NEA NUCLEAR SAFETY RESEARCH JOINT PROJECTS WEEK:
Success Stories and Opportunities for Future Developments
9-13 January 2023

NEA NUCLEAR SAFETY RESEARCH JOINT PROJECTS WEEK:
Success Stories and Opportunities for Future Developments
9-13 January 2023

WelcomeDay 2 – Tuesday 10 January

NEA NUCLEAR SAFETY RESEARCH JOINT PROJECTS WEEK:
Success Stories and Opportunities for Future Developments
9-13 January 2023

Questions: Questions, feedback and suggestions
Event public page: Nuclear Energy Agency (NEA) - NEA Nuclear Safety
Research Joint Projects Week: Success Stories and Opportunities for
Future Developments (oecd-nea.org)

Form: Questions, feedback and suggestions link

available in the registration confirmation email

Session 2

Joint Projects for Safety in Design, Learnings andPerspectives

SESSION MODERATOR

Dr Jinzhao ZHANG
Technical Director, Business Area Global
Nuclear, Tractebel (ENGIE), Belgium

SESSION 2: Joint Projects for Safety in Design, Learnings and Perspectives

SESSION 2: Joint Projects for Safety in Design, Learnings and Perspectives

Session 2
Joint Projects for Safety in Design, Learnings andPerspectives►

Examples of Nuclear Fuel Safety Projects

SESSION 2: Joint Projects for Safety in Design, Learnings and Perspectives

QUENCH-ATF

Dr-Ing Th. Walter TROMMHead of the Nuclear Waste Management, Safety and Radiation
Research Programme (NUSAFE), Karlsruhe Institute ofTechnology (KIT), Germany

SESSION 2: Joint Projects for Safety in Design, Learnings and Perspectives

OECD-NEA Joint Undertaking QUENCH-ATF
Juri Stuckert, Martin Steinbrück, Th. Walter Tromm (KIT)NEA Nuclear Safety Research Joint Projects Week, 9-13 Jan 2023
Institute for Applied Materials IAM-AWP and Program Nuclear Waste Management, Safety and Radiation Research (NUSAFE)

KIT – The Research University in the Helmholtz Associationwww.kit.edu

QUENCH-ATF Joint Undertaking

Worldwide first large-scale bundle experiments with Accident-Tolerant
Fuel (ATF) cladding materials under DBA and severe accident conditions
Fact sheet
Duration: Oct. 2021 – Oct. 2025
Budget: 1.6 Mio €20 participants from 9 countries

12

January 2023

Walter Tromm

OECD-NEA QUENCH-ATF project

KIT NUSAFE

Accident tolerant fuel (ATF) cladding
requirements
Excellent high-temperature properties
Oxidation resistance to steam
Reduced hydrogen production and chemical energy releaseHigh mechanical strength to maintain fuel rod integrity and core coolability
Extended coping time for operator to restore cooling system

Compatibility with current fuel and reactor designs
Meet regulatory and operational requirements
Equal or better operational behavior
Corrosion resistance and hydrogen uptake
Irradiation resistance

Licensing
Availability and economics
13

January 2023

Walter Tromm

OECD-NEA QUENCH-ATF project

KIT NUSAFEMost promising ATF cladding concepts
Cr-coated Zr alloys
Still Zr-based near-term solution
Improvement of oxidation, hydrogen uptake,mechanical properties up to 1300°C
Cr-Zr eutectic at 1332°C

FeCrAl alloys
Cheap and easy to produce
DBA LOCA degradation and HT oxidation significantly reduced
Neutronic penalty, relatively low melting temperature,
tritium permeability

SiCf-SiC ceramic matrix composites
First of a kind ceramic composite fuel cladding
Very good neutronic behavior and oxidation resistance up to >1700°C
In-pile hydrothermal corrosion and leak tightness
Images from Framatome, ORNL
and Westinghouse

14

January 2023

Walter Tromm

OECD-NEA QUENCH-ATF project

KIT NUSAFE

KIT-QUENCH activities and expertiseon ATF cladding
Participation in international programs

SiCf-SiC cladding after
64 h at 1600°C in steam

OECD-NEA (EGATFL, TOPATF)
IAEA CRP on Accident Tolerant Fuel Concepts forLight Water Reactors (ACTOF)
EC project IL TROVATORE
Westinghouse CARAT

Large-scale bundle tests in the QUENCH facility
High-temperature oxidation of ATF cladding
materials

Inductively heatedsingle-rod test

Coated Zr alloys (Cr, MAX phases)
Fe-based alloys
SiCf-SiC CMCQUENCH bundle forlarge-scale tests
15

January 2023

Walter Tromm

OECD-NEA QUENCH-ATF project

KIT NUSAFEQUENCH-ATF Joint Undertaking
Three bundle experiments with ATF cladding in the
QUENCH facility
Focus on Cr-coated Zr alloys
Option for SiC cladding in second phase of project
Tubes provided by Westinghouse (US) (and others?)
Design basis and beyond design basis accident conditions

Supporting separate-effects tests
at KIT
Complementary tests at IRSN (France)

Code support for test preparation and code benchmark
exercises, coordinated by GRS

16January 2023

Walter Tromm

OECD-NEA QUENCH-ATF project

KIT NUSAFEQUENCH Facility
Unique out-of-pile bundle facility toinvestigate reflood of an
overheated reactor core
So far, 21 experiments on SA
performed (1996-today)
7 DBA LOCA experiments withseparately pressurized fuel rods
QUENCH-19: First bundle test with
ATF cladding material (FeCrAl)

M. Steinbrück et al., Synopsis andoutcome of the Quench
experimental program,
NED 240 (2010) 1714-1727.
J. Stuckert et al., Results of the
QUENCH-LOCA experimental
program at KIT, JNM 534 (2020)
152143.

17

January 2023

Walter Tromm

OECD-NEA QUENCH-ATF project

KIT NUSAFE

Test matrix
QUENCH-ATF #1 (July 2022)
Cr coated Zr (provided by Westinghouse)
Bundle with 24 heated rods, 12.6mm pitch
Extended LOCA conditions: similar conditions as QUENCH-L3HT (LOCA 3 High
Temperature) with optimized ZIRLOTM, which will allow for comparison

QUENCH-ATF #2 (2023)
Cr coated Zr
Severe accident conditions(above Zr-Cr eutectic)
ATCR under discussion

18

January 2023

Walter Tromm

OECD-NEA QUENCH-ATF project

QUENCH-ATF #3 (2024)
Cr coated Zr or SiC
GO/NOGO on SiC in 2023
Scenario depending on material
and results of previous tests

KIT NUSAFE

Participants from 9 countries
SwedenCzech RepublicSSMÚJV Řež

Switzerland

France

PSI
ENSI

CEA
EDF
Framatome
IRSN

United Kingdom
NNL

Germany
GRS
KIT (Operating Agent)Japan
CRIEPI
JAEA
Russian Federation (suspended)
Bochvar InstituteKurchatov Institute
TVEL
19

January 2023

Walter Tromm

OECD-NEA QUENCH-ATF project

United States
EPRI
GA
USNRC
Westinghouse** under discussion
KIT NUSAFEQUENCH-ATF #1 experimentcross-section of bundle
instrumented with thermocouples

test bundle

Improved
handling of Cr
coated cladding
tubes during
bundleassembly

20

January 2023

Walter Tromm

OECD-NEA QUENCH-ATF project

KIT NUSAFEQUENCH-ATF #1 scenario
Extended LOCA conditions
Heating-up rate during theheating in superheatedsteam 5 K/sPeak cladding
temperature at the end of
the heat-up stage 1600 K
Duration of cool down
stage 200s
Final quenching by water

QUENCH-L3HT scenario

Should demonstrate enhanced performance athigh temperature during a LOCA transient
21

January 2023

Walter Tromm

OECD-NEA QUENCH-ATF project

KIT NUSAFEQUENCH-ATF #1 post-test examinations(KIT/IRSN)
Laser scanner profilometry (pre-/post-test)
Neutron radiography and tomography
protective
Cr2O3(green outerlayer)
Cr
diffusion zone

Zr alloytube cross-section

Burst opening parameters
22

January 2023

Walter Tromm

OECD-NEA QUENCH-ATF project

KIT NUSAFEQUENCH-ATF #1 benchmark exercise
Blind post-test benchmark
Coordinated by Th. Hollands, GRS (DE)Simulation of bundle temperatures and mechanical cladding behaviour (strainof ballooning region and burst parameters)
Strong impact of QUENCH tests on codes V&V
Many organisations are currently working on leveraging their code to address
ATF claddings incl. Cr coated Zr – therefore the timing is optimal
6 organisations volunteered to participate:
GRS/RUB PSS with AC2/ATHLET-CD
KIT with ASTECUSNRC/SNL with MELCOREPRI with MAAPUJV Rez with MELCOR (and TRANSURANUS?)

Launched shortly after test performance, finished 12/2022
After finalisation of blind phase, experimental data were delivered and open
benchmark analysis started
23

January 2023

Walter Tromm

OECD-NEA QUENCH-ATF project

KIT NUSAFEKIT activities and collaborations of relevancefor QUENCH-ATF
CSARP (Cooperative Severe Accident Research Program of the US Nuclear
Regulatory Commission)

EC Research Program IL TROVATORE (Innovative cladding materials foradvanced accident-tolerant energy systems)
IAEA Coordinated Research Project on Testing and Simulation of Advanced
Technology Fuels (ATF-TS)
OECD-NEA activities (e.g. TOPATF, EGATFL, EGIFE)
OECD-NEA project TCOFF (Thermodynamic Characterization of Fuel Debris
and Fission Products Based on Scenario Analysis of Severe Accident
Progression at Fukushima-Daiichi Nuclear Power Station)

24

January 2023

Walter Tromm

OECD-NEA QUENCH-ATF project

KIT NUSAFEExample:
IAEA Coordinated Research Project (CRP): “Testing and Simulation for Advanced
Technology and Accident Tolerant Fuels (ATF-TS)”
Start Date: 6 April 2020 Expected End Date: 1 August 2024
Participants:
29 organizations from 22 countries
Objectives:
To perform experimental tests including single rod and bundle tests on ATFs'
performance under normal, Design Basis Accident (DBA) and Design Extension (DE)conditions
To benchmark fuel codes against new test data either obtained during the CRP or fromexisting data relevant to advanced fuel and cladding concepts
To develop LOCA evaluation methodology for ATF performance with a view for NPP
applications
Outputs:
ATF performance dataset and bundle tests for normal operating conditions, DBAs andDECs
Experimentally validated fuel codes and modelling methods for ATF
A demonstration case for the application of the new ATF materials and validated fuelmodels for assessing fuel performance and quantifying fuel margins for LOCA
25

January 2023

Walter Tromm

OECD-NEA QUENCH-ATF project

KIT NUSAFE

Concluding remarks for the QUENCH-ATF
project
QUENCH-ATF project will provide valuable data on bundle behavior of ATF claddings through aseries of three tests
Making the most of the unique QUENCH facility
Test #1 on Cr coated Zr in LOCA conditions successfully conducted in July 2022
Second test with more severe conditions under preparation
Cooperation within QUENCH-ATF beneficial for many partners, e.g.:
QUENCH test and post-test examination by operating agent KIT,
cooperation with PSI, CEA and ILL/Grenoble for neutron-tomography

Complementary post-test examination kindly provided by IRSN
Analytical support and benchmark exercise kindly coordinated by GRS

26

January 2023

Walter Tromm

OECD-NEA QUENCH-ATF project

KIT NUSAFE

Possible Outlook for a futurecontinuation QUENCH-ATF project?
Continuation of the project with more detailed
investigation on Design Basis Conditions and Design
Extension Conditions for ATF Cladding Materials.
Maintaining and further developing facilities andcompetences in the field of material investigations
Attracting young scientists worldwide
E.g. 2 facilities at KIT available for theseinvestigations:
COSMOS-H for DBC up to critical heat flux
QUENCH for Severe Accident Conditions

27

January 2023

Walter Tromm

OECD-NEA QUENCH-ATF project

KIT NUSAFE

General Outlook and Closing Remarks
Various new cladding materials under development and investigation
Needs for experimentally accompanied material research for designbasis conditions and design extension conditions
Infrastructures and experimental facilities have to be maintained forout-of-pile tests and in-pile tests
Engagement of industry as the main driver for innovation needed
But as well strong involvement of regulators and TSOs mandatory
Collaboration of research centers and universities for:scientific expertise, experimental facilities, code development
Education and training of next generation of young scientists
Maintaining and further development of research facilities

OECD / NEA is a strong support for the cooperation
28

January 2023

Walter Tromm

OECD-NEA QUENCH-ATF project

KIT NUSAFESESSION 2: Joint Projects for Safety in Design, Learnings and Perspectives

CABRI CIPDr Vincent BUSSERCABRI CIP Project Manager, Institut de Radioprotection etde Sûreté Nucléaire (IRSN), FranceSESSION 2: Joint Projects for Safety in Design, Learnings and Perspectives

OVERVIEW OF THE OECD/NEA CIP PROJECTIRSN/FRM-296 ind. 06

Vincent BUSSER, François BARREGENERAL DESCRIPTION OF THE CIP PROJECT[ SEVERAL JOINT PROJECTS IN FUEL SAFETY

https://www.oecd-nea.org/jcms/pl_25445/studsvik-cladding-integrity-project-scip

Studsvik
Cladding
Integrity
Project SCIP

Halden

Reactor
Project

QUENCH-

ATF
https://www.oecd-nea.org/jcms/pl_36597/quench-atf-project

https://www.oecd-nea.org/jcms/pl_24970/halden-reactor-project-fuels-and-material

SecondFrameworkfor Irradiation
Experiments
FIDES IIhttps://www.oecd-nea.org/jcms/pl_70867/second-framework-for-irradiation-experiments-fides-ii

CABRI
International
Program CIP
https://www.oecd-nea.org/jcms/pl_24821/cabri-international-project-cip

SESSION 2: JOINT PROJECTS FOR SAFETY IN DESIGN, LEARNINGS AND PERSPECTIVES - OVERVIEW OF THE CIP PROJECT - 10 JANUARY 2023

32

GENERAL DESCRIPTION OF THE CIP PROJECT[ REP-NA PROGRAM… CIP

CIP : First tests in the sodium loop on fuelCIP0-1 2001
CIP0-2 2002
REP-Na1 1993(…)
REP-Na12 2000
Tests dedicatedto research
reactors

Sodium loop implementation
New core (UO2 rods)

1963

1977

1993

2003

CABRI
refurbishment
and PWL
implementation

2001
CIP partners sign theumbrella agreement on
4 February
International consensus on
the need for tests in PWR
representative conditions-> Pressurized Water Loop

CABRI firstneutron early 64

SESSION 2: JOINT PROJECTS FOR SAFETY IN DESIGN, LEARNINGS AND PERSPECTIVES - OVERVIEW OF THE CIP PROJECT - 10 JANUARY 2023

33

4

GENERAL DESCRIPTION OF THE CIP PROJECT[ IDENTITY CARD

▌ CABRI is owned and operated by A decree gives priority to IRSN in order to use the reactor for its research programs on fuel safety Refurbishment and operation of the reactor are 100 % sponsored by IRSN, CEA remaining the owner▌ 2nd phase: CABRI International Program CIP▌ 1st phase: CABRI REP-Na program 2 in sodium (CIP0-1/CIP0-2) + 10 in water Sodium loop, 12 tests
 Operating Agent: IRSN PCMI phase, rod failure threshold and mechanism, fuel ejection
 Partners: 12 foreign countries,
 1992-2002 with EDF cooperation + US-NRC support
16 partners representing:– Operators (EPRI, EDF)
– Safety authorities (CSN, NRI, HSE, …)– TSO (IRSN, NRC, GRS, …)
 Budget: Initial cost of EUR 74 million
 Main Objectives:
 To carry out integral tests performed on irradiated material
 To improve the knowledge for these phenomena during RIACABRI perfectly mimics Pressurized
Water Reactor conditions

The reactor design allows
performing rapid power transientstransient:– boiling crisis (post-DNB) phenomena,– post-failure (Fuel-Coolant Interaction FCI) events,– microstructure influence,– advanced cladding.

SESSION 2: JOINT PROJECTS FOR SAFETY IN DESIGN, LEARNINGS AND PERSPECTIVES - OVERVIEW OF THE CIP PROJECT - 10 JANUARY 2023

34

MAIN OUTCOMES AND BENEFITS[ CIP TEST  AMBITIOUS AND COMPLEX•Pool Type Reactor•Core Size: 65x65x80cm•Pulse: 10-80ms•Core Fuel Rods :1488 UO2 & Stainless Steel Cladding•Pressurized water circulation :P = 155 bar, T = 280°C, v = 4m/s

•Unique design and fabrication of
the test device•Embedding of the
instrumentation

Reactor

Test device• Hodoscope system• Non Destructive
Examinations• X-Ray radiography• X-Ray transmission
tomography• Gamma-scanning

Observation offuel in real time

Post-test
examination

 DEVELOPMENT OF A UNIQUE FACILITY WITH ADVANCED EXPERIMENTAL TOOLS TECHNOLOGICAL CHALLENGESESSION 2: JOINT PROJECTS FOR SAFETY IN DESIGN, LEARNINGS AND PERSPECTIVES - OVERVIEW OF THE CIP PROJECT - 10 JANUARY 2023

35

MAIN OUTCOMES AND BENEFITS[ CIP : TEST DEVICE▌ A unique test device
 Tube: 5 meters in length, 140 kg average diameter of approximately 100 mm
 Nominal operating conditions : those of the pressurizedwater loop (155 bar/280°C),
 But also, experimental conditions of the test during thetransient (up to 500 bar/350°C).
 The instrumentation embedded in the test device is specificthat meets very tough constraints and required several
years of development.
 ~ 50 detectors inside the test device
 2 Million EUR per test device

SESSION 2: JOINT PROJECTS FOR SAFETY IN DESIGN, LEARNINGS AND PERSPECTIVES - OVERVIEW OF THE CIP PROJECT - 10 JANUARY 2023

36

MAIN OUTCOMES AND BENEFITS[ CIP : OBSERVATION OF FUEL IN REAL TIME
▌ Hodoscope system On-line fuel motion detection (displacement, ejection,

relocation)
 measurement of the test rod fissile length measurement of the core power profile measurement of the pulse width
306 counting tracks:• 51 rows x 3 columns collimator• up to 1 ms acquisition rate
153 Fission Chambers (FC) power transient measurement
153 Proton Recoil Counters (PR)“low” power measurement (up to ~100 MW).

SESSION 2: JOINT PROJECTS FOR SAFETY IN DESIGN, LEARNINGS AND PERSPECTIVES - OVERVIEW OF THE CIP PROJECT - 10 JANUARY 2023

37

NUCLEAR SAFETY APPLICATION▌CABRI REP-Na program performed during the 1992-2002 period▌Tests performed in the CABRI facility with the sodium loop  Study limited to the PCMI (Pellet Clad Mechanical interaction)phase and on both UO2 and MOX (high BU up to 64 GWd/t, Zr-4 cladding) of the RIA situation▌Beyond an external corrosion layer thickness of 80 microns, cladding integrity for Zircaloy-4 (material used historically for

PWRs) can’t be guaranteed
Clad

Fuel

Highly corroded cladding
with spallation

CABRI REP-Na 8▌At the international level, results from CABRI REP-Na induced and provided input to considerations on the necessity torevisit safety criteria established on lightly irradiated fuel.
▌In France, pending the complete deployment of cladding materials with improved performance, IRSN recommended tolimit the operating conditions for Zircaloy-4 containing cores in order to mitigate the risk
Possible spalling Operating restrictions

No spalling
0 µm

80 µm

Unauthorized Reactor shutdown

Oxide
thickness

108 µmSESSION 2: JOINT PROJECTS FOR SAFETY IN DESIGN, LEARNINGS AND PERSPECTIVES - OVERVIEW OF THE CIP PROJECT - 10 JANUARY 2023

38

NUCLEAR SAFETY APPLICATION▌ CIP
 Study fuel rod behavior of PWR under RIA condition
 Integrity of fuel rods in prototypical irradiation conditions (pressurized water loop) Post-DNB and post-failure fuel behavior : main phenomena studied  Risk of rod failure / criteria
 Consequences in case of rod failure Content:
 12 experiments (3 tests already performed 2002/2018)
 UO2 and MOX Fuel Rods (BU up to 100GWd/t) with:
 Current cladding materials (M5, Zirlo,…) New cladding Materials (Optimized Zirlo, M-MDA)
 Validation of models in code (FUEL+/SCANAIR software, …)

SESSION 2: JOINT PROJECTS FOR SAFETY IN DESIGN, LEARNINGS AND PERSPECTIVES - OVERVIEW OF THE CIP PROJECT - 10 JANUARY 2023

39

NUCLEAR SAFETY APPLICATION▌ CIP Matrix
Objective

Completedin Na

Test in
NSRR

Test

Rod

CIP0-1UO2 Zirlo 70 GWd/t

CIP0-2UO2 M5 75 GWd/t

CIPQ

MOX Zr4 47 GWd/t

Loop qualification – Boiling
crisisX

CIP1-2BUO2 M5 77 GWd/t

UO2 - Boiling crisis

X

X

CIP3-1RUO2 Zirlo 75 GWd/t

Post failure events

X

X

CIP4-1P

MOX M5 65 GWd/t

MOX - Boiling crisis

X

CIP4-1PHP

MOX M5 65 GWd/t

Effect of filling pressure
90b vs 50b in other tests

X

CIP4-2

MOX-SBR Zr4 60 GWd/t

Post failure events

CIP3-3UO2 Opt. Zirlo

New cladding material

CIPX

UO2 M-MDA SR

New cladding material

CIPY

UO2 doped Cr2O3

New fuel (PCI-remed)

CIPZ

Optimized Zirlo withliner

New cladding material

Completedin CABRI
2001

Reference test in sodium loop

2002
2018
2022

X

SESSION 2: JOINT PROJECTS FOR SAFETY IN DESIGN, LEARNINGS AND PERSPECTIVES - OVERVIEW OF THE CIP PROJECT - 10 JANUARY 2023

40

NETWORKING▌ Stakeholders with different roles in their countries: Operators, Safety authorities, TSO▌ Collaboration required with industrial partners (EPRI, EDF, …) to provide irradiated material to be tested.
▌ RIA activities in the frame of WGFS

RIA benchmark

CIP AnalyticalGroup(CIP partners)▌ R&D complementary programs involving some of partners : FGD (IRSN/JAEA), PROMETRA (IRSN/CEA/EDF)

SESSION 2: JOINT PROJECTS FOR SAFETY IN DESIGN, LEARNINGS AND PERSPECTIVES - OVERVIEW OF THE CIP PROJECT - 10 JANUARY 2023

41

VALUES▌ Several Publications to share results with scientific community▌ Formation of young researchers throughout the project: many PhD thesis,

Post-doctoral positions in fuel safety connected to RIA issues▌ CIP Partners are main contributors of the SOAR RIA

Working Group on Fuel Safety (WGFS) / NEA▌ CIP will provide experimental data for code validation

SESSION 2: JOINT PROJECTS FOR SAFETY IN DESIGN, LEARNINGS AND PERSPECTIVES - OVERVIEW OF THE CIP PROJECT - 10 JANUARY 2023

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FEEDBACK ON MAIN CHALLENGES RELATED TO THE PROJECT DEVELOPMENT• To maintain the CABRI
experimental platform at a highlevel of operability for the whole
duration of the project with ahigh constraint of irradiatedmaterial• Difficulties to cover financial
aspects in case of technical issues
on the CABRI reactor

Experimental
Platform•

To maintain, throughout the
project, the skills at a high levelof competence on theexperimental devices, including
instrumentation and advanced
experimental techniques onirradiated materials.

Experimental
analysis•

To ensure that the objectives of
the test remain as high interestfor partners from the beginningof the project (Safety authorities,
TSO, operator – different
countries)

Safetyanalysis▌ CABRI is licensed by IAEA as Capacity Building: formation of young researchers, PhD Thesis, Post-

doctoral.
SESSION 2: JOINT PROJECTS FOR SAFETY IN DESIGN, LEARNINGS AND PERSPECTIVES - OVERVIEW OF THE CIP PROJECT - 10 JANUARY 2023

43

INSGHTS ON FUTURE R&D NEEDS AND TRENDS• Keep opportunity ofcompleting in-pile
special tests for
improving knowledge
on fuel safety studies.

FIDES-II,…

•

Need of qualification of
new advanced cladding
and fuel materials (ATF,…), in particular the
behavior in DBA-RIA
transient

Materials• Simulation implies
models that requires
more data to enablevalidation of numerical
tools.

Validationof codesSESSION 2: JOINT PROJECTS FOR SAFETY IN DESIGN, LEARNINGS AND PERSPECTIVES - OVERVIEW OF THE CIP PROJECT - 10 JANUARY 2023

44

CONCLUSIONS• Safety topic complex to address• Technical complexity of the refurbishment• Budget challenges to cover all technical hazards• Integral tests : irradiated conditions, Pressurized
Water Loop
• Several experimental ways (instrumentation,
hodoscope system, IRIS pre & post-test
examinations)

Complex•Throughout the project• Human capital investment• Advanced relevant skills• Discussions with partners todefine the test matrix• Formation of young researchers
throughout the project•

Improvement of the safety
approaches in RIA conditions• Improvement of theunderstanding and themodelling of RIA• Development of specificsoftwareSkills,Networking

Use for safety
analysis

SESSION 2: JOINT PROJECTS FOR SAFETY IN DESIGN, LEARNINGS AND PERSPECTIVES - OVERVIEW OF THE CIP PROJECT - 10 JANUARY 2023

45

Thank you for your attention
For more information on IRSN website :https://www.irsn.fr/EN/Research/Research-organisation/Researchprogrammes/CABRI-International-program/Pages/CABRI-CIP-program.aspx

SESSION 2: JOINT PROJECTS FOR SAFETY IN DESIGN, LEARNINGS AND PERSPECTIVES - OVERVIEW OF THE CIP PROJECT - 10 JANUARY 2023

46

Session 2
Joint Projects for Safety in Design, Learnings andPerspectives► Examples of Primary and Secondary Side Circuit
Thermal-Hydraulics, Passive Safety Systems Projects

SESSION 2: Joint Projects for Safety in Design, Learnings and Perspectives

ATLAS

Dr Kyoung-Ho KANG

Principal Researcher, Director, Korea Atomic Energy ResearchInstitute (KAERI), Korea

SESSION 2: Joint Projects for Safety in Design, Learnings and Perspectives

Examples of Primary and Secondary Circuit ThermalHydraulics, Passive Safety Systems Projects

ATLAS
10 January, 2023

Kyoung-Ho KANG (khkang@kaeri.re.kr), Director
Innovative System Safety Research Division

2

Overview of ATLAS Project (1) OECD/NEA ATLAS International Joint Project Operation of OECD/NEA ATLAS international joint project(2014.04. ~ present;
now 3rd project is on-going)• With an aim of safety evaluation of LWR and validation & improvement of safety
analysis codes

 Overview of 3rd OECD/NEA ATLAS international joint project• Period: 2021.01.01.~2024.12.31.
• Participants: 19 organizations from 11 countries• Contribute to resolution of safety issues and advancement of safety analysis
technology

Innovative System Safety Research Division3

Overview of ATLAS Project (2) ATLAS Integral Effect Test Facility

Innovative System Safety Research Division4Facilities and Competences Development (1) ATLAS Upgrade (1) ATLAS project has helped maintain and upgrade the ATLAS facility including
instrumentation.
• Manufacturing of a new heater rod bundle and RPV (‘17)– The TC locations were shuffled and capability to capture the CET behavior was improved.
– A profile TC was installed inside the RPV upper head to observe 3D flow behavior.

• Replacement of existing safety injection tanks (SITs) with Hybrid SITs (‘17)– Passive ECCS concept can be investigated by utilizing Hybrid SITs.

• Revamping steam generators (‘21)– U-tube bundles were replaced with new ones with additional fluid temperature
measurement.

New heater rod bundle
Innovative System Safety Research Division

Installation of TCs

Concept of PECCS
5

Facilities and Competences Development (2) ATLAS Upgrade (2) ATLAS project has sustained development of highly qualified competences ofexperiment.
• Reactor coolant system (RCS)-containment integrated IET• Multi-dimensional thermal-hydraulic behaviors inside RPV, SG, containment, andvarious passive safety systems such as PAFS and PECCS

 Collaboration with University
 One doctorate student stayed at KAERI2019 to October 2019.

from June• Ms. Lorduy Alos Maria (Polytechnic Univ. of Valencia, Spain) Work scope under framework of PhD thesis• Scaling method between test facilities• Evaluation of code predictive capability based on test data

RCS-containment IET

 KAERI is open to collaborate with universities under the framework of ATLAS
project.
Innovative System Safety Research Division6

Nuclear Safety Applications (1) Operation of Benchmark Activities ATLAS project has operated benchmark activities to allow participants to gain more insights
into code capabilities through sharing relevant information and experiences.
• Demonstration of code/model applicability to simulation of transient and relevant phenomena• Identification of the modelling options and uncertain parameters which have significant impact oncode prediction of main phenomena

 Test items for benchmark activities• ATLAS-1: A5.1 (1% cold leg break SBLOCA: Counterpart test of LSTF SB-CL-32)– 15 participants adopting 8 different codes• ATLAS-2: B5.1 (1” RPV top break LOCA: Counterpart test of LSTF SB-PV-07)– 11 participants adopting 8 different codes• ATLAS-3: C2.3 (Steam line break with operation of PAFS)– 9 participants adopting 5 different codes Recommendations for future benchmark exercise• During the preparation of benchmark activity, there should be a compilation of the list of key
parameters and identification of potentially significant parameters.
– Define clearly all requested parameters in advance– The OA should provide a list of physical phenomena significant for the designated test

• Combined uncertainty and sensitivity analyses are recommended for evaluation of code performance
Innovative System Safety Research Division7

Nuclear Safety Applications (2) Application of ATLAS Test Data ATLAS data have been used to assess the design of NPP under challenging
accident scenarios including DBA and DEC and also to assess the new design
features like the passive safety systems.
• Contribute to improvement of analytical applicability to get further steps in V&V ofsystem safety analysis codes and methods (BEPU, scaling etc.)

 Especially in Korea, ATLAS data have been used to improve the phenomenabased SPACE model and the scale-up applicability of SPACE code.
• CCFL, loop seal clearing, critical flow, liquid off-take, NC effect, etc.
• Compare other scale IET facilities ATLAS data have been used in the system code sensitivity crosswalk led by USNRC and EDF under the framework of ATLAS-3 project.
• Analyze and compare the sensitivities across several system codes as opposed to the
traditional system code sensitivity study that aims to find the sensitivities of a single
code• Comparison of figure of merit (FOM) predictions over time
Innovative System Safety Research Division

8

Networking
 Participants of OECD/NEA ATLAS Project Various stakeholders have participated in the ATLAS project.
• Regulators: NRC (USA), CSN (Spain), FANR (UAE)• TSO: BelV (Belgium), GRS (Germany), KINS (Korea)• Research Institutes: CEA (France), PSI (Switzerland), UJV (Czech), NPIC (China), KAERI(Korea)• Industry: EDF (France), CNPRI (China), Tractebel (Belgium), KHNP (Korea), KEPCO
E&C (Korea), KEPCO NF (Korea), DOOSAN (Korea)

 Modality of establishment of a common basis on safety issues• For each test item, OA performed a pre-test analysis with a system analysis code to
define initial & boundary conditions of test and to confirm major thermal-hydraulicresponses of interests.
• Very active pre- and post-analyses have been done by the project participants with
their analysis codes.
• Test specifications for each test item were finally settled by lively discussion
between participants.
Innovative System Safety Research Division9

Values Dissemination of Major Outcomes ATLAS data have been transferred to the NEA Data Bank for potential
distribution to NEA members.
• Final integration reports including benchmark activity report are available at the
NEA Data Bank.

 Joint workshops between ATLAS and PKL/ETHARINUS projects have beenorganized to share experiences and practices among both project participants.
• 1st workshop: Lucca, Italy (13-15 April 2016)• 2nd workshop: Barcelona, Spain (7-9 November 2018)• 3rd workshop: Barcelona, Spain (7-9 November 2023)

 “OECD/NEA international programs” was organized as a special session of
NUTHOS-11 conference (‘16): 12 papers were presented.
 Journal papers (8 papers) and many conference papers (11 papers) have been
published by OA and participants under the framework of ATLAS project.

Innovative System Safety Research Division10

Feedback on Main Challenges related to the Project
Development (1) Main Challenges in Establishing New Project Decline of needs for developing a new/advanced large PWR• The trend has been changing from a large PWR to a SMR.
– Key phenomena in the SMR still need to be investigated to validate the system analysis codes.
– Experimental investigation for the SMR-specific safety systems can be strongly related to theknowledge from the PWR research: Natural convection, boiling, condensation, andperformance of passive safety systems

 System analysis codes have reached an advanced state of development so thatnowadays, they are not being as actively developed as they were before.
• Investigation of multi-scale and multi-physics phenomena with the high-precision
data and analysis will be essential to enhancing the technology of reactor safety
analysis.

 Identification of test topics from a nuclear safety enhancement point of view
• Essential to launching a new joint research project• Feedback from potential participants are very important to establish research
priorities.
Innovative System Safety Research Division11

Feedback on Main Challenges related to the Project
Development (2) Main Challenges: Project Initiation & Development
 Delay or cancellation of participation in the international cooperation project
can occur as a result of political issues.
 Earlier set-up of the budget and project agreement among partners could
make project launching more efficient.
• Declaration in time from promising participants will be helpful to set-up the budget
and agreement.

 The chance to share and discuss program scope should be provided prior toinitiating a project.
• Identification of experimental topics from a safety enhancement point of view

Innovative System Safety Research Division

12

Insights on Future R&D Needs and Trends (1) Follow-up Phase of ATLAS Project KAERI has a plan to continue the ATLAS project in order to contribute toadvancement of safety analysis technology and resolution of safety issues.
• Further investigation of the thermal-hydraulic phenomena in the RCS-containment
integrated IET• Effectiveness of operator’s action against design extension conditions• Performance optimization of passive safety systems in PWR and SMR– Passive residual heat removal system; Passive ECCS
– Passive containment cooling system

Innovative System Safety Research Division13

Insights on Future R&D Needs and Trends (2) Future Research Program utilizing the ATLAS Facility
 Long-term domestic national R&D program: ‘22 ~ ‘29• Optimization technology for accident management (AM) strategy• Validation for performance of containment safety systems

 IET for Validation of Innovative SMR Safety
 Long-term domestic national R&D program: ‘23 ~ ’28• Construction of new IET facility– Commissioning test will be performed in 2027.

Innovative System Safety Research Division

14

Conclusion Perspectives for Nuclear Safety Research Although the numerical solution technology of thermal-hydraulic system
codes has matured, the uncertainty of the physical models is still large andneeds to be improved.
• Experiments that can be applied to physical model development by involving codedevelopers from experimental design are important.
• Development and V&V of multi-scale and multi-physics code system including
uncertainty quantification are essential to enhance nuclear safety analysis
technology.

 Exchange of information and experience in the field of SMR research would bevaluable by organizing network among experts (experimental & analytical side)• Provide the best guideline for the design of an integral effect test facility simulating
SMR, code qualification process, methodology development for advanced
technologies of SMR.

Innovative System Safety Research Division

15

SESSION 2: Joint Projects for Safety in Design, Learnings and Perspectives

RBHT

Dr Stephen M. BAJOREK

Senior Technical Advisor for Thermal Hydraulics, Nuclear
Regulatory Commission (NRC), United States

SESSION 2: Joint Projects for Safety in Design, Learnings and Perspectives

NEA Joint Project on
Rod Bundle Heat Transfer(RBHT)
Stephen M. Bajorek, Ph.D.
Senior Technical Advisor
Office of Nuclear Regulatory Research
United States Nuclear Regulatory Commission

2023 NEA Nuclear Safety Research Joint Projects Week
January 10, 2023

52

Introduction
 The RBHT joint project was a 3-year effort involving 20

organizations from 12 countries.
 The test facility is located at the Pennsylvania State

University. It was designed primarily to produce detailed
data to assist in code development and assessment. RBHT
is heavily instrumented.
 The present joint project was a combined experimental /analytical study. The project conducted a series ofcoordinated reflood experiments and benchmark activities.
 Safety Benefit:

Improved accuracy in simulation of large andsmall break LOCA. Supports power uprates.
53

Project Member Organizations


The United States Nuclear Regulatory Commission (USNRC), USA

Tractebel Engineering S.A., Belgium

Bel V, Belgium

UJV Rez, Czech Republic

Teknologian tutkimuskeskus VTT Oy (VTT), FinlandInstitut de Radioprotection et de Sûreté Nucléaire (IRSN), FranceCommissariat à l’Energie Atomique et aux énergies alternatives (CEA), FranceElectricité de France (EDF), France

Framatome, France

Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) gGmbH, Germany



Nuclear and Industrial Engineering (NINE), Italy



Nuclear Regulation Authority (NRA), Japan

Korea Atomic Energy Research Institute (KAERI), Korea

Korea Institute of Nuclear Safety (KINS), Korea

Korea Hydro and Nuclear Power – Central Research Institute (KHNP CRI), Korea



KEPCO Nuclear Fuel (KEPCO NF), Korea

Consejo de Seguridad Nuclear (CSN), Spain

Swedish Radiation Safety Authority (SSM), Sweden



Swiss Federal Nuclear Safety Inspectorate (ENSI), Switzerland

The Paul Scherrer Institute (PSI), Switzerland

54

Project Schedule Initial Proposal:

2018 Expert’s Meeting: July 2019 at PSU. (In-Person)
 Test Series 1: 2019 – 2020 (testing & QLR prep) Workshop #1: October 2020 (test series 1 results, series 2planning)(Virtual) Test Series 2:2021 (testing & QLR prep) Workshop #2: March 2021 (test series 1 benchmark, distribution oftest series 2 results) (Virtual) Workshop #3:

November 2021 (test series 2 results) (Virtual) Workshop #4: May 2022 (test series 2 benchmark and finalization)(Virtual) Two Volume Final Report :November 2022

55

Bundle Cross-Section(9.5 mm)(12.6 mm)

Instrumentation•(280) embedded heater rod thermocouples•(50) spacer grid strap and steam thermocouples - unique•(39) steam probe rake thermocouples•(~75) other miscellaneous thermocouples throughout system•(21) fine span DP cells on flow housing•(6) pairs of quartz window openings, spanning grid locations•

Pressure measurement throughout system•

Carry-over tanks with temperature and level measurements•

Flow meters on inlet and exhaust lines•

HP (512) channel data acquisition system, up to 20Hz data rate•(2) Laser illuminated, high speed droplet cameras- high detail- high detail- high detail- uniqueSpacer Grid Droplet Breakup

Rapid measurement of drop size and velocity enables
improved models and correlations . . .

58

Test Series / Benchmark Studies

Two test series with Two benchmark studies

 “Open” Tests•••

Wide range of tests considered. Provide basic assessment.
Data provided to all participants as it became available.
Benchmark 1 simulations for wide range of conditions.

 “Semi-Blind” Tests•••

Test conditions similar to Open Tests, but with additional complexity.
Only initial and boundary conditions released to participants.
Benchmark 2 simulations included uncertainty methods.

59

Products
Products of the Project:• Volume 1:

Facility Description, Open Tests and Simulations• Volume 2:

Blind Tests and Simulations, Uncertainty Results• Quick Look Reports• Experimental Data

Code Assessment
Correlation Development• Data on spacer grid phenomena.
60

Benchmarks
 Identified several potential deficiencies inthermal-hydraulic codes, which may affect:•

Large Break LOCA prediction of PCT•

Thermal-Hydraulic predictions for low flow rates(i.e. SMRs and Passive Cooling Systems)

 Allowed participants to improve code
uncertainty methodologies (important tosimulation of LOCA and power uprates).
61

Simulation Findings II
Overprediction of Peak Temperatures:

62

Conclusions / Findings


Reflood thermal-hydraulics remains a challenge.
Modeling some phenomena such as entrainment andspacer grid effects are difficult.



The RBHT project has provided new and unique datafor code assessment and model development.

Benchmark studies identified areas where analysis
codes and BEPU can be improved.



Safety benefits include more accurate thermalhydraulic codes and a better understanding of accident
phenomena.

63

Challenge Question #2
Challenge 2 – How safety research joint projects could better serve development andmaintenance of key competences and research infrastructures, including education of
new talents for the future of the nuclear energy sector?

(1) Joint projects help preserve the research infrastructure by
spreading the cost to several organizations. No single
organization must provide all of the resources.
(2) Education and new talents are enhanced by providing research
opportunities to younger staff members.
(3) Innovation is important for the future. Solving “old”
problems with new instrumentation or new computational
methods is something joint projects can support (and possibly
require).
64

Challenge Question #3
Challenge 3 – What could be the new approaches and arguments for facilitating
decisions from public and private stakeholders of the nuclear energy sector to fundthe future safety research joint projects for the benefit of nuclear innovation andsafety in general?

There is a large variety of stakeholders in the nuclear sector, andwhat is important to a regulator may or may not be important tovendors/utilities, or of interest to academic community.
Consider the formation of an international advisory board tosuggest research needs and joint projects. Board members should
be independent and represent a range of stakeholders.

65

66

Panel SESSION 2: Joint Projects for Safety in Design, Learnings and Perspectives

Dr-Ing Th. Walter
TROMM

Dr François
BARRÉ

Dr Kyoung-Ho
KANG

Dr Stephen M.
BAJOREK

Head of NUSAFE
Programme,
Karlsruhe Institute ofTechnology (KIT),
Germany

Deputy Director of
Safety Research,Institut deRadioprotection etde Sûreté Nucléaire
(IRSN), France

Principal
Researcher,
Director, Korea
Atomic Energy
Research Institute(KAERI), Korea

Senior TechnicalAdvisor for Thermal
Hydraulics, Nuclear
Regulatory
Commission (NRC),
United States

Mr Jonathan
WRIGHT

Dr Ki Yong
CHOI

Head of the Fuel
Materials Centre of
Excellence,
Westinghouse
Electric Company,
SwedenSenior Vice-Presidentof Intelligent Nuclear
Safety Research
Department, Korea
Atomic Energy
Research Institute(KAERI), Korea

Panel SESSION 2: Joint Projects for Safety in Design, Learnings and Perspectives

Jonathan WRIGHT

Head of the Fuel Materials Centre of Excellence, Westinghouse
Electric Company, Sweden

Panel SESSION 2: Joint Projects for Safety in Design, Learnings and Perspectives

Panel SESSION 2: Joint Projects for Safety in Design, Learnings and Perspectives

Dr Ki Yong CHOI
Senior Vice-President of Intelligent Nuclear Safety Research
Department, Korea Atomic Energy Research Institute (KAERI), Korea

Panel SESSION 2: Joint Projects for Safety in Design, Learnings and Perspectives

Panel SESSION 2: Joint Projects for Safety in Design, Learnings and Perspectives

Dr François BARRÉ

Deputy Director of Safety Research, Institut de Radioprotection
et de Sûreté Nucléaire (IRSN), FrancePanel SESSION 2: Joint Projects for Safety in Design, Learnings and Perspectives

NEA NUCLEAR SAFETY RESEARCH JOINT PROJECTS WEEK:
Success Stories and Opportunities for Future Developments
9-13 January 2023

Thank you for your participation today and see you all tomorrow!
Questions, feedback and suggestions link available in the registration confirmation email
Event public page: Nuclear Energy Agency (NEA) - NEA Nuclear Safety Research Joint Projects Week:

Success Stories and Opportunities for Future Developments (oecd-nea.org)