 13th International Symposium on Heat Transfer and Renewable Sources of Energy, Szczecin – Międzyzdroje 2010.
• [28] Bury T., Składzień J.: Simulations of loss-of-coolant accidents for containments of the second and the third PWR generation. Archives of Thermodynamics 27(2006), 4, 45–56.