Nuclear Thermal-Hydraulic Safety for the Future I: The Present and Future of Nuclear Thermal-
Hydraulic Safety Research
Nov. 19, 2018
Ki-Yong CHOI, Ph.D. ([email protected])Director
Thermal-Hydraulics & Severe Accident Research Division
Korea Atomic Energy Research Institute
Plenary Lecture 3
Thermal Hydraulics & Severe Accident Research Division
Contents
Safety Research of KoreaI
TH Safety Research OutcomesII
Opportunities and ChallengesIV
ObservationsIII
2
SummaryV
Thermal Hydraulics & Severe Accident Research Division
Contents
Safety Research of KoreaI
TH Safety Research OutcomesII
Opportunities and ChallengesIV
ObservationsIII
3
SummaryV
Thermal Hydraulics & Severe Accident Research Division
General Information of KAERI
4
Nuclear R&D Ecology in Korea
Safety Research in Korea
MSIT-Driven Safety Technology R&D- Basic, Common, Cross-cutting-
(KAERI, Academia, etc)
RegulatoryApplication
IndustrialApplication
Safety Goal AssuranceEconomic Competitiveness
Public Acceptance
NSSC-Driven Regulatory R&D(KINS/KAERI/Academia, etc)
MOTIE-Driven Application R&D(Industry/KAERI/Academia, etc)
Methodology, Tool, Tech. Background
Methodology, Tool, Data, new concepts
License
RegulatoryNeed
FutureNeed
IndustryNeed
Thermal Hydraulics & Severe Accident Research Division
General Information of KAERI
5
Founded in 1959, KAERI has been a cradle for Korean nuclear Industry and other organizations including regulatory bodies
Personnel : 1463 employees (865 Doctors, 598 Masters and Bachelors)Budget(2017) : 615 Million USDDae-jeon(H/Q), Jeong-eup(ARTI), Gyeong-ju(KOMAC), Busan(KIRR, UC)
KIRAMS(Korea Institute of
Radiological & Medical Sciences)
KOPEC(Korea Power
Engineering Co., Inc.)
KNFC(Korea Nuclear Fuel
Company)
KINS(Korea Institute of
Nuclear Safety)
KINAC(Korea Institute of Nuclear
Nonproliferation and Control)
Technology Transfer•NSSS Design (KOPEC)•Fuel Design(KNFC)•Rad-waste Management (KHNP/NETEC)
59.2 KAERI
07.396.12
1960s 1970s 1980s 1990s 2000s
75.10 82.11 90.2 04.12
Safety Research in Korea
Thermal Hydraulics & Severe Accident Research Division
Research Branches of KAERI
6
Neutron Science R&D
(HANARO)
Frontier for Nuclear
Technology Industrialization
KAERI (Daejeon)
Proton Accelerator(100MeV)
Nuclear and Material Research
KOMAC (Gyeongju)
2nd Research Reactor for RI
Production (under construction)
KJRR (Busan) Radiation Fusion Tech.,
Materials, Biological Application
ARTI (Jeongeup)
KAERI
Safety Research in Korea
Thermal Hydraulics & Severe Accident Research Division
Nuclear Safety R&D of KAERI
7
PhysicalIntegrity
Reactor Cooling(Thermal-hydraulics)
RiskManagement
EnvironmentalProtection
Severe AccidentManagement
TH integral effect tests for accident conditions of APR1400, OPR1000 & APR+ at prototypic press & temp
Steam explosion tests with prototypic reactor core materials of up to 30 kg
Test of corium behavior in a reactor cavity for up to 300
kg of prototypic corium
Various test facilities for new design features of APR1400, APR+ & SMART
ATLAS TROI VESTA
ATLAS : Advanced Thermal-hydraulic Test Loop for Accident Simulation
TROI : Test for Real corium Interaction VESTA : Verification of Ex-vessel corium
stabilization
Safety Research in Korea
Thermal Hydraulics & Severe Accident Research Division
Contents
Safety Research of KoreaI
TH Safety Research OutcomesII
Challenges and OpportunityIV
ObservationsIII
8
SummaryV
Thermal Hydraulics & Severe Accident Research Division
TH Safety Research Outcomes
9
APR1400 NRC-DC SUPPORTRAI Responses- VAPER, MIDAS & ATLAS, …- Electrical 4-train tests- License approval in 2018
APR+, IPOWER DEVELOPMENTImproved design features- DVI+, FD+, PAFS, PCCS- HEMS, PECCS, A/W mixed cooling, …
CODE DEVELOPMENTTH system code + BEPU- MARS-KS, SPACE- PAPIRUS, SimulatorMulti-scale code- CUPID, CINEMA
GEN.IV DEVELOPMENTSMART, SFR, sCO2 system- VISTA, SMART-ITL, FESTA, FINCLS, SISTA- iHELPS, PRESCO- SCIEL
APR1400 DEVELOPMENTComprehensive R&D- IET: ATLAS, RCP, FD- SET: THETA, B&C, MIDAS
BASIC EXPS. & MODELINGPhenomena understanding and modeling- Boiling, condensation, CHF, convection, NC, etc- Multi-physics (Fuel + TH + CTMT)
Break
Flow
LBLOCA
DVI Nozzle
Thermal Hydraulics & Severe Accident Research Division
Basic Experiments and Modelling
10
CHF Tests for License Support and Phenomena Identification
R&D Contents Phenomena identification of CHF and Post-CHF for fuel safety and performance
improvement
Major Outcomes License support of 3-pin Fuel Test Loop for HANARO (water CHF)
7-rod water CHF, 5x5 bundle Freon CHF tests for SMART SDA support
License support of JRTR export
Technology transfer to NFI by performing Freon CHF tests
Technology transfer to KHNP-CRI, KINS for validation of SPACE, MARS codes
TH Safety Research Outcomes
Thermal Hydraulics & Severe Accident Research Division
Basic Experiments and Modelling
11
Safety Relevant Local Thermal-Hydraulic Basic Tests
R&D Contents Basic & high heat flux boiling phenomena study
ECC bypass tests and model development
PAFS pool boiling and local condensation in HX
Major Outcomes Experimental database and
Physical model improvement of the codes
TH Safety Research Outcomes
High heat flux boiling structure (left)Boiling visualization (right)
PAFS pool boiling (left) Condensation model of the HX (right)
ECC water film behavior and model
Thermal Hydraulics & Severe Accident Research Division
Basic Experiments and Modelling
12
3-D Rod Bundle Turbulence Behavior: NEA Benchmark (IBE-2)
R&D Contents 5x5 rod bundle single-phase turbulence mixing test (MATiS-H)
4x4 rod bundle two-phase flow behavior test (MATiS-V)
International cooperation for CMFD validation
Major Outcomes Experimental DB for local turbulence structure by LDV, PIV
IBE-2; 12 countries, 22 organizations, 25 groups; standard DB for CFD validation
Contribution to special session of the CFD4NRS4
TH Safety Research Outcomes
MATiS-H MATiS-V
X(mm
)
-10
0
10
Y(mm
)
-10
0
10
Vo
idF
ractio
n[- ]0
0.2
X Y
Z
0.3
0.26
0.22
0.18
0.14
0.1
0.06
0.02
Void fraction at sub-channel Velocity, TI profiles
Thermal Hydraulics & Severe Accident Research Division
Basic Experiments and Modelling
13
3-D Two-Phase Mixing Tests
R&D Contents 3-D two-phase behavior taking place in the DC during LOCA
Major Outcomes 3-D void distribution in large plate facility (DYNAS)
Validation and improvement of system-analysis codes (SPACE, MARS, etc)
Cooperation with CEA by data exchange (DYNAS vs. REGARD)
TH Safety Research Outcomes
DYNAS facility DYNAS flow visualization and void fraction test results
Thermal Hydraulics & Severe Accident Research Division
Basic Experiments and Modelling
14
3-D Code Validation Technology Development
R&D Contents Effects of spacer grid on reflood heat transfer
Coolability assessment of the deformed fuel during LB-, IB-, SB-LOCAs
Major Outcomes Reflood heat transfer DB with 6x6 rod bundle
Investigation of the effects of relocated fuel on coolability
Verification of ECCS performance under ballooned fuel conditions
Validation DB against the ECCS rule revision (10CFR50.46c)
TH Safety Research Outcomes
FR2 test results (KfK, 1983)
Thermal Hydraulics & Severe Accident Research Division
Comprehensive Thermal-Hydraulic R&D for the APR1400 Development and Licensing Support (’97~’06)
THETA : SET ProgramATLAS : IET ProgramMARS : Safety AnalysisSevere Accident Mitigation Measures
15
DB
Model
DB
APR1400 Develop. & NRC DC Support TH Safety Research Outcomes
Thermal Hydraulics & Severe Accident Research Division
RCP Shaft Seal Assembly (SSA) Performance TestR&D Contents RCP SSA characterization during SBO conditions with a full-scale test loop
Major Outcomes Generation of seal leakage characterization data for SBO analysis
Reduction of safety analysis uncertainty associated with RCP seal leakage
APR1400 Develop. & NRC DC Support TH Safety Research Outcomes
Thermal Hydraulics & Severe Accident Research Division
R&D Contents In case of 4-train EDG design, the ECC bypass rate
was found to increase
How to avoid an increase of ECC bypass rate?
Major Outcomes DVI+ concept proposed which found to reduce
the ECC bypass rate
The performance of DVI+ was experimentally confirmed by 1/5 air-water test and more margin for LBLOCA late reflood core heating was achieved
17
APR+, iPOWER Development TH Safety Research Outcomes
1/5 scale ECC Bypass Air-Water Test
ECC Duct (DVI+)
Optimization of Safety System for APR+
ECC Bypass Fraction
CL-1CL-2
CL-3
DVI-4
Break
DVI-2
DVI-3
DVI-1
DVI-4
Break
DVI-3
DVI-1
CL-1
CL-3
DVI-4
DVI-1
Break
CL-2
4-EDG designDVI combination
Thermal Hydraulics & Severe Accident Research Division
R&D Contents Can the SIT injection period be extended to get
more safety margin?
Injection of non-condensable gas should be avoided to prevent adverse effects on core cooling
Major Outcomes Noble idea to use the dead volume of the lower
hemisphere of the SIT was suggested to extend injection time
Design change to prevent the non-condensable gas from getting out of the SIT was suggested
18
APR+, iPOWER Development TH Safety Research Outcomes
Optimization of Safety System for APR+
잔여수위
잔여수위
SIT lower dead volume
SIT SIT+
Gravity driven SIT to prevent nitrogen gas from released
Thermal Hydraulics & Severe Accident Research Division
R&D Contents Performance verification and design optimization of the PAFS
Major Outcomes Separate effect tests with a single-HX-installed facility PASCAL showed reliable
heat transfer performance enough to remove the decay core power
The following integral effect tests with three-HX-installed ATLAS-PAFS confirmed safety requirements required for SDA of APR+
19
APR+, iPOWER Development TH Safety Research Outcomes
Supporting Licensing of SDA of APR+
APR+ PAFS conceptual drawing
Steam
Generator
PCHX
PCCT
PCCT
Pre-Heater
PCCT
Circulation Pump
Nitrogen Gas
Pre-Heater
Steam
flow
Condensate
flow
PASCAL test results ATLAS-PAFS facility
Thermal Hydraulics & Severe Accident Research Division
R&D Contents Development of air or air-water hybrid cooling concepts
sustainable for 72 hours to deal with SBO
Major Outcomes Air-water hybrid cooling concept
PCCT water pool empty time can be extended up to 72 hours by air-water cooling concept
Direct SG secondary side air cooling concept
The cooling time can be even extended infinitely when air cooling concept is applied
20
APR+, iPOWER Development
Development of Innovative Safety Features Direct SG Air-cooling concept
Air-water hybridCooling concept
Containment air cooling concept
72 hour cooling during SBO
TH Safety Research Outcomes
Thermal Hydraulics & Severe Accident Research Division
R&D Contents Development of hybrid pressure emergence makeup system (HEMS) for APR+
Development of passive ECCS (PECCS) technology for iPOWER
Major Outcomes Separate and integral effect tests for performance verification
Development of a best-optimized guideline for design and operation
21
APR+, iPOWER Development
Development of Innovative Safety Features
TH Safety Research Outcomes
HEMS PECCS
Thermal Hydraulics & Severe Accident Research Division
SMART Development
22
Integral Effect Tests for SDA, PS Validation, SMART-PPE
R&D Contents SDA support tests with VISTA-ITL for SMART (1:1 in height, 1/310 in volume)
SER (Safety Enhancement Research) and PPE (Pre-Project Engineering) with SMART-ITL (1:1 in height, 1/49 in volume) for confirmation of design
Major Outcomes VISTA-ITL: SBLOCA (5 runs), CLOF (1 run), PRHRS (4 runs) used for fluid system
design & safety analysis and answers to RAI (4 times, 64 items) helped to achieve SDA (2012.7.4)
SMART-ITL: VISTA-ITL CTs (6 runs), PSS 1/2 train (14/5 runs), DBAs (14 runs), performance (8 runs)
TH Safety Research Outcomes
VISTA-ITL
0 200 400 600 800 1000 1200 1400 1600 1800 20000.0
0.1
0.3
0.4
0.5
0.6
0.8
0.9
1.0
No
rma
lize
d W
ate
r L
eve
l (-
)
Time from Break (seconds)
RPV Level (LT-BPV-01)
SB-SIS-07
SB-SCS-04
SB-PSV-02
Active Core Region
RPV w. level dur. LOCA SMART-ITL SMART passive systems
Thermal Hydraulics & Severe Accident Research Division
□PGSFR Flow Distribution Facilities (PRESCO) Euler No. scaling, 1/5 length reduced and preserved internal geometry
Core/IHX simulators for flow measurement and DP control & calibration
□On Going Subject (2018) Local multi-D flow visualization.
Gas entrainment- on IHX inlet
0.57070.5451
0.3697
0.47400.3052
0.5211 0.5300 0.4641 0.51350.54000.58470.57300.3392
0.2490
0.3180
0.3870
0.4560
0.5250
0.5940
Mass Flow Rate [kg/s]
0.3411
0.3912 0.4474 0.5614 0.5468 0.5734 0.5492 0.5217 0.3872 0.3136
0.3111 0.3775 0.4703 0.4765 0.4730 0.3730 0.3290
0.2958 0.3036 0.3032 0.3242 0.2821
0.47310.3132 0.3891 0.4640 0.5417 0.5114 0.5357 0.5708 0.3879 0.3353
0.37560.3136 0.3809 0.4712 0.4139 0.4710 0.3070
0.2795 0.3204 0.3160 0.3044 0.2763
0.51820.2917 0.3740 0.4788 0.5237 0.5691 0.5933 0.4657 0.2739
0.37050.54510.2818 0.3747 0.4733 0.58410.5327 0.5682 0.4863 0.2710
0.51870.4731 0.30950.47680.3073 0.4147 0.52260.5291 0.5389 0.4031
0.37280.3386 0.34830.2492 0.37450.3992 0.3912 0.2805
0.51430.4944 0.32310.47850.3034 0.4009 0.52090.5523 0.5562 0.3886
0.38680.3412 0.34070.2763 0.37060.3977 0.3910 0.2730
0 1 2 3 4 5 6 7 8 9 100.20
0.25
0.30
0.35
0.40
0.45
0.50
0.55
0.60
0.65
Flo
w R
ate
(kg/s
)
Group
EXP.
CFD
Target
Facilities (PRESCO)
Core Flow (Exp) Groupwise Core Flow
①
①
②
②
③
Facilities
(GETS)
Fuel Assembly Simulator
Venturi Tube(for Flow
Rate Meas.)
VRROS(for
DPControl)
Inlet Plenum
Core Shroud
UISRedan
□Highly Accurate Exp. DB Acquired Uncertainties & mass balance error ≤ 1%
Exp, Calc: No Bypass
Valve network for 300 Dp’s
DP lines assembling
0 2 4 6 8 10 12 14 16 18 20 22 24-20
0
20
40
60
80
100
120
140
160 EXP.
CFD
Pre
ssure
(kP
a)
Location
Pressures around 1 IHX
Pressures along
The flow path
Gen. IV Development TH Safety Research Outcomes
Thermal Hydraulics & Severe Accident Research Division
Code Development
24
MARS & MARS-SIM Development and Maintenance
R&D Contents Improvement of MARS-KS and development of MARS-SIM
Major Outcomes Transfer of MARS-KS to KINS and keep it updated
Development of multi-D post-processor
Full-scope simulator of Shin-Hanul 1/2, JRTR, HANARO
Technology transfer to WSC and KHNP-CRI
PC-based OPR1000 simulator used by IAEA for training
TH Safety Research Outcomes
Research reactor model
Multi-D post-processor
Full-scope simulator, MARS-SIM
Thermal Hydraulics & Severe Accident Research Division
Code Development
25
SPACE Development, License Approval and Even More
R&D Contents B.E., multi-D, system-scale code with 3-fields formulation
Multi-purpose: LOCA & Non-LOCA safety analysis;
Licensing approved by NSSC in 2017
TH Safety Research Outcomes
Code RequirementModel Developemnt- Hydro Solver/Model & Corr.- Special Process/ComponentModel integration & verificationDemo versionExperiments
SPACE Validation (1D)- Separate effect- Component effect- Integral effect- Plant dataSPACE Environment- GUIIndependent V&V by 3rd partyValidation Experiments
Answers for KINS questions- Prepare answers- Error improvementSPACE Validation (3D)- Experiment / Plant dateCode maintenance
Validation/Preparing for Licensing
Licensing
March 2007 March 2010 Dec. 2012 Dec. 2015
Code Development
Dec. 2018
Multi-D ModelingExtension to Gen-IV Reactor- SPACE-SFR- SPACE-RRExtension to DECs capability- High burn-up- Fuel coupling
Application
Licensed on March 3, 2017
Thermal Hydraulics & Severe Accident Research Division
Code Development
Parallel computing Platform IntegRated for Uncertainty and Sensitivity Analysis (PAPIRUS)
PAPIRUS performs:- Sensitivity analysis- Linearity test- Data assimilation- Development of surrogate models/reduced
order models- Uncertainty propagation- Quantification of code accuracy using Fast
Fourier Transform Based Method (FFTBM)- Optimization of reactor design, using e.g.,
simulated annealing
26
1.List parameters and define mean and uncertainty 2.Select parameters to
perturb in code input
3.Parallel calculation
4.Monitoring the process
5.Examine the graphical results
TH Safety Research Outcomes
Thermal Hydraulics & Severe Accident Research Division
Code Development
UQ with Scaling Analysis (Data Assimilation Module of PAPIRUS)
1.Calibration of parameters, e.g. physical models, boundary conditions, etc., using 1D small scale FEBA test data.
2.Uncertainty quantification of cladding temperatures for larger scale FLETCH-SEASET and 2D PERICLES tests using the parameter distributions obtained by step 1 (blind calculation).
27
Physical Property Space
FEBA
PERICLES
FLETCH-SEASET
…...
FEBA test data
A PosterioriParameter
Distributions
FEBA
FLETCH-SEASET
PERICLES
Nuclear Power Plants
A Priori Parameter Distributions
TH Safety Research Outcomes
Thermal Hydraulics & Severe Accident Research Division
Code Development
28
TH Safety Research Outcomes
The CUPID code has been developed for steady-state and transient analyses of single- and two-phase flows in nuclear reactor in component- or CFD-scale
Scale (m)
… 10-3 10-2 10-1 100
Component-scale System-scale
MARS,SPACE
Application to passive system
Component Analysis
DNS
CUPID/MARS
Multi-scale TH
액체온도
CUPID
MARSModeled
by MARS
Modeled
by CUPID
Setup Numerical Method
• 3D 2-fluid 3-field• Implicit scheme• Unstructured FVM• Verifications
2012~2016Development
of CUPID
2007~2011High-resolution
Numerical Method
Pin wise kinetics code coupling
CUPID/DeCART
CUPIDDeCART
fuel q’’’
ρl , Tl , αl , Tfuel
Meso-scale
2017~20213D Safety
Analysis using CUPID
PAFS AnalysisCUPID
-SG
High-fidelity Containment
Analysis
3D TH Analysis for Reactor Core
Coupling of TH and Fuel Structure
Codes
• 3D TH and 2D Fuel Structure Codes
• Coupled simulation of fuel deformation
Thermal Hydraulics & Severe Accident Research Division
Code Development
29
TH Safety Research Outcomes
CUPID Code Development
Major Outcomes Reasonable ranking in IBE-4 benchmark with
GEMIX experiments of PSI (2016)
Ranked first in the boron dilution benchmark with ROCOM data(2017)
CUPID
Ranking for turbulent kinetic energy
* Cho et al., Nuclear Engineering and Design, 2018.
Thermal Hydraulics & Severe Accident Research Division
Code Development
30
TH Safety Research Outcomes
CUPID Code Development
Major Outcomes Successful replication of
the siphon break line to prevent core uncovery
Interface drag model based on inter-phase topology map
Outlet: atmosphere (LOCA)
Atmosphere Pressure
Maindrainpipe
Siphon break line
Water Tank
Rx
* I.K.Park et al., Nuclear Engineering and Design, 2018.
Thermal Hydraulics & Severe Accident Research Division
CUPID Development
System TH Code Coupling (CUPID-MARS)
31
MARS
CUPID 2-phasecoupling
1-phasecoupling
• Heat Structure Coupling (Explicit): APR+ PAFS simulation
• Flow Field Coupling (Implicit): Pressure Matrices are Merged
Verification of implicit coupling
* Park et al., ANE, 2013.
CUPID
Neutronics Code Coupling (CUPID-MASTER/DeCART)
CUPIDMASTER
orDeCART
fuel q’’’
ρl , Tl , αl , Tfuel
CEA Drop Accident Analysis(CUPID/MASTER)
Rod wise 1/4 core steady state calculation (CUPID/DeCART)
TH Safety Research Outcomes
Thermal Hydraulics & Severe Accident Research Division
CUPID Development
CUPID-CT is now being validated against HYMERES-2(H2P1-0) and THAI (HM-2) tests
Simulation of HYMERES-2 H2P1-0 case
32
TH Safety Research Outcomes
simulation
test data
Helium layer erosion transient
Thermal Hydraulics & Severe Accident Research Division
Contents
Safety Research of KoreaI
TH Safety Research OutcomesII
Opportunities and ChallengesIV
ObservationsIII
33
SummaryV
Thermal Hydraulics & Severe Accident Research Division
Observations from R&D Outcomes (1/2)
34
Development of advanced LWRs was the main initiative of most thermal-hydraulic safety researches in Korea in the past decades
OPR1000 APR1400 APR+ iPOWER ???
Nuclear thermal-hydraulics society was busy verifying advanced new systems by conducting various-scale tests
There were not so many activities on safety improvement of the existing NPPs, whereas new systems have been continuously developed DVI, FD, SDS DVI+, FD+, PAFS PECCS, PCCS, etc
Experiment has played a major role in providing industries & safety authority with confidence whether the new designs can be adopted
Safety analysis tools have been continuously expanded to embrace the unprecedented designs by improving or adding empirical models
Thermal-hydraulic society lacked initiative of first principle-based predictive tool development
Observations
Thermal Hydraulics & Severe Accident Research Division
Observations from R&D Outcomes (2/2)
35
Thermal hydraulic safety research almost reached the limit in particular in the territory of LWRs
Indication of that the current “macroscopic” research SHOULD move to“microscopic” research to figure out what remaining technology gaps are
Or we have to come up with ground-breaking ideas, taking into account the contemporary technology evolution (e.g., ICBM)
A drastic paradigm change in safety analysis is imminent
A deterministic analysis is complemented by UQ due to its inherent uncertainties of model parameters and giving a control to PSA
A CFD-like calculation is growing at a rapid pace and expanding its territory before getting into the regulatory framework
Multi-physics coupled calculation is becoming tangible as a future tool to figure out un-identified remaining safety gaps
Observations
Thermal Hydraulics & Severe Accident Research Division
Contents
Safety Research of KoreaI
TH Safety Research OutcomesII
Opportunities and ChallengesIV
ObservationsIII
36
SummaryV
Thermal Hydraulics & Severe Accident Research Division
Opportunities of Nuclear Thermal-Hydraulics - general
37
Current nuclear safety analyses are based on the computer codes developed in the late 1980s for which desktop PC is usually used
* Joshua Kaizer (NRC), Multi-physics Model Validation Workshop, UCSU, 2017.
Is it really sufficient to deal with all the complex physics of PWRs ?
Validation of the conservative method or model needs more accurate analysis tools
Speedup of HPC (High Performance Computing) continues and leads the current industrial revolution
Thermal Hydraulics & Severe Accident Research Division
Opportunities of Nuclear Thermal-Hydraulics - specific
38
Safety always comes first!Thermal-hydraulic research is a critical field to ensure safety
Increasing affordability of advanced experimental techniques*High resolution imaging, combination of different diagnostics, etc.
Advancement of data science*Data mining, pattern recognition, data assimilation, etc.
Improved UQ tools*Sensitivity analysis, GRS method, Monte Carlo, etc.
Advanced computer science and software engineering*Software architecture, modeling framework, AI, etc.
Affordable data storage and computational power*Cloud, cluster-PC, supercomputer, etc.
Success in theory and application of CFD*CMFD, interface tracking, etc.
Accumulated experience and database Highly experienced experts, databank, etc.
*Courtesy of Nam Dinh, NURETH15, 2013
Thermal Hydraulics & Severe Accident Research Division
Challenges of Nuclear Thermal-Hydraulics - general
39
The major challenges facing the nuclear industry include ensuring power plant safety, protecting reactors from natural disasters and external aggression, and finding effective solutions for long-term waste management.
Thermal hydraulics plays a critical role in ensuring safety to prevent severe accident from taking place in any case zero release
Remaining knowledge gaps threatening safety SHOULD be identified and safety-relevant R&D to be pursued as first action
The challenges to the nuclear safety authority are related to how to rationalize their regulatory practice taking into account continuous evolution in safety technology
Safety margin needs to be evaluated not only by conservative but also by best-estimate tools
Collaborative leadership based on safety knowledge is essential for the nuclear community to survive in recent social environment
Thermal Hydraulics & Severe Accident Research Division 40
① Higher fidelity versus engineering design tool
Technology always evolves; “want” vs. “need” ?What is “good enough” versus “needs basic R&D”? Safety SHOULD be not only assessable but also predictableHigh-fidelity is a right thing to do!
② 5M-featured approach; Multi-D, multi-scale, multi-physics with mechanistic modeling and multi-field approach*
It is more like a new paradigm to thermal hydraulic safety and helps us to find knowledge gap with a link to a target-oriented research to fill the gap
Do we have enough data or need more? This approach will stimulate generation of multi-D,S,P data
*Courtesy of C.H Song, Nuclear Technology, 2016
Challenges of Nuclear Thermal-Hydraulics - specific
Thermal Hydraulics & Severe Accident Research Division 41
③ Convergence research to be pursued
Innovative technology (e.g. ICBM) needs to be coupled with TH Virtual environment for thermal-hydraulic safety (VR, AR, AI, etc.)
Well-prepared for integrated analysis of DSA and PSA UQ by utilizing data science needs to be advanced
④ Best utilization of safety resources by multi-lateral cooperation
Each nuclear stakeholder SHOULD work together to best use the limited safety resources by sharing responsibility for safety Safety authority, industry, institute, academia, etc.
Data sharing, facility sharing, man-power exchange, etc.
It can be realized either in a country or in an international framework, but an independent leadership would be a critical element for success e.g. NUGENIA, CASL, NEAMS, …
Challenges of Nuclear Thermal-Hydraulics - specific
Thermal Hydraulics & Severe Accident Research Division
Challenges of Nuclear Thermal-Hydraulics - specific
42
High Fidelity Simulation in Nuclear Safety
Traditional science approach Theory drives design of experiments Experiments provides discoveries to drive
theory Empirically based modeling and simulation
heavily dependent on staying close to experimental basis
Addition of science based modeling and simulation Science (1st principles) based modeling and
simulation used to extrapolate and predict beyond tested states
Can quickly confirm or disprove theory hypotheses
Improve experiments by predicting “areas of interests” and expected results
Courtesy of Alex R. Larzelere, NEAM Simulation, DOE
Thermal Hydraulics & Severe Accident Research Division
Challenges of Nuclear Thermal-Hydraulics - specific
43
Multi-physics in Nuclear Reactor
Complicated phenomena do not occur independently Radial power distribution in MSLB, CRUD in fuel rod
Necessity for co-simulation with different physics T/H, N/K, Material, Chemistry, etc.
Platforms for multi-physics simulation VERA, MOOSE, SALOME
Neutron
kineticsWater
chemistry
MOOSE Framework
SALOME
I&C
Thermal
hydraulics
Structure/material
Heat
conduction
Multi-physics in nuclear reactor Multi-physics platform
Courtesy of J.R. Lee, CUPID workshop, NPRE/UIUC, 2018
Thermal Hydraulics & Severe Accident Research Division
Challenges of Nuclear Thermal-Hydraulics - specific
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Multi-scale in Nuclear Thermal-Hydraulics
World is multi-scale!
In modeling, a common challenge is determining the correct scale to capture a phenomenon of interest Capture the details you care about and ignore those you don’t
But multiple phenomena interact, often at different scales
So far, current technology isn’t really multiscale. It has just used fine information to build the best coarse model. But it’s a needed part of the process
There is an increasing demand for accurate and realistic simulation of some multi-physics phenomena, e.g. PCI
Thermal Hydraulics & Severe Accident Research Division
Challenges of Nuclear Thermal-Hydraulics - specific
45
Multi-scale in Nuclear Thermal-Hydraulics
Technologies providing high fidelity TH predictions by combining codes with different analysis scale
System, Component, CFD, and DNS scale codes are used Leaded by EU
Courtesy of H.Y. Yoon CUPID workshop, NPRE/UIUC, 2018
DNS-scaleCUPID-SG
CUPID-RV CUPIDMARS
Component-scale
System-scale
CFD-scale
Multi-scale coupled calculations
Upscaling
Thermal Hydraulics & Severe Accident Research Division
Contents
Safety Research of KoreaI
TH Safety Research OutcomesII
Opportunities and ChallengesIV
ObservationsIII
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SummaryV
Thermal Hydraulics & Severe Accident Research Division
Summary
A huge amount of thermal-hydraulic research has been carried out in a well-balanced manner between experiments and analysis in Korea
Its main enabler was a strong initiative toward advanced LWRs and some Gen.IV reactors such as APR1400, APR+, iPOWER, SMART, PGSFR.
Great progress was already achieved in developing advanced analysis tools/methodologies; MARS, SPACE, CUPID, PAPIRUS etc.
Very good research practice was established that experiments and analysis work together to discover phenomena and improve the analysis tools
But, thermal-hydraulic safety research seems to be almost saturated and encounters big challenge for moving forward
It is a formidable task to gain additional safety by following the current research practice.
It is right time to best utilize the contemporary technology to fill the remaining safety gaps. (e.g. 5M-featured approach, ICMB)
In this context, modeling and simulation (M&S) would be a promising new enabler
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Thermal Hydraulics & Severe Accident Research Division
Summary
Cooperation with people living in another discipline starts to be a essential element to achieve the “safety goal”
Convergence research, in particular, cooperating with the modern science technologies is necessary
Data science would play a role to get safety insight from the existing rich data (field data, image data, etc.) and to improve UQ
Best utilization of safety resources by multi-lateral cooperation
Data sharing, facility sharing, man-power exchange, etc.
Japan and Korea have to pursue continuing cooperation to contribute to nuclear safety worldwide and give people an easy mind against nuclear
Cooperation network between J and K is critical to make sure safety in the East Asian area
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Thermal Hydraulics & Severe Accident Research Division 49
Thank for Your Kind Attention!
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