Loads on RPV Internals in a PWR due to Loss-of-Coolant Accident considering Fluid-Structure Interaction

Similar documents
Improved PWR Simulations by Monte-Carlo Uncertainty Analysis and Bayesian Inference

The Dynamical Loading of the WWER440/V213 Reactor Pressure Vessel Internals during LOCA Accident in Hot and Cold Leg of the Primary Circuit

Department of Engineering and System Science, National Tsing Hua University,

QUALIFICATION OF A CFD CODE FOR REACTOR APPLICATIONS

VVER-1000 Reflooding Scenario Simulation with MELCOR Code in Comparison with MELCOR Simulation

Safety Analyses for Dynamical Events (SADE) SAFIR2018 Interim Seminar Ville Sahlberg

AREVA Fatigue Concept IAEA TM Lab Tour, JULY 8, 2016

ATLAS Facility Description Report

Stratification issues in the primary system. Review of available validation experiments and State-of-the-Art in modelling capabilities.

3D CFD and FEM Evaluations of RPV Stress Intensity Factor during PTS Loading

THERMAL HYDRAULIC REACTOR CORE CALCULATIONS BASED ON COUPLING THE CFD CODE ANSYS CFX WITH THE 3D NEUTRON KINETIC CORE MODEL DYN3D

Stress and fatigue analyses of a PWR reactor core barrel components

Experiences of TRAC-P code at INS/NUPEC

Developments and Applications of TRACE/CFD Model of. Maanshan PWR Pressure Vessel

CANDU Safety #3 - Nuclear Safety Characteristics Dr. V.G. Snell Director Safety & Licensing

Presenters: E.Keim/Dr.R.Trewin (AREVA GmbH) WP6.9 Task leader: Sébastien Blasset (AREVA-G) NUGENIA+ Final Seminar, Helsinki August, 2016

External Pressure... Thermal Expansion in un-restrained pipeline... The critical (buckling) pressure is calculated as follows:

THERMAL HYDRAULIC ANALYSIS IN REACTOR VESSEL INTERNALS CONSIDERING IRRADIATION HEAT

Grid supports design for dual-cooled fuel rods

Title: Development of a multi-physics, multi-scale coupled simulation system for LWR safety analysis

BUOYANCY DRIVEN MIXING STUDIES OF NATURAL CIRCULATION FLOWS AT THE ROCOM FACILITY USING THE ANSYS CFX CODE

DEVELOPMENT OF A COUPLED CODE SYSTEM BASED ON SPACE SAFETY ANALYSIS CODE AND RAST-K THREE-DIMENSIONAL NEUTRONICS CODE

Preventing Thermal Runaways of LENR Reactors. Jacques Ruer sfsnmc

Transactions on Modelling and Simulation vol 9, 1995 WIT Press, ISSN X

THERMAL STRATIFICATION MONITORING OF ANGRA 2 STEAM GENERATOR MAIN FEEDWATER NOZZLES

Investigation of falling control rods in deformed guiding tubes in nuclear reactors using multibody approaches

Evaluating the Safety of Digital Instrumentation and Control Systems in Nuclear Power Plants

Instability Analysis in Peach Bottom NPP Using a Whole Core Thermalhydraulic-Neutronic Model with RELAP5/PARCS v2.7

APPLICATION OF THE COUPLED THREE DIMENSIONAL THERMAL- HYDRAULICS AND NEUTRON KINETICS MODELS TO PWR STEAM LINE BREAK ANALYSIS

NUMERICAL EVALUATION OF SLOSHING EFFECTS IN ELSY INNOVATIVE NUCLEAR REACTOR PRESSURE VESSELS SEISMIC RESPONSE

CFX CODE APPLICATION TO THE FRENCH REACTOR FOR INHERENT BORON DILUTION SAFETY ISSUE

Steady-State and Transient Neutronic and Thermal-hydraulic Analysis of ETDR using the FAST code system

FINITE ELEMENT COUPLED THERMO-MECHANICAL ANALYSIS OF THERMAL STRATIFICATION OF A NPP STEAM GENERATOR INJECTION NOZZLE

Thermal Hydraulic System Codes Performance in Simulating Buoyancy Flow Mixing Experiment in ROCOM Test Facility

RELAP5 to TRACE model conversion for a Pressurized Water Reactor

VVER-1000 Coolant Transient Benchmark - Overview and Status of Phase 2

ANALYSIS OF THE OECD MSLB BENCHMARK WITH THE COUPLED NEUTRONIC AND THERMAL-HYDRAULICS CODE RELAP5/PARCS

Thermal Analysis. with SolidWorks Simulation 2013 SDC. Paul M. Kurowski. Better Textbooks. Lower Prices.

Safety Analysis of Loss of Flow Transients in a Typical Research Reactor by RELAP5/MOD3.3

ANALYSIS OF A REACTOR PRESSURE VESSEL SUBJECTED TO PRESSURIZED THERMAL SHOCKS

Comparison of Silicon Carbide and Zircaloy4 Cladding during LBLOCA

A Repeated Dynamic Impact Analysis for 7x7 Spacer Grids by using ABAQUS/ Standard and Explicit

Applied Nuclear Science Educational, Training & Simulation Systems

SUB-CHAPTER D.1. SUMMARY DESCRIPTION

Nonlinear Modeling for Health Care Applications Ashutosh Srivastava Marc Horner, Ph.D. ANSYS, Inc.

Natural Frequencies Behavior of Pipeline System during LOCA in Nuclear Power Plants

Introduction to Continuous Systems. Continuous Systems. Strings, Torsional Rods and Beams.

Analysis of Shear Lag Effect of Box Beam under Dead Load

Sebastian Buchholz, Daniel von der Cron, Andreas Schaffrath. System codes improvements for modelling passive safety systems and their validation

«CALCULATION OF ISOTOPE BURN-UP AND CHANGE IN EFFICIENCY OF ABSORBING ELEMENTS OF WWER-1000 CONTROL AND PROTECTION SYSTEM DURING BURN-UP».

The moderator temperature coefficient MTC is defined as the change in reactivity per degree change in moderator temperature.

English text only NUCLEAR ENERGY AGENCY COMMITTEE ON THE SAFETY OF NUCLEAR INSTALLATIONS

Authors : Eric CHOJNACKI IRSN/DPAM/SEMIC Jean-Pierre BENOIT IRSN/DSR/ST3C. IRSN : Institut de Radioprotection et de Sûreté Nucléaire

Sensitivity Analyses of the Peach Bottom Turbine Trip 2 Experiment

Mechanical Engineering Ph.D. Preliminary Qualifying Examination Solid Mechanics February 25, 2002

CFD STUDIES IN THE PREDICTION OF THERMAL STRIPING IN AN LMFBR

Mechanical Design in Optical Engineering

Thermal Analysis with SOLIDWORKS Simulation 2015 and Flow Simulation 2015

Subsystem Response Review

ANALYSIS OF THE OECD PEACH BOTTOM TURBINE TRIP 2 TRANSIENT BENCHMARK WITH THE COUPLED NEUTRONIC AND THERMAL-HYDRAULICS CODE TRAC-M/PARCS

Presented at the COMSOL Conference 2010 Paris

Uncertainty Quantification of EBR-II Loss of Heat Sink Simulations with SAS4A/SASSYS-1 and DAKOTA

Scaling Analysis as a part of Verification and Validation of Computational Fluid Dynamics and Thermal-Hydraulics software in Nuclear Industry

Simplified Method for Mechanical Analysis of Safety Class 1 Piping

Name: 10/21/2014. NE 161 Midterm. Multiple choice 1 to 10 are 2 pts each; then long problems 1 through 4 are 20 points each.

A Numerical Estimate of Flexible Short-Tube Flow and Deformation with R-134a and R-410a

EUROPEAN COMMISSION 5th EURATOM FRAMEWORK PROGRAMME KEY ACTION : NUCLEAR FISSION

EUROPEAN COMMISSION 5th EURATOM FRAMEWORK PROGRAMME KEY ACTION : NUCLEAR FISSION FLOMIX-R FIKS-CT

Buckling, Postbuckling, and Collapse Analysis with Abaqus. Abaqus 2017

Sensitivity Analysis of a Nuclear Reactor System Finite Element Model

Reactivity Coefficients

Severe accident risk assessment for Nuclear. Power Plants

Composite FEM Lab-work

Project ALLEGRO: ÚJV Group Activities in He-related Technologies

Lectures on Applied Reactor Technology and Nuclear Power Safety. Lecture No 6

Extension of the Simulation Capabilities of the 1D System Code ATHLET by Coupling with the 3D CFD Software Package ANSYS CFX

NPP Simulators for Education Workshop - Passive PWR Models

1. Introduction. 1.1 Overview of Study:

COUPLED USE OF FEA AND EMA FOR THE INVESTIGATION OF DYNAMIC BEHAVIOUR OF AN INJECTION PUMP

Thermal Analysis with SOLIDWORKS Simulation 2016 and Flow Simulation 2016

Chapter 8. Design of Pressurizer and Plant Control

Evaluating the Core Damage Frequency of a TRIGA Research Reactor Using Risk Assessment Tool Software

A COUPLED CFD-FEM ANALYSIS ON THE SAFETY INJECTION PIPING SUBJECTED TO THERMAL STRATIFICATION

COPPER FOR BUSBARS CHAPTER 4: SHORT-CIRCUIT EFFECTS

Finite Element Analysis Lecture 1. Dr./ Ahmed Nagib

FIELD TEST OF WATER-STEAM SEPARATORS FOR THE DSG PROCESS

Influence of residual stresses in the structural behavior of. tubular columns and arches. Nuno Rocha Cima Gomes

On Nonlinear Buckling and Collapse Analysis using Riks Method

Quantitative Phenomena Identification and Ranking Table (QPIRT) for Bayesian Uncertainty Quantification

CONVERSION OF THE THERMAL HYDRAULICS COMPONENTS OF ALMARAZ NPP MODEL FROM RELAP5 INTO TRAC-M

Shree Tirupati Polypack

Downloaded from Downloaded from / 1

HWR Moderator Sub-cooling Requirements to Demonstrate Back-up Capabilities of Moderator During Accidents

OF A HELIUM-COOLED MODULE

Structural Analysis of Truss Structures using Stiffness Matrix. Dr. Nasrellah Hassan Ahmed

A Study on Hydraulic Resistance of Porous Media Approach for CANDU-6 Moderator Analysis

GATE SOLUTIONS E N G I N E E R I N G

CFD and Thermal Stress Analysis of Helium-Cooled Divertor Concepts

A Hybrid Approach to Modeling LOCA Frequencies and Break Sizes for the GSI-191 Resolution Effort at Vogtle

TOWARDS A COUPLED SIMULATION OF THERMAL HYDRAULICS AND NEUTRONICS IN A SIMPLIFIED PWR WITH A 3x3 PIN ASSEMBLY

Transcription:

Loads on RPV Internals in a PWR due to Loss-of-Coolant Accident considering Fluid-Structure Interaction Dr. P. Akimov, Dr. M. Hartmann, L. Obereisenbuchner Fluid Dynamics Stuttgart, May 24, 2012

Content Motivation Fluid-structure interaction Introduction of computer codes used STRAC / HAUPT LOCAFLEX Sample LOCA calculations Summary Interaction Dr. P. Akimov May 24, 2012 - p.3

Motivation In the case of a LOCA (e.g. break at the main coolant line): Allowable deformations of Control Rod Guide Tubes must not be exceeded ensure safe reactor shutdown Control Rod Guide Tubes Upper Support Plate (USP) Upper Support Columns Integrity of the RPV internals Upper Plenum (e.g. Core Support Barrel, Support Plates, Support Columns) ensure sufficient core cooling Downcomer Core Support Barrel Core Upper Core Plate (UCP) Lower Support Plate Take into account the main effects of the fluid-structure interaction in the computational models RPV and internals Interaction Dr. P. Akimov May 24, 2012 - p.4

Fluid-structure interaction (FSI) FSI effects with regard to the Core Support Barrel (CSB): FSI effects in the Upper Plenum: Upper Support Plate Cold Nozzle Hot Nozzle Downcomer RPV CSB Core Postulated break buckling of the CSB deflection of the CSB Zone of reduced pressure Strong reduction of incoming pressure waves due to FSI Upper Core Plate bending of the tubular structures in the Upper Plenum Interaction Dr. P. Akimov May 24, 2012 - p.5

Sequential vs. coupled FSI approaches Sequential approach: FSI approach: Fluid Displacements No FSI effects! Fluid Forces Fluid Structure Structure Displacement & Stresses Assessment of the results Fluid Forces Stresses from displacements Covering global loads, not realistic More realistic, load reduction effects due to structural reaction Interaction Dr. P. Akimov May 24, 2012 - p.6

Computer codes Upper plenum internals Code STRAC / HAUPT Upper Plenum Lower internals, RPV, and Fuel Assemblies (FAs) Code LOCAFLEX Core Support Barrel Core with FAs Lower internals RPV and internals Interaction Dr. P. Akimov May 24, 2012 - p.7

Code STRAC / HAUPT (upper plenum internals)

STRAC / HAUPT Main features CSB RPV STRAC is based on code TRAC (Transient Reactor Analysis Code) Loop 1 1D and 3D components PIPE, TEE, VALVE, PUMP, STGEN, VESSEL, CORE Guide Tube for Measurement Loop 1 (hot) Loop 2 (hot) Support Column Control Rod Guide Tube Loop 2 Non-homogeneous, non-equilibrium modelling 2 fluid flow, 2 phase flow Modifications and model extensions Evaluation of forces, valve models, interface to structure programs, e.g., HAUPT Fluid-structure interaction of the upper plenum internals is considered Loop 4 Loop 4 (hot) Loop 3 (hot) 3D grid for the RPV (horizontal cross section) Loop 3 Interaction Dr. P. Akimov May 24, 2012 - p.9

Code LOCAFLEX (RPV, the core, and RPV internals)

LOCAFLEX Main features Fluid: The whole primary coolant volume is included. Modeled are: the RPV with internals, the core, steam generators, pumps, etc 1D, 2 phase flow Structure: The model consists of the RPV, the core, and the RPV internals (except for the upper plenum internals) New object-oriented developments allow for complex structural models generated by ANSYS software with nonlinear couplings Fluid-structure interaction in the downcomer is considered Interaction Dr. P. Akimov May 24, 2012 - p.11

LOCAFLEX Fluid model of the RPV Upper Support Plate Cold Nozzle R327 R345 R363 R381 R9 R144 R8 R143 Loop Level R309 R308 R307 Upper Support Plate R310 R328 R346 R364 Hot Nozzle R7 R142 R306 Downcomer R6 R141 R5 R140 R4 R139 R3 R138 R2 R137 R1 R136 R303 R289 R296 R386 R393 R304 R302 Core Bypass R305 R301 R300 R299 R298 R297 Upper Core Plate Core Lower Support Plate Downcomer Lower Support Plate Core RPV and internals Upper Core Plate = 2D Network Interaction Dr. P. Akimov May 24, 2012 - p.12

LOCAFLEX Structural model (RPV) Rigid beam model of the RPV Coupling to the CSB Flange 26 25 24 23 9 22 21 20 19 27 28 29 14 30 31 32 18 33 13 8 Coupling of the RPV support 12 17 to the building fixpoint 7 6 11 5 4 15 3 2 Coupling to the Lower Support Plate 10 1 Parts of the RPV and internals that are in this model Interaction Dr. P. Akimov May 24, 2012 - p.13

LOCAFLEX Structural model (Lower internals) Extended FE-model with over 2000 MDOFs *) Hold-Down HOLD-DOWN-SPRING Spring USP Hold-Down Spring CRGA-C Control assemblies UCP USP Core HR schroud CRGA-C Control assemblies CSB CB UCP LSP *) MDOFs = Master Degrees of Freedom Parts of the RPV and internals that are in this model Interaction Dr. P. Akimov May 24, 2012 - p.14

LOCAFLEX Structural model (Upper plenum internals) The fluid forces due to the pressure waves (from STRAC/HAUPT) are applied in LOCAFLEX directly to the tubular structures Control Rod Guide Tubes Upper Support Plate USP CRGA-C Control assemblies UCP Fluid forces on the tubular structures Upper Core Plate Realistic coupling mechanism between the UCP and the USP thanks to the detailed FE models of the upper plenum internals RPV and internals Interaction Dr. P. Akimov May 24, 2012 - p.15

Horizontal row models of Fuel Assemblies (FAs) (5 row models for 1 direction are shown with colors) LOCAFLEX Structural model (Core) Loop 1 (hot) Core-Y Loop 2 (hot) Impact couplings Shroud / FA and FA / FA Coupling to UCP Loop 1 Loop 2 Core-X Loop 4 Loop 3 Loop 4 (hot) Loop 3 (hot) Coupling to LSP Interaction Dr. P. Akimov May 24, 2012 - p.16

LOCAFLEX Coupling between fluid & structure models Downcomer Downcomer inside CSB By means of Fluid CSB RPV CSB Control Rod Guide Tubes Upper Support Plate Upper Support Columns Fixpoint CSB Flange LSP UCP RPV By means of Structure RPV FAs FAs Building fixpoint Downcomer Core Support Barrel Lower Support Plate Core Upper Core Plate Core Shroud RPV and internals Interaction Dr. P. Akimov May 24, 2012 - p.17

Sample LOCA calculations

LOCA calculation: Cold leg break Results Scaled force (CSB/CSBmax) 0.5 0.0-0.5-1.0 without FSI X direction Y direction 0.0 0.1 0.2 0.3 0.4 0.5 0.6 TIME (S) RPV Loop 3 (hot) Loop 3 Loop 2 Y Loop 4 (hot) RPV with 4 loops CSB Loop 4 Loop 1 Loop 1 (hot) Deflection of the CSB due to rapid decompression X Total Force on the CSB Interaction Dr. P. Akimov May 24, 2012 - p.19

LOCA calculation: Cold leg break FSI effects without FSI with FSI Scaled force (CSB/CSBmax) 0.5 0.0-0.5-1.0 X direction Y direction Scaled force (CSB/CSBmax) 0.5 0.0-0.5-1.0 X direction Y direction 0.0 0.1 0.2 0.3 0.4 0.5 0.6 TIME (S) 0.0 0.1 0.2 0.3 0.4 0.5 0.6 TIME (S) With FSI more realistic analysis as to peak loadings & frequencies Interaction Dr. P. Akimov May 24, 2012 - p.20

Summary Accurate and reliable numerical tools STRAC / HAUPT and LOCAFLEX for loads analyses in the assessment of safety for nuclear power reactors Allowing for Fluid-Structure Interaction and a complete consideration of dynamics lead to more realistic analyses of loads on the RPV internals New developments in LOCAFLEX allow for more realistic and accurate calculations Interaction Dr. P. Akimov May 24, 2012 - p.21

[Editor and Copyright, 2011]: AREVA NP GmbH Paul-Gossen-Straße 100 91052 Erlangen, Germany. It is forbidden to reproduce the present publication in its entirety or partially in whatever form without prior written consent. Legal action may be taken against any infringer and/or any person breaching the aforementioned conditions. Subject to change and error without notice. Illustrations may differ from the original. The statements and information contained in this brochure are for advertising purpose only and do not constitute an offer of contract. They shall neither be construed as a guarantee of quality or durability, nor as warranties of merchantability or fitness for a particular purpose. These statements, even if they are future-oriented, are based on information that was available to us at the date of publication. Type, quantity and characteristics of goods and services are subject to formal individual formal contracts. Interaction Dr. P. Akimov May 24, 2012 - p.22

Loads on RPV Internals in a PWR due to Loss-of-Coolant Accident considering Fluid-Structure Interaction Dr. P. Akimov, Dr. M. Hartmann, L. Obereisenbuchner Fluid Dynamics Stuttgart, May 24, 2012 Weitere Fragen? Besuchen Sie uns am AREVA-Stand (HS 5) Further questions? Visit us at the AREVA booth (HS 5)