FZK/IRS-AS W. Hering, Ch. Homann1 Forschungszentrum Karlsruhe in der Helmholtz-Gemeinschaft 11th International QUENCH Workshop, Karlsruhe, Germany W. Hering,

Slides:



Advertisements
Similar presentations
Slide Nov 2006, EFDA PWI meeting, LjubljanaI.S. Landman, FZ-Karlsruhe Modelling on Wall Surfaces and Tokamak Plasma Consequences of ITER Transient.
Advertisements

Topical Information Meeting for the 3rd Call for Proposals of FP6
Forschungszentrum Karlsruhe in der Helmholtz-Gemeinschaft 1 J. Stuckert, FZK/IMF-III11 th International QUENCH Workshop, Karlsruhe, October 25-27, 2005.
0 Tilman Drath, Ingo D. Kleinhietpaß, Marco K. Koch Lehrstuhl für Energiesysteme und Energiewirtschaft (LEE) Ruhr-Universität Bochum (RUB) 11 th International.
Use of Fission Product Release results for IRSN safety studies Dr. Grégory NICAISE IRSN / DSR (Reactor Safety Department) International VERCORS Seminar.
Forschungszentrum Karlsruhe in der Helmholtz-Gemeinschaft M. Steinbrück, FZK/IMF-I 11th Int. QUENCH Workshop, Karlsruhe, October 25-27, Oxidation.
25-27/10/2005, 1 Laboratory for Thermal Hydraulics Nuclear Energy and Safety 11th QUENCH Workshop, FZ Karlsruhe Calculational Support for the QUENCH-10.
IAEA International Atomic Energy Agency Simulators in an e-learning platform D. Beraha.
OVERVIEW - RELAP/SCDAPSIM
11 th International QUENCH Workshop - Karlsruhe - October 25-27, Reflooding of a degraded core with ICARE/CATHARE V2 Florian Fichot 1 - Fabien Duval.
ITER Computational Shield Benchmark IAEA TM on Nuclear Data Library for Advanced Systems – Fusion Devices. Vienna, October 31 – November 2, 2007 U. Fischer,
INRNE-BAS MELCOR Pre -Test Calculation of Boil-off test at Quench facility 11th International QUENCH Workshop Forschungszentrum Karlsruhe (FZK), October.
11th International QUENCH Workshop, October 25-27, 2005, Forschungszentrum Karlsruhe, Germany 1 Evaluation of the Oxygen-Induced Zircaloy Embrittlement.
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,
1 Application of the SVECHA/QUENCH code to the simulation of the QUENCH bundle tests Q-07 and Q-08 Presented by A.V.Palagin* Nuclear Safety Institute (IBRAE)
PRACTICAL EXAMPLES OF THE ANALYSIS OF SEVERE ACCIDENTS Presented Dr. Chris Allison Regional Workshop on Evaluation of Specific Preventative and Mitigative.
Modeling boiling water reactor main steam isolation valve leakage using MELCOR Presented at the 21 st Annual Regulatory Information Conference March 10-12,
UNIVERSITÀ DI PISA GRUPPO DI RICERCA NUCLEARE – SAN PIERO A GRADO (GRNSPG) Any reproduction, alteration, transmission to any third party or publication.
October 25-27, th International QUENCH Workshop 1 Top Flooding Experiments and Modeling Estelle Brunet-Thibault (EDF), Serge Marguet (EDF)
Reliability Prediction of a Return Thermal Expansion Joint O. Habahbeh*, D. Aidun**, P. Marzocca** * Mechatronics Engineering Dept., University of Jordan,
HTTF Analyses Using RELAP5-3D Paul D. Bayless RELAP5 International Users Seminar September 2010.
Prior investigation of absorber rod Dy 2 O 3  TiO 2 behavior after severe accident test FSUE SRI SIA “LUCH” ICP MAE Presented by Dmitry N. Ignatiev Forschungszentrum,
Investigation of "dry" recriticality of the melt during late in-vessel phase of severe accident in Light Water Reactor D.Popov, KNPP, BG O.Runevall, KTH,
EUROTRANS DESIGN WP 1.5 Meeting May 22nd/23rd 2007 Stockholm, Sweden Recommendations for the MOX fuel conductivity and heat transfer correlations to be.
May 22nd & 23rd 2007 Stockholm EUROTRANS: WP 1.5 Task Containment Assessment IP-EUROTRANS DOMAIN 1 Design WP 1.5 Safety Assessment of the Transmutation.
Forschungszentrum Karlsruhe in der Helmholtz-Gemeinschaft NDACC H2O workshop, Bern, July 2006 Water vapour profiles by ground-based FTIR Spectroscopy:
1 Safety studies for MYRRHA B. Arien, S. Heusdains, H. Aït Abderrahim on behalf of the MYRRHA Team and Support IP-Eurotrans Workshop DM1-WP1.5Brussels,
KIT – University of the State of Baden-Wuerttemberg and National Laboratory of the Helmholtz Association Institute for Nuclear and Energy Technologies.
SISIFO-GAS A COMPUTERIZED SYSTEM TO SUPPORT SEVERE ACCIDENTS TRAINING AND MANAGEMENT WGRisk Workshop March 29-31, 2004 Köln, Germany César Serrano.
KIT – Universität des Landes Baden-Württemberg und nationales Forschungszentrum in der Helmholtz-Gemeinschaft Karlsruhe Institute of Technology, Germany.
KIT – University of the State of Baden-Wuerttemberg and National Research Center of the Helmholtz Association Institute for Nuclear and Energy Technologies.
RIC 2009 Thermal Hydraulics & Severe Accident Code Development & Application Ghani Zigh USNRC 3/12/2009.
Idaho National Engineering and Environmental Laboratory Assessment of Margin for In-Vessel Retention in Higher Power Reactors 2004 RELAP5 International.
17th Symposium of AER, Yalta, Crimea, Ukraine, Sept , 2007.
Generation Aino Ahonen CABABILITY OF APROS IN THE ANALYSES OF DIESEL LOADING SEQUENCES E. Raiko, H.Kontio, K.Porkholm, presented by A. Ahonen.
Nuclear Research Institute Řež plc 1 DEVELOPMENT OF RELAP5-3D MODEL FOR VVER-440 REACTOR 2010 RELAP5 International User’s Seminar West Yellowstone, Montana.
Department of Mechanical and Nuclear Engineering Reactor Dynamics and Fuel Management Group Comparative Analysis of PWR Core Wide and Hot Channel Calculations.
Accuracy Based Generation of Thermodynamic Properties for Light Water in RELAP5-3D 2010 IRUG Meeting Cliff Davis.
1 Context-dependent Product Line Practice for Constructing Reliable Embedded Systems Naoyasu UbayashiKyushu University, Japan Shin NakajimaNational Institute.
EUROTRANS – DM1 ENEA Activities on EFIT Safety Analysis ENEA – FIS/NUC Bologna - Italy WP5.1 Progress Meeting Tractebel / Brussels, March 17, 2006 G. Bandini,
IAEA Meeting on INPRO Collaborative Project “Performance Assessment of Passive Gaseous Provisions (PGAP)” December, 2011, Vienna A.K. Nayak, PhD.
RELAP5-3D Uncertainty Analysis A.J. Pawel and Dr. George L. Mesina International RELAP Users’ Seminar 2011 July 25-28, 2011.
TACIS Project: R8.01/98 – TRANSLATION, EDITING AND DIFFUSION OF DOCUMENTS (Result Dissemination) Probabilistic Safety Analysis Technology (PSA) TACIS R3.1/91.
1 ROD BUNDLE HEAT TRANSFER TEST RESULTS; Spacer Grid Effects and Potential Impacts on LOCA Evaluation Models Stephen M. Bajorek, Ph.D. Senior Technical.
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.
ERMSAR 2012, Cologne March 21 – 23, ON THE ROLE OF VOID ON STEAM EXPLOSION LOADS.
ERMSAR 2012, Cologne March 21 – 23, 2012 In-vessel retention as retrofitting measure for existing nuclear power plants M. Bauer, Westinghouse Electric.
ERMSAR 2012, Cologne March 21 – 23, 2012 Experimental and computational studies of the coolability of heap-like and cylindrical debris beds E. Takasuo,
Bulatom Confernce June 2007
Results of First Stage of VVER Rod Simulator Quench Tests 11th International QUENCH Workshop Forschungszentrum Karlsruhe October 25-27, 2005 Presented.
Paul Alexander 2 nd SKADS Workshop October 2007 SKA and SKADS Costing The Future Paul Alexander Andrew Faulkner, Rosie Bolton.
Forschungszentrum Karlsruhe in der Helmholtz-Gemeinschaft FZK, H & HQWS11, KA, Analysis and Comparison of Experimental Data of QUENCH-07.
ERMSAR 2012, Cologne March 21 – 23, 2012 OECD Benchmark Exercise on the TMI-2 Plant: Analysis of an Alternative Severe Accident Scenario G. Bandini (ENEA),
ERMSAR 2012, Cologne March 21 – 23, 2012 GENERIC CONTAINMENT A first step towards bringing (European) containment simulations to a common level St. Kelm.
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.
Use and Conduct of Safety Analysis IAEA Training Course on Safety Assessment of NPPs to Assist Decission Making Workshop Information IAEA Workshop Lecturer.
Hot-Spot Temperature Experiment Chats Workshop 10 th October 2013 Kamil Sedlak, Pierluigi Bruzzone EPFL-CRPP, Villigen, Switzerland.
Validation of Traditional and Novel Core Thermal-Hydraulic Modeling and Simulation Tools Issues in Validation Benchmarks: NEA OECD/US NRC NUPEC BWR Full-size.
A.Borovoi, S.Bogatov, V.Chudanov, V.Strizhov
Panel Discussion: Discussion on Trends in Multi-Physics Simulation
Design of the thermosiphon Test Facilities 2nd Thermosiphon Workshop
VICTOR HUGO SANCHEZ ESPINOZA and I. GÓMEZ-GARCÍA-TORAÑO
IRSN work and perspectives
I. Di Piazza (ENEA), R. Marinari, N. Forgione (UNIPI), F
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:

FZK/IRS-AS W. Hering, Ch. Homann1 Forschungszentrum Karlsruhe in der Helmholtz-Gemeinschaft 11th International QUENCH Workshop, Karlsruhe, Germany W. Hering, Ch. Homann W. Hering, Ch. Homann Forschungszentrum Karlsruhe Programme NUKLEAR P.O. Box 3640, D Karlsruhe, Germany 11th International QUENCH Workshop, October Table of Contents Motivation and objectives Imbedding of Q-11 into Reflood Database Status of Q-11 preparation Summary and conclusions Pre-test calculations of QUENCH-11 (Q-L2) using S/R5 and ASTEC

FZK/IRS-AS W. Hering, Ch. Homann2 Forschungszentrum Karlsruhe in der Helmholtz-Gemeinschaft 11th International QUENCH Workshop, Karlsruhe, Germany Motivation and objectives Open issues in SFD core refloodOpen issues in SFD core reflood – –Reduction in H 2 uncertainty – –Perform a dry-out-reflood sequence test (Q-11) – –Reflood with low capability systems (Q-11) – –Assess risk of unintended core reflood (in case of LOOP) Pre-test work for QUENCH-11Pre-test work for QUENCH-11 – –Feasibility study (QWS-10) – –Upgrade facility to meet requirements – –First results of Q-11 pre-test experiments (Juri Stuckert) – –Specification of step-by step approach to meet requirements of QUENCH-11 (“Vorversuche”)

FZK/IRS-AS W. Hering, Ch. Homann3 Forschungszentrum Karlsruhe in der Helmholtz-Gemeinschaft 11th International QUENCH Workshop, Karlsruhe, Germany  Data base – –EU-Programs – –FZK Experiments – –IRSN Phebus – –OECD/NEA – –USNRC – –Plant accidents – –LUTCH  Reactor type – –PWR – –VVER – –BWR Imbedding of Q-11 into Reflood Database

FZK/IRS-AS W. Hering, Ch. Homann4 Forschungszentrum Karlsruhe in der Helmholtz-Gemeinschaft 11th International QUENCH Workshop, Karlsruhe, Germany Q-L2 in the reflood map Experimental data base: Depending on Reflood Mass Flow Rate (RMFR) and Core Damage State (CDS) Steam starved (PARAMETR) Q-11 (Q-L2)

FZK/IRS-AS W. Hering, Ch. Homann5 Forschungszentrum Karlsruhe in der Helmholtz-Gemeinschaft 11th International QUENCH Workshop, Karlsruhe, Germany Step-by-step approach to QUENCH-11  Test objectives -Extend QUENCH facility to low mass flow rate scenarios (ceasing pumps or AMM) -Prepare facility for experiments with free water surface -Investigate scenario with low steam availability -Investigate scenario with low steam availability (app. 1g/s  0.04 g/rod*s)  Stepwise approach - Component tests q11v1 - Component tests q11v1 - Guidance and control test q11v2 (T < 600 K)   qualification of input decks - Design basis reflood test q11v3 (T < 1400 K)   extend database also for DBA codes  update of input decks  QUENCH-11 (Q-L2)

FZK/IRS-AS W. Hering, Ch. Homann6 Forschungszentrum Karlsruhe in der Helmholtz-Gemeinschaft 11th International QUENCH Workshop, Karlsruhe, Germany Best simulation of reactor conditions with Q-L2  Reactor  Consider real volumes in the RPV: Contribution of downcomer: additional 80 to 120 % of the free core flow area Pre-test calculationsPre-test calculations – –Pre-test experiments to assess input decks – –Check independent control of: 1. evaporation rate and 2. bundle heat-up

FZK/IRS-AS W. Hering, Ch. Homann7 Forschungszentrum Karlsruhe in der Helmholtz-Gemeinschaft 11th International QUENCH Workshop, Karlsruhe, Germany Tools for pre-test calculations and post-test analyses  SCDAP/RELAP5 mod 3.2.irs: specially modified for out-of-pile facilities  basic tool  ASTEC V1.x (contribution to SARNET):  Check more possible test scenarios (after qualification using S/R5)  Parameter studies (fast running code)  Due to manpower restrictions: Code validation focussed on DIVA (~ICARE2)

FZK/IRS-AS W. Hering, Ch. Homann8 Forschungszentrum Karlsruhe in der Helmholtz-Gemeinschaft 11th International QUENCH Workshop, Karlsruhe, Germany Pre-test Q-11v2 Objectives:   Steam flow control with Auxiliary heater power   Control bundle heat-up   Response time of additional water inflow   Qualification of fluid measurement   Test low mass flow-rate reflood (< 0.7g/s*rod)

FZK/IRS-AS W. Hering, Ch. Homann9 Forschungszentrum Karlsruhe in der Helmholtz-Gemeinschaft 11th International QUENCH Workshop, Karlsruhe, Germany First post test analysis Draft findings:   Water ejected due to flashing even   Bundle voided z> 0.25 m   Temperature rise linearly until bundle power reduced  Not observed in experiment

FZK/IRS-AS W. Hering, Ch. Homann10 Forschungszentrum Karlsruhe in der Helmholtz-Gemeinschaft 11th International QUENCH Workshop, Karlsruhe, Germany Post-test Q-11v2 Draft findings:   Boil-off rate larger: z > 0.5 m, smaller: z < 0.5 m   Bundle temperatures underestimated   Flashing observed (due to initial conditions)  Check initial conditions and heat losses to environment at “low” temperatures

FZK/IRS-AS W. Hering, Ch. Homann11 Forschungszentrum Karlsruhe in der Helmholtz-Gemeinschaft 11th International QUENCH Workshop, Karlsruhe, Germany Whole scenario Pre-test calculation (before Q-11v2): – –Q-11v3 delivers µm oxide layer (reactor specific) – –Q-11v3 reflood phase simulates Accumulator driven core reflood – –Max temperatures: - Q-11v3: < 1350 K - Q-11 ~ 2600 K Next steps: 1. 1.update input deck 2. 2.Check sequence Q-11v3 and Q-11

FZK/IRS-AS W. Hering, Ch. Homann12 Forschungszentrum Karlsruhe in der Helmholtz-Gemeinschaft 11th International QUENCH Workshop, Karlsruhe, Germany   Reasonable simulation of axial temperature profile during heat-up phase   Onset of final transient OK (t< 7000s)   Temperature peak prior / during reflood underestimated (like most of the codes in ISP-45)  ASTEC V1.2 in work Validation: ASTEC results for ISP-45

FZK/IRS-AS W. Hering, Ch. Homann13 Forschungszentrum Karlsruhe in der Helmholtz-Gemeinschaft 11th International QUENCH Workshop, Karlsruhe, Germany ASTEC: draft results for Quench-11   Comparable to S/R5 calculations   Deviations during cool- down are due to lacking reflood model in (ASTEC V1.1)   Much faster than S/R5   Shows Temperature evolution in the core as well as shroud insulation

FZK/IRS-AS W. Hering, Ch. Homann14 Forschungszentrum Karlsruhe in der Helmholtz-Gemeinschaft 11th International QUENCH Workshop, Karlsruhe, Germany Summary and conclusions (1)  Post test analysis identified unexpected deviations in the S/R5 facility model:  not required before because experiments start at higher temperatures  Q-11v2 proved successfully step-by-step approach to prepare Q-11  Free water level and steam mass flow rate could be controlled, although predictions of pre- calculations differ from experiment  Free water level and steam mass flow rate could be controlled, although predictions of pre- calculations differ from experiment  QUENCH facility, originally not designed for a free water level at lower end of the bundle is now able to simulate that feature

FZK/IRS-AS W. Hering, Ch. Homann15 Forschungszentrum Karlsruhe in der Helmholtz-Gemeinschaft 11th International QUENCH Workshop, Karlsruhe, Germany  Q-11v3 – Q11 pre-test calculations initiated using results of Q-11v2 – –Q-11 can be performed as proposed (QWS-10) – –Database will be released to LACOMERA partners – –Post-test analyses with ASTEC will be continued with know-how from S/R5  SFD-Research focused on:   SARNET: Validation of ASTEC V1.x by code to code and code to data   Improvement and extension of the FZK reflood map   Detailed simulation with S/R5 to supply integral codes with reliable boundary conditions for Q-11 Summary and conclusions (2)

FZK/IRS-AS W. Hering, Ch. Homann16 Forschungszentrum Karlsruhe in der Helmholtz-Gemeinschaft 11th International QUENCH Workshop, Karlsruhe, Germany