Perspective on Thermal Hydraulic Researches in the Nuclear Field

Contents Introduction Status of the Nuclear Power Domestic & international status Thermal hydraulic researches Active T-H Research Areas Computational tools Development & application Experimental study Facility & instrumentation Concluding Remarks

Introduction Background Objectives Scope & Depth

Status of the Nuclear Power Korean Status Stable Phase Gyeongju disposal facility CP for APR1400 6th Generating Capacity Electricity Generation 39% of total (’06.12) Enhancing the Use of Nuclear Energy Problems

Status of the Nuclear Power (Cont’d) International Status 435 commercial NPPs operating in 30 countries 370,000 Mwe 16% of the world's electricity Enhancing the Use of Nuclear Energy Energy demand, global warming, pollution, etc Problems

Status of T-H Researches Researches on Existing NPPs (to GEN-III+) Focused on safety To resolve safety issues To ensure NPP safety due to power uprates and continued operation for operating NPPs To ensure NPP safety due to new safety features for new NPP designs Researches on Next Generation NPPs Developmental researches to makeup technology gap Safety researches to ensure NPP safety

Status of T-H Researches (Cont’d) NPP Generation GEN-III OPR1000 GEN-III+ APR1400 GEN-IV Water-Cooled Reactor Systems Gas-Cooled Reactor Systems Liquid-Metal-Cooled Reactor Systems Nonclassical Reactor Systems

Status of T-H Researches (Cont’d) T-H Research Categories System code application and development Computer simulation is the only way of knowing how the plant will respond to normal and accidental conditions. Experiment To investigate real phenomena and/or prove system performance To provide technical basis for the development of models and correlations Computational Fluid Dynamics (CFD) code application and development CFD has now been recognized as an important tool. Widely used for single phase problems, but for multi-phase ones

Status of T-H Researches (Cont’d) Relationship among Research Categories System Code Models & Correlations, Validation Scailability, Facility & test design Experiment Models & Correlations, Complement Problem Definition IC & BC Mechanistic models, Validation Facility & test design CFD

System Code Application Wide Use of Realistic System Codes WCOBRA/TRAC, CATHARE, CATHENA, RELAP5/MOD3, TRACE Problems What is the uncertainty? Application of realistic evaluation model (REM) What is the safety margin? Application to safety and license issues  Is a design feasible? Application to test applicability of codes and issue finding Uncertainty evaluation and methodology development for existing reactors and issue finding for new reactors

System Code Application (Cont’d) Activities in Korea (원자력안전해석심포지움, CAMP/MUG Meeting) ECCS realistic evaluation model (KINS, KAERI, KNF, KOPEC) Application of KINS-REM to APR1400

System Code Application (Cont’d) Activities in Korea (Cont’d) Safety issue resolution (KINS, NETEC, KNF, KOPEC) Mixing Volume Calculation (RELAP5) Boron Concentration after LOCA (BORON) Application of RELAP5 to Long-term Cooling Issue

System Code Application (Cont’d) Activities in Korea (Cont’d) Safety margin evaluation (KINS) Event Tree Safety Margin Changes in PCT PDF Application of RELAP5 to Safety Margin Evaluation

System Code Application (Cont’d) Activities in Korea (Cont’d) Application to new reactors (KAERI) MARS Code V/V for LBE (Liquid lead-bismuth eutectic) Properties Application of MARS to KALIMER

System Code Application (Cont’d) Activities in Korea (Cont’d) Application to experiment for new reactors (SNU) SNU RCCS Facility for VHTR Bulk Temperature in the Water Tank Application of MARS to SNU RCCS Experiment

System Code Application (Cont’d) International Activities (International CAMP Meeting) Very similar to Korean activities Code application to evaluate safety issues Code application to check the existing code applicability for new reactors Fuel Cladding Temperatures during Blowdown for a Feedwater Line Break Application of RELAP5 to HPLWR (European Supercritical Reactor

System Code Development Background Accumulation of T-H research results Drastic increase in machine performance Need to efficient NPP operation Problems How to reduce the uncertainty? Develop more sophisticated modeling (code & nodalization) Develop more mechanistic models  How to makeup technology gaps? Improve the existing code capability or develop new codes Develop new physical models Uncertainty reduction for existing reactors and development of a system code for new reactors

System Code Development (Cont’d) Korean Activities for the Existing Reactors MARS code consolidation (KINS) DVI model for APR1400 Validation of 3-D capabilities Code coupling Development of numerical technique with a high precision (KAERI) Multi-dimensional models for 2-Ф flow and interfacial transport phenomenon Numerical scheme Multi-dimensional component modules (RCS, SG, etc) Development of domestic design code system (KHNP) Secure his own property

System Code Development (Cont’d) International Activities for the Existing Reactors TRACE code development (USNRC) Spacer grid models Droplet field & entrainment modeling Interfacial drag models NEPTUNE project (France IRSN) Multi-scale and multi-disciplinary calculations (CATHARE + CFD) To prepare the new generation of industrial two-phase flow codes To structure and coordinate the R&D works around two-phase flow

System Code Development (Cont’d) Analysis Needs for New Reactor Systems Design tools for the reactor and associated components and processes Predictive capabilities for ensuring safety of future reactor designs Evaluation tools for design certification and licensing processes Analysis results must present a strong safety case with a high level of proof for reactor licensing.

System Code Development (Cont’d) Current Status and Challenges A current focus is to modify LWR safety codes. Challenges New fuels Different coolants Higher temperature operating conditions Events relevant to high-temperature gas-cooled reactors differ from those in LWR’s New severe accident models are needed for phenomena peculiar to the future reactors

System Code Development (Cont’d) Activities in Korea for New Reactors Extension of MARS code to new reactors (KAERI) MARS-SMART MARS-GCR MARS-LBE

System Code Development (Cont’d) Activities in the States for New Reactors Gas-cooled reactor codes GRSAC LWR codes ATHENA + FLUENT + CONTAIN MAAP4 MANTRA MELCOR RELAP5-3D Major problem is lack of data for development of models & correlations and code validation

CFD Code Application Background Problems CFD codes commonly used in many industrial sectors are now applied for nuclear reactor safety of for design purposes. CFD codes are mainly capable of obtaining at a lower cost, valuable information on some physical phenomena. Enable to test virtually any configuration, from both qualitative and quantitative points of view Problems How to validate CFD code and make results reliable? How to apply multi-component and multi-Ф problems? Quality assurance for CFD codes, and identification of the weakness of CFD techniques

CFD Code Application (Cont’d) Direct Numerical Simulation (DNS) “Costly” Fluid Dynamics used exceptionally 100% accurate with NO modeling hypothesis Large Eddy Simulation (LES) “Colorful” Fluid Dynamics with great detail Applicable to eng. problems but at some cost Almost as reliable as DNS, know-how required and not well established Reynolds Averaged (RANS) Conventional Fluid Dynamics widely used in eng Economical, and full reactor or sub-component design problems Needs improvement & validation for new range of applications

CFD Code Application (Cont’d) Journal of Fluids Engineering Editorial Policy The Journal of Fluids Engineering will not consider any paper reporting the numerical solution of a fluids engineering problem that fails to address the task of systematic truncation error testing and accuracy estimation. Authors should address the following criteria for assessing numerical uncertainty. The basic features of the method including formal truncation error of individual terms in the governing numerical equations must be described. Methods must be at least second order accurate in space. Inherent or explicit artificial viscosity (or diffusivity) must be assessed and minimized. Grid independence or convergence must be established. When appropriate, iterative convergence must be addressed. In transient calculations, phase error must be assessed and minimized. The accuracy and implementation of boundary and initial conditions must be fully explained. An existing code must be fully cited in easily available references. Benchmark solutions may be used for validation for a specific class of problems. Reliable experimental results may be used to validate a solution.

CFD Code Application (Cont’d) USNRC ACRS Position on CFD Application Anticipate increasingly use of CFD codes To resolve multidimensional effects important to safety Very little effort on CFD work in NRC CFD codes are validated to a much less rigorous standard than codes for nuclear use, and the source codes are not available. CFD codes appear to include a number of ad hoc fixes to improve stability and robustness Comments on NRC plan for CFD model development NPHASE for regulatory applications Feasibility of the plan within the resource constraints of the agency. Standards of reliability and accuracy required for CFD codes

CFD Code Application (Cont’d) 1-Ф Problems for the Existing Reactors Boron dilution Mixing of cold and hot water in Steam Line Break event Hot-leg temperature heterogeneity PTS (pressurised thermal shock) Counter-current flow of hot steam in hot leg for severe accident investigations of a possible “Induced Break” Thermal fatigue (e.g. T-junction) Hydrogen distribution and combustion in containment

CFD Code Application (Cont’d) 1-Φ Problems for New Reactors Natural circulation in LMFBRs Coolability & Flow induced vibration of APWR radial reflector Flow in lower plenum of ABWR Depressurization of a GCR Decay heat removal in a GCR Thermal loading on structures Containment integrity (peak pressure) during the long-term cooling of innovative reactors with passive safety systems

CFD Code Application (Cont’d) 2-Φ Problems Where CFD Is Recommended 2-Φ CFD tools are far less mature than single phase tools. Identification of the relevant basic phenomena which need to be modeled for a given application Assessment of the model including verification, validation and demonstration tests Definition of new R&D work for a more detailed validation, for a better numerical efficiency, and a better accuracy and reliability of predictions

CFD Code Application (Cont’d) CFD Application to Boron Dilution Issue

CFD Code Application (Cont’d) CFD Application to 2-Φ Problems

CFD Code Development Immature Field with Many Challenges Multi-component and multi-phase flow Turbulent models Numerical techniques Verification & validation Highlighted by technology development, enhanced machine performance, and relative low costs compared to experiments

CFD Code Development (Cont’d) New T-H Many Challenges for VHTR

CFD Code Development (Cont’d) INL 10-Year Plan for CFD

CFD Code Development (Cont’d) Limited Scope CFD Analysis

Experimental Study Mostly Accompanied by Code Development and Validation Code development Experiments to capture physics Experiments to develop models & correlation Component level experiments to develop component models in a code Code validation System or component level experiments (SETs & IETs) Proof Experiments with Real Scale Facility

원자력 열수력 실험의 국내 기관별 역할 열수력 실증실험 기술 개발(계측, 고온고압 실험, 척도해석 등) 원자력 전력 연구원 학 계 산업계 안전 기술원 원자력 연구소 열수력 실증실험 기술 개발(계측, 고온고압 실험, 척도해석 등) 중대형 열수력 실증실험체계 구축 및 실험 수행 중장기적 열수력 실증실험 연구 방향 도출 열수력 실증실험 기반 기술 개발 중소형 열수력 실험체계 구축 및 실험 수행 산업체 응용 실험 수행 실증실험 결과 활용 성능/안전성 향상 및 코드 검증을 위한 실증실험 필요 분야 도출 안전규제 현안/관심사 관련 실증실험 필요 분야 도출

주요 대학의 핵심 연구 분야 KAIST 서울대 포항공대 경희대 제주대 기 타 CHF, Post-CHF 열전달 및 관련 메커니즘 비등 및 응축 열전달 신형원자로 계통 및 기기의 열수력 특성 (DVI 관련 현상, 피동 노심 냉각, 피동 격납 용기 냉각, 가스냉각로 등) 서울대 2상유동 및 Subcooled Boiling 특성 규명 (계측기 개발 포함) 신형원자로 계통 및 기기의 열수력 특성 (DVI 관련 현상, 피동 피동 격납 용기 냉각, 가스냉각로 등) 소형 열수력 종합효과실험 포항공대 비등 및 응축 열전달 특성 (다양한 열전달 증진 메커니즘 연구 포함) 계측 기술 개발: 전자기 유량계, Heat Flux Meter 등 비원자력 분야 산업체 응용 연구 경희대 Thermal Mixing & Stratification Fluid-Structure Interaction 제주대 Electrical Impedance Imaging 기 타 Fundamental Research in Various Topics

국내 실험 기술 및 인프라 수준 개별효과실험 체계 종합효과실험 체계 고온/고압 실험 및 측정기술은 세계적 수준에 접근 원자력 중장기 연구개발사업 등을 통해 열수력 핵심 모델 관련 실험시설을 확보하고 연구 진행 중 코드 모델 평가를 위한 상세 측정 데이터 생산 본격화 국내 개발 원자로 특성을 반영한 고유 모델 일부 개발 국내 실험에 기반한 고유 모델 개발 성과는 크게 미흡 종합효과실험 체계 ATLAS 구축으로 국내 원전 산업/규제에 필수적인 최소한의 종합효과실험체계 확보 국외 실험 데이터만으로는 미흡한 APR1400 및 OPR1000 관련 대부분의 사고 시나리오 모의 가능  산업체 및 규제검증 코드의 경쟁력에 필수 SMART 등 설계 특성이 크게 다른 원자로의 경우 새로운 종합효과실험시설 필요

Experimental Study for Advanced Reactor Systems APEX Facility Advanced Plant Experiment Complete 2x4 Loop Primary System: Length scale 1/4 Volume Scale 1/192 Core Power ~ 1 MW Passive Safety Systems: 2 Core Makeup Tanks (CMTs) , 2 Accumulators, a Passive Residual Heat Removal (PRHR) Heat Exchanger, IRWST, and a 4-Stage ADS System. Hot & Cold Leg SBLOCAs, MSLB, Inadvertent ADS, Double-Ended DVI Line Break, Station Blackout and Long Term Recirculation.

Experimental Study for Advanced Reactors Systems (Cont’d) PUMA Facility Purdue University Multi-Dimensional Integral Test Assembly Design a well-scaled integral test facility having proper and sufficient instrumentation Pressure ratio 1/1 Area ratio 1/100 Length (height) ratio 1/4 Power ratio 1/200 Maximum heater rods power 650 kW Assess important SBWR phenomena for LOCA and other transients

Experimental Study for New Reactor Systems Experiments for CFD Model Assessment

Experimental Study for New Reactor Systems (Cont’d) Experiments for CFD Model Assessment (Cont’d)

Concluding Remarks Importance of T-H Researches Lots of T-H Researches Safety issue resolution for the operating NPPs Advanced and future reactor systems Increased needs for CFD technology development Keep your eyes on TREND Integrate information gathered and extract points Not predicting the future would be like driving a car without looking through the windshield.