Nuclear Thermal-Hydraulic Safety for the Future I · Nuclear Thermal-Hydraulic Safety for the Future I: The Present and Future of Nuclear Thermal-Hydraulic Safety Research Nov. 19,

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


Contents




ObservationsIII

3

SummaryV


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


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



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



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



Contents



Challenges and OpportunityIV

ObservationsIII

8

SummaryV


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


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




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


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



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


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



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)


DYNAS facility DYNAS flow visualization and void fraction test results



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)


FR2 test results (KfK, 1983)


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


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


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


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


Optimization of Safety System for APR+

잔여수위

잔여수위

SIT lower dead volume

SIT SIT+

Gravity driven SIT to prevent nitrogen gas from released


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


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


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



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


HEMS PECCS


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)


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


□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


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


Research reactor model

Multi-D post-processor

Full-scope simulator, MARS-SIM


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


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


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



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



Code Development

28


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


Code Development

29


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.


Code Development

30


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.


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)



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


simulation

test data

Helium layer erosion transient


Contents




ObservationsIII

33

SummaryV


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


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


Contents




ObservationsIII

36

SummaryV


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


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


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


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, …




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



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



44

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



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


Contents




ObservationsIII

46

SummaryV


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

47


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

48


Thank for Your Kind Attention!

Any Question or Comment?

Documents

Nuclear Thermal-Hydraulic Safety for the Future I · Nuclear Thermal-Hydraulic Safety for the Future I: The Present and Future of Nuclear Thermal-Hydraulic Safety Research Nov. 19,