ERMSAR 2012, Cologne March 21 – 23, 2012 Post-test calculations of CERES experiments using ASTEC code Lajos Tarczal 1, Gabor Lajtha 2 1 Paks Nuclear Power.

Slides:



Advertisements
Similar presentations
Hongjie Zhang Purge gas flow impact on tritium permeation Integrated simulation on tritium permeation in the solid breeder unit FNST, August 18-20, 2009.
Advertisements

INRNE-BAS MELCOR Pre -Test Calculation of Boil-off test at Quench facility 11th International QUENCH Workshop Forschungszentrum Karlsruhe (FZK), October.
Lesson 17 HEAT GENERATION
Author: Cliff B. Davis Evaluation of Fluid Conduction and Mixing Within a Subassembly of the Actinide Burner Test Reactor.
Jiří Duspiva Nuclear Research Institute Řež, plc. Nuclear Power and Safety Division Dept. of Reactor Technology 11 th International QUENCH Workshop Karlsruhe,
Thermal-Hydraulic Transient Analysis of the Missouri University Research Reactor (MURR) TRTR Annual Meeting September 17-20, 2007 Dr. Robert C. Nelson1,
Institute for Electric Power Research Co. International Workshop On Level 2 PSA and Severe Accident Management Cologne, Germany 29.
October 25-27, th International QUENCH Workshop 1 Top Flooding Experiments and Modeling Estelle Brunet-Thibault (EDF), Serge Marguet (EDF)
HTTF Analyses Using RELAP5-3D Paul D. Bayless RELAP5 International Users Seminar September 2010.
Jennifer Tansey 12/15/11. Introduction / Background A common type of condenser used in steam plants is a horizontal, two- pass condenser Steam enters.
Dakota Nickerson, Kyle Pflueger, Adam Leschber, Andrew Dahlke M.E. Undergrad 1.
Spiral Condensors. Working Principle The Spiral Condensor consists of two sheets stainless steel strips which have been wounded from the centre.
1 Wilfrid Farabolini 23 feb. 04Direction de l’Energie Nucléaire Tritium Control A Major Issue for a Liquid Metal Blanket.
GT – MHR Reactor Cavity Cooling System
Argonne National Laboratory 2007 RELAP5 International User’s Seminar
23 Jan 2007 LASA Cryogenics Global Group 1 ILC Cryomodule piping L. Tavian for the cryogenics global group.
Thermal hydraulic analysis of ALFRED by RELAP5 code & by SIMMER code G. Barone, N. Forgione, A. Pesetti, R. Lo Frano CIRTEN Consorzio Interuniversitario.
Thermal Hydraulic Simulation of a SuperCritical-Water-Cooled Reactor Core Using Flownex F.A.Mngomezulu, P.G.Rousseau, V.Naicker School of Mechanical and.
RIC 2009 Thermal Hydraulics & Severe Accident Code Development & Application Ghani Zigh USNRC 3/12/2009.
Advanced Test Reactor.
17th Symposium of AER, Yalta, Crimea, Ukraine, Sept , 2007.
ASTEC validation on PANDA tests A. BENTAIB, A. BLEYER Institut de Radioprotection et de Sûreté Nucléaire BP 17, Fontenay aux Roses Cedex, FRANCE.
LBE-Water interaction in LIFUS V facility under different operating conditions A. Ciampichetti, D. Bernardi - ENEA T. Cadiou - CEA N. Forgione – Università.
KIT – University of the State of Baden-Württemberg and National Large-scale Research Center of the Helmholtz Association Institute for Nuclear and Energy.
Department of Mechanical and Nuclear Engineering Reactor Dynamics and Fuel Management Group Comparative Analysis of PWR Core Wide and Hot Channel Calculations.
2-D Heat Transfer Model of A Horizontal U-Tube M. S. Islam 1, A. Fujimoto 2, A. Saida 2 and T. Fukuhara 2 2-D Heat Transfer Model of A Horizontal U-Tube.
Nuclear Thermal Hydraulic System Experiment
1 Numerical study of the thermal behavior of an Nb 3 Sn high field magnet in He II Slawomir PIETROWICZ, Bertrand BAUDOUY CEA Saclay Irfu, SACM Gif-sur-Yvette.
IAEA Meeting on INPRO Collaborative Project “Performance Assessment of Passive Gaseous Provisions (PGAP)” December, 2011, Vienna A.K. Nayak, PhD.
Experimental and numerical studies on the bonfire test of high- pressure hydrogen storage vessels Prof. Jinyang Zheng Institute of Process Equipment, Zhejiang.
Convection: Internal Flow ( )
ERMSAR 2012, Cologne March 21 – 23, The Experimental Results of LIVE-L8B: Debris Melting Process in a Simulated PWR Lower Head X. Gaus-Liu, A. Miassoedov,
ERMSAR 2012, Cologne March 21 – 23, 2012 ESTIMATION OF THERMAL-HYDRAULIC LOADING FOR VVER-1000 UNDER SEVERE ACCIDENT SCENARIO Barun Chatterjee 1, Deb Mukhopadhyay.
ERMSAR 2012, Cologne March 21 – 23, 2012 MELCOR Severe Accident Simulation for a “CAREM-like” Integral Reactor M. Caputo, J. M. García, M. Giménez, S.
ERMSAR 2012, Cologne March 21 – 23, 2012 CONDUCT AND ANALYTICAL SUPPORT TO AIR INGRESS EXPERIMENT QUENCH-16 J. BIRCHLEY 1, L. FERNANDEZ MOGUEL 1, C. BALS.
Wir schaffen Wissen – heute für morgen A. Dehbi, D. Suckow, T. Lind, S. Guentay Paul Scherrer Institut, Switzerland Large Scale Experimental Program at.
ERMSAR 2012, Cologne March 21 – 23, 2012 Analysis of Corium Behavior in the Lower Plenum of the Reactor Vessel during a Severe Accident Rae-Joon Park,
©SJA Søren Juhl Andreasen and Søren Knudsen Kær Aalborg University Institute of Energy Technology Dynamic Model of High Temperature PEM Fuel Cell.
ERMSAR 2012, Cologne March 21 – 23, 2012 Analysis and interpretation of the LIVE-L6 experiment A. Palagin, A. Miassoedov, X. Gaus-Liu (KIT), M. Buck (IKE),
E3 Teacher Summer Research Program. Willie L. Smith - IPC, Physics Tidehaven ISD Tidehaven High School.
ERMSAR 2012, Cologne March 21 – 23, 2012 Authors: PANTYUSHIN S.I., FRIZEN Е.А., SEMISHKIN V.P., BUKIN N.V, BYKOV M.А., MOKHOV V.А. (OKB «GIDROPRESS», Podolsk.
Modeling a Steam Generator (SG)
ERMSAR 2012, Cologne March 21 – 23, 2012 In-vessel retention as retrofitting measure for existing nuclear power plants M. Bauer, Westinghouse Electric.
Enhanced heat transfer in confined pool boiling
ERMSAR 2012, Cologne March 21 – 23, 2012 Experimental and computational studies of the coolability of heap-like and cylindrical debris beds E. Takasuo,
DCLL ½ port Test Blanket Module thermal-hydraulic analysis Presented by P. Calderoni March 3, 2004 UCLA.
Results of First Stage of VVER Rod Simulator Quench Tests 11th International QUENCH Workshop Forschungszentrum Karlsruhe October 25-27, 2005 Presented.
ERMSAR 2012, Cologne March 21 – 23, 2012 EXPERIMENTAL STUDY OF HYDROGEN COMBUSTION DURING DCH EVENTS IN TWO DIFFERENT SCALES Giancarlo Albrecht Leonhard.
EUROPEAN ORGANIZATION FOR NUCLEAR RESEARCH Design of the thermosiphon Test Facilities Thermosiphon Cooling Review A. MORAUX PH Dpt / DT Group CERN SEPTEMBER.
Bay Zoltán Foundation for Applied Reseach Institute for Logistics and Production Systems BAY-LOGI Assessment of crack like defect in dissimilar welded.
1 Improvement of RELAP5 Models for Condensation of Steam and Steam-Gas Mixture in Horizontal and Inclined Tubes Pavel KRÁL NURETH-16, Chicago, 2015.
Plant & Reactor Design Passive Reactor Core Cooling System
CONTROL AND SAFETY of Nuclear Steam Supply Systems (NSSS)
PHWR Safety 2014 / CANSAS-2014 Workshop 2014 June Jun Yang AECL
Algirdas Kaliatka, Audrius Grazevicius, Eugenijus Uspuras
A.Borovoi, S.Bogatov, V.Chudanov, V.Strizhov
Design of the thermosiphon Test Facilities 2nd Thermosiphon Workshop
International Topical Meeting on Nuclear Reactor Thermal Hydraulics
Chapter 8: Internal Flow
Thermodynamics Thermal Hydraulics.
Phase III Indo-UK Collaboration
Produktentwicklung und Maschinenelemente
From: Analysis of Late Phase Severe Accident Phenomena in CANDU Plant
Analysis of Reactivity Insertion Accidents for the NIST Research Reactor Before and After Fuel Conversion J.S. Baek, A. Cuadra, L-Y. Cheng, A.L. Hanson,
Session Name: Lessons Learned from Mega Projects
ESS elliptical cryomodule
NEW ARISTON STORAGE BOILERS March, 2009.
NUMERICAL STUDY OF IN-VESSEL CORIUM RETENTION IN A BWR REACTOR M
First results of the bundle test QUENCH-L2 with M5® claddings
State Scientific Center– Research Institute of Atomic Reactors
Presentation transcript:

ERMSAR 2012, Cologne March 21 – 23, 2012 Post-test calculations of CERES experiments using ASTEC code Lajos Tarczal 1, Gabor Lajtha 2 1 Paks Nuclear Power Plant 2 NUBIKI Nuclear Safety Research Institute, Budapest

ERMSAR 2012, Cologne March 21 – 23, 2012 Outline Concept of VVER-440/213 reactor vessel external cooling in Paks NPP Introduction of CERES facility ASTEC code, ASTEC model of the CERES facility Post test calculations, comparison with Test 2 results Conclusions 2

ERMSAR 2012, Cologne March 21 – 23, 2012 Background – RV external cooling in VVER Ventillation duct Drain valve Reactor cavity Connection corridor

ERMSAR 2012, Cologne March 21 – 23, 2012 Background – RV external cooling in VVER-440 Natural circulation cooling circuit Problems: Significant uncertainty of heat flux assessment on reactor vessel wall caused by molten pool (surface, heat flux) VVER-440 specific flow channel geometry: The width of the narrowest gap in the flow channel can be 8-10 mm. 4

ERMSAR 2012, Cologne March 21 – 23, 2012 CERES test facility Cooling Effectiveness on Reactor External Surface Aim of experiment: Experimental verification the effectiveness of reactor vessel external cooling Geometry: 1:40 volume ratio of flow channel and elliptic bottom of reactor vessel ( 9° degree section of vessel) Vertical dimensions in 1:1 height rate, because of correct modelling of the natural circulation Main features: Variable minimal gap width in the narrowest cross sections Variable heat flux profile: Varying the amount of heater rods and heater power 5

ERMSAR 2012, Cologne March 21 – 23, 2012 CERES test facility 6

ERMSAR 2012, Cologne March 21 – 23, 2012 CERES test facility Heat flux profile along the reactor vessel wall 7 LOCA 200 mm: Distribution of heat flux at RPV outer surface at different times (x axis represents spread distance from RPV bottom to the top along the surface) Preset heat flux distribution in CERES facility Heat flux profile calculated with ASTEC code Total heat power: 160 kW Heat flux peak occurs right at the elvation of narrowest gap 105 kW (516 kW/m 2 )

ERMSAR 2012, Cologne March 21 – 23, 2012 Summary of the performed experiments 8 1. test2. test3. test4. test Critical gap width [mm]209,7 4,7-9,7 (asymmetric) 3-10,1 (asymmetric) Cross section of critical gap [mm 2 ] Heat power (narrow gap/total) [kW]98/145105/160104/159108/166 Maximum heat flux [kW/m 2 ] Flooding water maximum temperature [C°] Diameter of outlet [mm] test2. test3. test4. test Maximum surface temperature of heater blocks [C°] Permanent surface temperature of heater blocks[C°] Maximum flooding flow rate [liter/hour] The 2. Test is conservative and the most realistic case

ERMSAR 2012, Cologne March 21 – 23, 2012 ASTEC severe accident code 9 CESAR module Modelling the primary and secondary circuits thermal hydraulics 5 equation model Capability of treatment of non- condensable gases (H 2, N 2 ) Basic elements: VOLUME JUNCTION WALL CONNECTI (Boundary conditions) SYSTEMS (system elements)

ERMSAR 2012, Cologne March 21 – 23, 2012 ASTEC model of CERES facility ASTEC v2.0 rev0 5 nodes for elliptic vessel bottom model (V01- V05) 6 nodes for vertical flow channel model (V06- V11) Flooding tank temperature control with heated wall, as user defined boundary condition Electrical heater blocks power defined with „HEAT connection” as boundary condition Modelling the water surface behaviour on atmospheric pressure („open pool”) in the upper region of flow channel and in the flooding tank, the environment node was filled with N 2 gas 10 Narrow gaps

ERMSAR 2012, Cologne March 21 – 23, 2012 Post-test calculations of Test 2 Heater power – boundary condition 11

ERMSAR 2012, Cologne March 21 – 23, 2012 Post-test calculations of Test 2 Inlet flow 12 ~11 min~4 min

ERMSAR 2012, Cologne March 21 – 23, 2012 Post-test calculations of Test 2 Inlet flow 13 Backflow The maximum values of black-flow are higher than in the experiment. Calculated period ~600 s Measured peridod ~500 s

ERMSAR 2012, Cologne March 21 – 23, 2012 Post-test calculations of Test 2 Flooding water temperature – boundary condition 14

ERMSAR 2012, Cologne March 21 – 23, Post-test calculations of Test 2 Water and wall temperatures in bottom part

ERMSAR 2012, Cologne March 21 – 23, 2012 Post-test calculations of Test 2 Water and wall temperatures in bottom part 16

ERMSAR 2012, Cologne March 21 – 23, Post-test calculations of Test 2 Water and wall temperatures in narrowest section The periodic „build up-collapse” of the steam plug shows good agreement with the experiment. Small difference in temperature maximum and minimum values.

ERMSAR 2012, Cologne March 21 – 23, 2012 Post-test calculations of Test 2 Water and wall temperatures in narrowest section 18 The periodic „build up-collapse” of the steam plug shows good agreement with the experiment. Small difference in temperature maximum and minimum values.

ERMSAR 2012, Cologne March 21 – 23, 2012 Post-test calculations of Test 2 Water temperatures in the upper section 19

ERMSAR 2012, Cologne March 21 – 23, 2012 Conclusions The ASTEC calculated values are in a reasonable agreement with the experimental results. The code can correctly model the main physical phenomena in the heated channel; Well predicted tendencies, flow and temperature oscillations; The deviations between the measured and calculated values are in 20% range and can be explained; The VVER-440 reactor vessel wall in the narrowest section can be cooled near stable, °C temperature. 20 Implementation at Paks Unit 1 in 2011:

ERMSAR 2012, Cologne March 21 – 23, 2012 Thank you for your attention! 21