Nuclear Safety2021

Nuclear Safety Research

Support Facilities
for Existing and Advanced
Reactors: 2021 Update

NE A

Nuclear Safety

Nuclear Safety Research Support Facilities for
Existing and Advanced Reactors: 2021 Update© OECD 2021
NEA No. 7565
NUCLEAR ENERGY AGENCY
ORGANISATION FOR ECONOMIC CO-OPERATION AND DEVELOPMENTORGANISATION FOR ECONOMIC CO-OPERATION AND DEVELOPMENTThe OECD is a unique forum where the governments of 38 democracies work together to address the economic,
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Nuclear Energy Agency.

NUCLEAR ENERGY AGENCYThe OECD Nuclear Energy Agency (NEA) was established on 1 February 1958. Current NEA membership consists of
34 countries: Argentina, Australia, Austria, Belgium, Bulgaria, Canada, the Czech Republic, Denmark, Finland, France,
Germany, Greece, Hungary, Iceland, Ireland, Italy, Japan, Korea, Luxembourg, Mexico, the Netherlands, Norway,
Poland, Portugal, Romania, Russia, the Slovak Republic, Slovenia, Spain, Sweden, Switzerland, Turkey, theUnited Kingdom and the United States. The European Commission and the International Atomic Energy Agency also
take part in the work of the Agency.
The mission of the NEA is:–

to assist its member countries in maintaining and further developing, through international co-operation,
the scientific, technological and legal bases required for a safe, environmentally sound and economicaluse of nuclear energy for peaceful purposes;–

to provide authoritative assessments and to forge common understandings on key issues as input togovernment decisions on nuclear energy policy and to broader OECD analyses in areas such as energy
and the sustainable development of low-carbon economies.

Specific areas of competence of the NEA include the safety and regulation of nuclear activities, radioactive waste
management and decommissioning, radiological protection, nuclear science, economic and technical analyses of
the nuclear fuel cycle, nuclear law and liability, and public information. The NEA Data Bank provides nuclear dataand computer program services for participating countries.

This document, as well as any data and map included herein, are without prejudice to the status of or sovereignty over any territory, to thedelimitation of international frontiers and boundaries and to the name of any territory, city or area.
Corrigenda to OECD publications may be found online at: www.oecd.org/about/publishing/corrigenda.htm.
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Cover photos: PKL facility, Germany (Areva); The PANDA reactor pressure vessel (Paul Scherrer Institute, Switzerland).

FOREWORD

Foreword

At its 61st meeting in June 2017, the Nuclear Energy Agency (NEA) Committee on the Safety of
Nuclear Installations (CSNI) decided to establish a Senior Expert Group on Safety Research(SESAR) to update previous assessments of capabilities and facilities required to support thesafety of nuclear installations. The NEA issued a report on this activity in 2001 entitled Nuclear
Safety Research in OECD Countries: Major Facilities and Programmes at Risk. Six years later, a followon activity resulted in the publication SESAR/SFEAR: Nuclear Safety Research in OECD Countries –Support Facilities for Existing and Advanced Reactors (2007).
The SESAR/SFEAR2’s mandate is as follows: “The Senior Expert Group on Safety Research/Support Facilities for Existing and Advanced Reactors 2 (SESAR/SFEAR2) is responsible forreviewing and updating the previous SESAR assessments of research facilities required tosupport the safety of nuclear installations. The group shall recommend actions to be taken by
the CSNI and its member countries to facilitate broader use and sustained operation of essential
research facilities required to support nuclear safety.”
The activity described in this report builds upon and updates the previous work, expanding
its scope to cover advanced reactors, including evaluations performed by the Task Group on
Advanced Reactor Experimental Facilities (TAREF) and the emergence of several proposed
molten salt and small modular reactor designs. In addition, although the need to maintainexperimental databases was recognised as an important issue, it was not treated specifically inprevious reports; the present report therefore makes some direct recommendations regarding
database maintenance.

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ACKNOWLEDGEMENTS

Acknowledgements

The contributions of the following members of the Senior Group of Experts on Nuclear Safety
Research/Support Facilities for Existing and Advanced Reactors 2 (SESAR/SFEAR2) to review andwrite this report are gratefully acknowledged. In particular, Joanne Ball provided capable
leadership as chair of the group and ensured the timely and efficient completion of the report.

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Fernand Vermeersch

Belgium

Joanne Ball (chair)

CanadaCarlo Vitanza

Czech Republic

Eero Virtanen

FinlandFrancois Barre

FranceMartin Sonnenkalb

Germany

Walter Tromm

Germany

Ivan Tóth

Hungary

Francesco D’Auria

Italy

Koji Kitano

Japan

Taisuke Yonomoto

Japan

Ki-Yong Choi

Korea

Carlos Castelao Lopez

Spain

Inmaculada Córdoba

Spain

Anna Alvestav

SwedenReiner Mailänder

Switzerland

Domenico Paladino

Switzerland

Robert Tregoning

United States

Andrew White (Secretariat)

NEA

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TABLE OF CONTENTS

Table of contents

List of abbreviations and acronyms .................................................................................................... 7
Executive summary ................................................................................................................................ 9
Short-term conclusions and recommendations ........................................................................... 10
Longer-term conclusions and recommendations ......................................................................... 11
General conclusions and recommendations ................................................................................. 12
References .......................................................................................................................................... 12
1. Introduction ....................................................................................................................................... 13
1.1. Purpose ........................................................................................................................................ 13
1.2. Scope ............................................................................................................................................ 14
1.3. Approach ..................................................................................................................................... 15
1.4. Co-ordination .............................................................................................................................. 15
1.5. Organisation of report ............................................................................................................... 15
References .......................................................................................................................................... 17
2. Overview of reactor designs and safety issues ........................................................................... 19
2.1. Boiling water reactors ................................................................................................................ 19
2.2. Pressurised water reactors ........................................................................................................ 222.3. Water-water energetic reactors................................................................................................ 24
2.4. Pressurised heavy water reactors and advanced PHWRs ..................................................... 26
2.5. Advanced light water reactors.................................................................................................. 28
2.6. Advanced pressurised heavy water reactors .......................................................................... 29
2.7. Gas-cooled reactors .................................................................................................................... 30
2.8. Sodium fast reactors .................................................................................................................. 33
2.9. Molten salt reactors ................................................................................................................... 36
2.10. Small modular reactors ........................................................................................................... 39
References .......................................................................................................................................... 39
3. Technical issues and associated facilities.................................................................................... 41
3.1. Issues and facilities unique to the nuclear industry ............................................................. 42
3.2. Severe accident and containment phenomena ..................................................................... 59
3.3. Integrity of equipment and structures .................................................................................... 70
3.4. Issues and facilities not unique to the nuclear industry ...................................................... 74
References .......................................................................................................................................... 88
4. Conclusions and recommendations.............................................................................................. 91
4.1. Introduction ................................................................................................................................ 91
4.2. Short-term conclusions and recommendations .................................................................... 91
4.3. Longer-term conclusions and recommendations .................................................................. 95
4.4. General conclusions and recommendations .......................................................................... 97
Reference ............................................................................................................................................ 98
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TABLE OF CONTENTS

List of tables
1.1.

Impact of SESAR/FAP facility recommendations (2001) on SESAR/SFEAR
findings (2007) ........................................................................................................................ 16
1.2. Impact of SESAR/SFEAR findings (2007) on current facility status .................................. 16
2.1. TAREF gas-cooled reactor summary ratings ...................................................................... 33
2.2. Highest ranked issues and facilities for sodium fast reactors ......................................... 35
3.1. Current thermal-hydraulic safety issues ............................................................................ 46
3.2. Current nuclear fuel issues and facilities ........................................................................... 52
3.3. Fuels and materials reactors worldwide............................................................................. 53
3.4. Current reactor physics issues ............................................................................................. 59
3.5. Severe accident and containment-related issues – past and current ............................. 64
3.6. Current integrity of equipment and structures issues ..................................................... 72
3.7. Current human and organisational factors issues and facilities .................................... 75
3.8. Current plant control and monitoring issues..................................................................... 79
3.9. Cybersecurity research facilities .......................................................................................... 82
3.10. Current seismic effects issues .............................................................................................. 84
3.11 Current fire assessment issues ............................................................................................ 87
4.1. Facilities in danger in the short term (2019-21) and facilities identified in 2007
as being at risk........................................................................................................................ 92
4.2. Critical facilities to be monitored in the long term ........................................................... 96

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LIST OF ABBREVIATIONS AND ACRONYMS

List of abbreviations and acronyms

CEA

Alternative Energies and Atomic Energy Commission (France)

CFD

Computational fluid dynamics

CFP

Coated fuel particle

CHF

Critical heat flux

CNL

Canadian Nuclear Laboratories

CSNI

Committee on the Safety of Nuclear Installations (NEA)

CTF

Containment Test Facility (Canada)

D-LOFC

Depressurised loss-of-coolant

DOE

Department of Energy (United States)

ECCS

Emergency core cooling system

FBTR

Fast Breeder Test Reactor (India)

FP

Fission product

GFR

Gas fast reactor

HFRHigh Flux Reactor

HTGR

High-temperature gas reactor

HTR

High-temperature reactor

I&C

Instrumentation and control

IAEA

International Atomic Energy Agency

IFE

Institute for Energy Technology (Norway)

IRSN

Institute for Radiological Protection and Nuclear Safety (France)

JAEA

Japan Atomic Energy Agency

JMTR

Japan Material Testing Reactor

LOCA

Loss-of-coolant accident

LOFC

Loss-of-forced circulation

LSCF

Large-Scale Containment Facility (Canada)

LSTF

Large Scale Test Facility (Japan)

LWR

Light water reactor

MCCI

Molten core concrete interaction

MOX

Mixed oxide

MSFR

Molten salt fast reactor

MSR

Molten salt reactor

MSRE

Molten Salt Reactor Experiment

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LIST OF ABBREVIATIONS AND ACRONYMS

8NC

Natural circulation

NEA

Nuclear Energy Agency

NRU

National Research Universal (Canada)

NSC

Nuclear Science Committee (NEA)

NSRR

Nuclear Safety Research Reactor (Japan)

P-LOFC

Pressurised loss-of-coolant

PHWR

Pressurised heavy water reactor

PSI

Paul Scherrer Institute (Switzerland)

PWR

Pressurised water reactor

RIA

Reactivity insertion accident

RPV

Reactor pressure vesselSA

Severe accident

SARNET

Severe Accident Research Network

SBO

Station blackout

SESAR

Senior Expert Group on Safety Research (NEA)SETH

SESAR Thermal-hydraulics (NEA)SFEAR

Support Facilities for Existing and Advanced Reactors (NEA)

SFP

Spent fuel pool

SFR

Sodium fast reactor

SG

Steam generator

SGTR

Steam generator tube rupture

SiC

Silicon carbide

SMR

Small modular reactor

SSC

Systems, structures and components

STEM

Source Term Evaluation and Mitigation (NEA)

TAREF

Task Group on Advanced Reactor Experimental Facilities (NEA)

TREAT

Transient Reactor Test Facility (United States)

VENUS

Vulcan Experimental Nuclear Study (Belgium)

VVER

Water-water energetic reactor

WGAMA

Working Group on Analysis and Management of Accidents (NEA)

WGFS

Working Group on Fuel Safety (NEA)

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EXECUTIVE SUMMARY

Executive summary

The OECD Nuclear Energy Agency (NEA) Committee on the Safety of Nuclear Installations (CSNI)
decided in June 2017 to establish a Senior Expert Group on Safety Research (SESAR) to update
previous assessments of capabilities and facilities required to support the safety of nuclear
installations. The NEA issued a report on this topic in 2001 entitled Nuclear Safety Research inOECD Countries: Major Facilities and Programmes at Risk (NEA, 2001). In 2007, a follow-on activity
resulted in the publication of a second report: SESAR/SFEAR: Nuclear Safety Research in OECDCountries – Support Facilities for Existing and Advanced Reactors (NEA, 2007).
This report builds on and updates the previous work, expanding its scope to cover additional
advanced reactor types, including evaluations performed by the Task Group on Advanced Reactor
Experimental Facilities (TAREF) for high-temperature reactors and sodium fast reactors. It also
identifies general safety issues for several proposed molten salt and small modular reactor designs.
The present report – Nuclear Safety Research Support Facilities for Existing and Advanced Reactors: 2021
Update – can be considered as a follow-on to the previous two reports. In addition, although theneed to maintain experimental databases was recognised as an important issue, it was not treatedspecifically in previous reports; the present report makes some direct recommendations regarding
experimental data preservation.
Several facilities have been shut down since the publication of the SESAR/SFEAR reportin 2007. Accordingly, loss of critical research infrastructure (i.e. facilities, capabilities andexpertise) remains a concern and was a major factor in conducting the current study. However,
several key facilities were preserved between 2007 and 2019 thanks to the SESAR/SFEAR project.
These are discussed in Chapters 1 and 4.
The present report discusses the safety issues, research needs and supporting researchfacilities associated with currently operating water-cooled reactors in NEA member countries.
These reactors include pressurised water reactors (PWRs), boiling water reactors (BWRs),
pressurised heavy water reactors, water-water energetic reactors, and advanced or evolutionary
designs of these types. The main purpose of this report is to:•summarise current safety issues, whose resolution will depend on additional research
work;

•

provide the current status of research facilities that support resolution of the safety issues;•recommend actions the CSNI could take in the short term to help maintain facilities which
require a substantial investment of resources and are in danger of premature closure;•provide recommendations on long-term nuclear safety research facility infrastructureneeds and preservation.

The report also provides information on safety issues and research needs not unique to the
nuclear industry and those associated with advanced reactor designs such as gas-cooled reactors,
sodium fast reactors, molten salt reactors and small modular reactors. This information ispresented for completeness and for use by designers, operators and researchers in planning andconducting future work.
This report addresses technical issues unique to the nuclear industry as well as those thatare relevant to the nuclear industry as well as other industries.
•

Issues unique to the nuclear industry:–thermal-hydraulics;–

nuclear fuel;

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EXECUTIVE SUMMARY•–

reactor physics;–

severe accident and containment phenomena;–

integrity of equipment and structures.

Issues not unique to the nuclear industry:–human and organisational factors;–

plant control and monitoring;–

cybersecurity;–

external events;–

fire assessment.

The second group of issues includes factors that are relevant to the nuclear industry but arealso supported in a major way by other industries and thus may not require support from the
CSNI. In general, in developing recommendations for consideration, the group focused on thosefacilities that have unique capabilities and high relevance to the resolution of safety issues for
Generation II designs as well as the potential to be highly relevant in support of the resolution ofsafety issues for new and emerging Generation III and IV designs. The uniqueness, replacement
cost (>EUR 5 million) and operating cost of facilities was an important consideration in both theevaluation and recommendations. Due to the costs of operating many larger facilities, further
co-operative efforts will most likely be needed to maintain them in the longer term.

Short-term conclusions and recommendations
To assess short-term concerns, the facilities were examined (see Chapter 3), and those that wereunique, versatile and in danger of being shut down in the period 2021-2024. The CSNI is
encouraged to support, to the extent possible, joint projects proposed for these facilities.
In the thermal-hydraulics area, three facilities are in short-term danger. These include PKL
and Large Scale Test Facility (LSTF, Japan), which support PWR thermal-hydraulic work, and
PANDA, which supports both light water reactors (LWR) and BWR thermal-hydraulic safetyissues, as well as severe accident and containment phenomena. It should be noted that both
PANDA and PKL were identified as being at short-term risk as early as 2001, but have been used
to conduct several important CSNI research initiatives via NEA joint projects in the intervening
years. Both PANDA (PSI, Switzerland) and PKL (Areva, Germany) proponents are anticipated topropose new projects before the current projects come to an end (2021 for PANDA and 2020 for
PKL). The short-term prognosis of the LSTF (Japan) is similar: indications are that Japan will
likely propose a new joint project in the near future.
In the severe accident and containment phenomena technical area, the aforementioned
PANDA facility, the VERDON facility (France) and the THAI facility (Germany), identified as beingat risk in the near term, had all been identified as being at risk, or to require long-termmonitoring, in the 2007 report. There are future NEA joint projects planned for each of them inorder to justify their operation for the coming years. The CHROMIA platform (fission productrelease and transport facilities at the Institut de Radioprotection et de Sûreté Nucléaire [IRSN]
in France) is also at risk, and, like the THAI facility, has been extensively utilised by NEA projects(STEM 1, 2 and 3) in the 2010s. A new NEA joint project, ESTER, is now underway and will use
both the CHROMIA and the VERDON facilities. Another new NEA joint project, THEMIS, is also
underway and will use the THAI facility.
In the fuel and materials areas, several large, versatile reactors have been closed since 2007.
The National Research Universal Reactor in Canada and the Halden Reactor at IFE Norway bothclosed in 2018, representing a significant loss worldwide, for both materials and fuel testing.
Additionally, the Japan Material Testing Reactor (JMTR) in Japan, a test reactor used primarily
for materials testing, has been shut down. Despite these facilities being previously identified as
being at short-term risk (PHEBUS) or facilities to monitor in the long term, substantial national
support is required to sustain or refurbish such ageing or specialised reactors.

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EXECUTIVE SUMMARY

Reactors worldwide, including CABRI, BR-2, LVR-15, MIR, TREAT, HFR, ATR and others havebeen identified as being suitable to replace some of the capabilities lost by the closure of thesereactors. The NEA has recently taken some measures to protect existing infrastructure byproposing the establishment of the FIDES network, which, along with joint projects for a varietyof irradiation activities, will allow members to access various reactors and test programmes
for materials and fuels. This step towards collaborative irradiation projects will provide both
short- and medium-term activities that can be supported by the CSNI to mitigate the current
irradiation gap.
A review of sub-critical and zero-power reactors by the NEA Nuclear Science Committee foundthat several of these reactors have been closed since the last SESAR report, and recommended
that they should closely monitor the remainder (VENUS in Belgium, ZED-2 in Canada, LR-0 in the
Czech Republic and KUCA in Japan). The STACEY facility (Japan) was mentioned as a candidate to
replace some capacity, but it is currently shut down. The CROCUS facility in Switzerland may also
have some capacity. None of the aforementioned facilities have been identified as being in
imminent danger of shutdown.
Finally, although the scope of the 2007 activity focused solely on facilities unique to the
nuclear industry, the GALAXIE fire platform at the IRSN is identified in this evaluation as beingat risk in the near term. Although other fire research facilities exist, the co-location of separate
effects, intermediate and large-scale facilities in this platform, as well as the experience gained
from three phases of NEA joint projects such as PRISME, make GALAXIE ideal for addressing fire
safety issues unique to the nuclear industry (e.g. glovebox fires, propagation through ventilation
systems, fire initiated in cable trays). An NEA joint project is planned to be proposed once thecurrent PRISME project ends (in 2021).

Longer-term conclusions and recommendations
Many of the factors used in the last two reports to arrive at conclusions and recommendations
have resulted in effective measures for retaining key facilities at risk. These measures should
continue to be used in the future, with consideration of the following factors:•

Cost of facility operation and replacement (i.e. limit CSNI involvement to large facilities
needing multinational support).

•

Consistency with SFEAR recommended list of facilities for long-term preservation (seebelow).

•

Ability to define a useful experimental programme (i.e. one that will provide information
useful to the resolution of one or more safety issues).

•

Long-term planning to ensure that the most important facilities receive the highest priority
for long-term preservation (i.e. not first come, first served). This would include assessing
the long-term resource implications (i.e. consider impact of cost of a co-operative
programme on resources available for other projects) and the host country’s long-termplans for the facility.

•

Industry participation.

•

Host country commitment and ability to support NEA projects.

Building on previous SESAR evaluations and the safety issues contained in Section 3.1 of thisreport, a table of critical research facility infrastructure needs was developed along with a list, byreactor type, of existing facilities that could fulfil those needs. This list is discussed in Chapter 4.
The facilities are those considered unique, hard to replace and identified as playing a significant
role in resolving issues in their technical area. It is recommended that the CSNI continue to focus
on these facilities in developing a strategy for long-term infrastructure preservation, continuing
to take action, as appropriate, to ensure that critical facilities are available for each reactor type tomeet the critical research infrastructure needs. The same should be done for facilities developed
for new reactor types identified in Chapter 2. Similar to the short-term recommendations above,
host country commitment and long-term facility plans will be important factors in determining
which facilities to preserve.

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EXECUTIVE SUMMARY

General conclusions and recommendations
The following conclusions and recommendations pertain to both the short and long term. They
result from the group’s observations and experience in carrying out the current activity and desire
to develop a practical set of recommendations, with facility preservation being a co-ordinatedeffort among the NEA standing committees. Specific general conclusions and recommendations
are listed below. It is worth noting that many of the conclusions reached from previous
SESAR/SFEAR activities are still valid, and are included below.
•

CSNI members are encouraged to continue their excellent support for facilities at risk,
which has already resulted in several valuable projects (e.g. PKL, PANDA, GALAXIE).

•As recommended in the previous report, test reactor availability should be given special
scrutiny, due to the high cost of operation and replacement. The new Framework for
Irradiation Experiments (FIDES) framework and associated Joint Experimental Programmes(JEEPs) is an essential step in maintaining global capability.

•

Regardless of FIDES and the success of JEEP, continued and ongoing attention must be
focused on smaller unique facilities at risk.

•

The Nuclear Science Committee (NSC) should maintain a close watch on facilities used
to support criticality and reactor physics codes.

The shutdown of the Halden Reactor and subsequent project activities have highlighted theneed for preserving key experiments in international databases. Consequently, the groupdeveloped a series of recommendations specifically targeting data preservation:•

NEA Joint Safety Research projects should clearly outline their plan for data preservation,and should stipulate that a copy of the primary data needs to be sent to the NEA for storage.

•

CSNI working groups should be asked to identify key datasets in their areas. Some of thismay have been done with code validation matrices and datasets to support thedevelopment and implementation of standards.

•

There should be a cross-functional (CSNI, NSC, etc.) NEA task group established toconsider what should be done to preserve the key experimental datasets. This could
include possible options for data libraries, how to screen datasets, what informationneeds to accompany the primary data, etc.

•

CSNI working groups should select an appropriate option for preserving each key dataset
and develop an activity to put it in place (CSNI Activity Proposal Sheet [CAPS], joint
project, etc.).

References
NEA (2001), Nuclear Safety Research in OECD Countries: Major Facilities and Programmes at Risk, OECD
Publishing, Paris, www.oecd-nea.org/SESAR.
NEA (2007), “Nuclear Safety Research in OECD Countries: Support Facilities for Existing andAdvanced Reactors (SFEAR)”, OECD Publishing, Paris, www.oecd-nea.org/SESAR-SFEAR.

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1. INTRODUCTION

1. Introduction

The Nuclear Energy Agency (NEA) Committee on the Safety of Nuclear Installations (CSNI) has
used Senior Expert Groups on Safety Research (SESARs) to identify capabilities and facilities
essential for the safety of nuclear installations, and to recommend how to preserve key
capabilities and facilities. The CSNI has acted on SESAR recommendations and established cooperative research projects that focused research on areas where there were significant safetyissues to be addressed, and that preserved key facilities which were in imminent danger of being
shut down.
Tables 1.1 and 1.2 summarise the recommendations from the previous Senior Expert Groups,
identify actions taken by the NEA as a result, and provide the status of facilities previously
identified as being at risk or requiring monitoring.
This report describes the outcomes from the latest SESAR group established by the CSNI. The
Senior Expert Group on Safety Research/Support Facilities for Existing and Advanced Reactors 2(SESAR/SFEAR2) was asked to review and update the previous SESAR assessments. They were also
asked to recommend actions to be taken by the CSNI and its member countries to facilitate
broader use and sustained operation of research facilities essential to support nuclear safety.
The recommendations of the Senior Expert Group on Safety Research/Support Facilities for
Existing and Advanced Reactors 2 (SESAR/SFEAR2) have prompted numerous co-operative
research projects to preserve key facilities that were in near-term danger and to improve thestate of knowledge in selected areas. The actions taken by the CSNI as a result of two previous
evaluations by the expert group, in 2001 and 2007, have kept a number of facilities active thatwould otherwise have been lost, and focused research on areas where there were significant
safety issues that remained to be addressed.
Tables 1.1 and 1.2 summarise the recommendations from the previous Senior Expert Groups,
identify actions taken by the NEA as a result, and provide the status of facilities previously
identified as being at risk or requiring monitoring.

1.1. Purpose
The SESAR/SFEAR2 mandate defines the purpose as follows: “In conducting its review,
SESAR/SFEAR2 should ensure that the facilities in all NEA member countries are considered, either
through direct representation on the group, or through opportunities for non-group members toprovide input. To the extent possible, consideration should also be given to infrastructure of othermajor nuclear countries like China and India. SESAR/SFEAR2 should interface with groups within
the NEA that conduct similar reviews, with the International Atomic Energy Agency and withother international organisations as necessary.”
For operating reactors, the purpose of the SESAR/SFEAR activity is to: summarise the currently
identified safety issues whose resolution depends on additional research work; provide thecurrent status of those research facilities unique to the nuclear industry that support resolutionof the safety issues; where such facilities represent a substantial investment of resources and arein danger of premature closure, recommend actions the CSNI could take to help maintain thefacilities; and provide recommendations on long-term nuclear safety research facility
infrastructure needs and a strategy for its preservation. Note that many facilities, such as hot cells
and test loops, that are owned by a commercial entity and primarily used to support nuclear
vendors or operators are not typically included in this report, except where they provide uniqueinfrastructure to resolve ongoing safety concerns. In addition, where research facilities do not

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1. INTRODUCTION

exist, but would be useful to address currently identified safety issues, facility needs are identified.
For future reactors, the purpose of the SESAR/SFEAR activity is to summarise known safety issues,
safety research and facility needs for use by designers, operators and researchers.

1.2. Scope
The mandate given to the SESAR/SFEAR2 defines the scope as follows: “The scope for the SESAR
review should first focus on the support required for water cooled reactors (boiling water reactor[BWR], pressurised water reactor [PWR], pressurised heavy water reactor [PHWR]), particularly
those currently deployed in NEA member countries. Nevertheless, recognising the increased
importance of innovative reactor designs, requirements for non-water cooled reactors should
be included with generic considerations in the various technical areas (e.g. materials, fuel,severe accidents, etc.).”The scope of this activity is primarily limited to safety issues and facilities associated with
nuclear reactor design, construction and operation. However, in contrast to the previous reports,
the safety of spent fuel pools has also been added under the sub-chapter of nuclear fuel, given theevents at Fukushima. In addition, the current report adds cybersecurity as an issue warranting aseparate section because it has emerged as an area of concern for operating plants. Finally, theprevious focus on seismic effects has been expanded to include other external events. The reportfocuses on the following technical areas:•

thermal-hydraulics;•

nuclear fuel;•reactor physics;•

severe accidents and containment phenomena;•

integrity of equipment and structures;•

human and organisational factors;•

plant control and monitoring;•

cybersecurity;•

external events;•

fire assessment.

The first five address phenomena, safety issues and facilities unique to the nuclear industry.
The last five address phenomena, safety issues and facilities that are relevant to the nuclear
industry, but which are also relevant to other industries and where research and facilities may
be supported by others. The safety issues identified in this report are those where additional
research is needed to support their resolution or to reduce uncertainties, thus supporting more
realistic treatment of the issue.
The reactor designs to be assessed in each of the technical areas are:•

Currently operating PWRs, BWRs, PHWRs and water-water energetic reactors (VVERs) inmember countries.

•

Future designs including advanced light water reactors (LWRs), PHWRs and gas-cooled
reactors, as well as sodium fast reactors, molten salt reactors and finally, small modular
reactors that could belong to any of the aforementioned types.

It should also be recognised that research in some of the technical areas listed above does
not require large facilities, but rather large-scale co-ordinated programmes.

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1. INTRODUCTION

1.3. Approach
A set of recommendations was developed for each of the technical areas above that are uniqueto the nuclear industry. No technical recommendations are provided for those technical areas
previously identified as not unique to the nuclear industry, but the safety issue and facility
information is provided for information only. The exception is for fire research, for which it wasrecognised that there is a need for facilities that can evaluate fire propagation and behaviour inconfined spaces (ventilation, cable trays, gloveboxes) unique to nuclear facilities, and that there
is a very versatile facility in France that should be maintained.
For the issues identified for operating and future plants, members re-evaluated the safetyissues described in the previous SESAR/SFEAR report and identified research work that was stillrequired. Further, the major facilities currently performing or available to perform research
directly relevant to these issues have been identified. Those that were operable in the early2010s but which have since been closed, or are due to be closed soon, are also specified, in orderto provide an overview of recent developments.

1.4. Co-ordination
The issues, conclusions and recommendations contained in this report represent the personalviews of the SESAR/SFEAR participants, but not necessarily the views of their organisations.
However, in assembling the information contained in the report, the participants benefited from
input from the CSNI working groups, special expert groups and CSNI members. In addition,
input was obtained from the NEA Nuclear Science Committee on the reactor physics section.

1.5. Organisation of report
Chapter 2 provides a short overview of the reactor designs and their safety issues addressed
within the scope of this report. Chapter 3 addresses each of the technical areas listed above for
operating reactors as well as advanced ones. It is split into two sections pertaining to thosetechnical areas unique to the nuclear industry, and those that are not. It also describes in moredetail the organisation and purpose of both of these sections. Chapter 4 provides the group’s
conclusions and recommendations regarding critical facilities unique to nuclear safety researchin danger of being lost that deserve and need international support and possible actions for
CSNI consideration.
The previous report provided a detailed and systematic ranking of the importance of eachsafety issue as well as a ranking of facilities on the basis of their importance to resolving theissue, their uniqueness, their versatility to address more than one issue, and the cost of theiroperation or replacement. The same general approach was followed here; however, since theissues faced today are very similar to those faced in the early 2010s, it was unnecessary to rank
the issues in terms of their safety significance. Furthermore, except where identified in the text,
it is assumed that all of the issues identified in the tables have a medium to high safety
relevance, and that additional research is required to resolve the issue.
The conclusions and recommendations are divided into near-term (CSNI action needed in
the next one to three years) and long-term actions. However, any action by the CSNI would becontingent upon the willingness of the host country of the facility to contribute substantially to
its continued operation in accordance with NEA/CSNI guidelines (i.e. the CSNI does not intend
to serve as a host country for facility preservation).

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1. INTRODUCTION

Table 1.1. Impact of SESAR/FAP facility recommendations (2001)
on SESAR/SFEAR findings (2007)
SESAR/FAP recommendation

Resulting CSNI action

1) Maintain the PANDA, PKL and SPES
facilities in the thermal-hydraulic area (thesefacilities were in near-term danger ofclosure).

Initiated the SETH programme
PANDA maintained through 2005, but identified
utilising the PANDA and PKL facilities as being in near-term danger in the 2007 SFEAR(no host country support for SPES).
study.
PKL active and not in near-term danger.

2) Monitor and maintain key thermalhydraulic facilities in the long term. Thermalhydraulic facilities should be maintained in
North America, Europe and Asia.

Facility status monitored. Initiatedprogramme utilising the ROSA facility
when it was in danger of being shut
down.

3) Maintain the RASPLAV and MACE facilities
in the severe accident area (these facilities
were in near-term danger of closure).

Initiated the MASCA programme as a MASCA and MCCI remained active were and not
follow-on to RASPLAV to maintainin near-term danger in 2007.
facilities. Initiated the MCCI
programme utilising the MACE facility.
Initiated the SERENA programme
KROTOS facility maintained through 2007.
(group of experts to discuss status offuel coolant interaction and future
experimental needs). FARO
shutdown. KROTOS kept in standby.

4) Develop centre of excellence on fuelcoolant interaction in consideration ofpotential loss of the FARO and KROTOS
facilities.
5) Develop centre of excellence on iodine
chemistry and fission product behaviour.

Impact on facilities until 2007

ROSA remained active and was not deemed as
being in near-term danger in 2007. Other
thermal-hydraulic facilities continued to bemonitored.

Joint projects planned in Canada and New facilities developed in Canada and France
France respectively.
to support NEA joint projects, Behaviour ofiodine and STEM projects operating by 2007-08.

Table 1.2. Impact of SESAR/SFEAR findings (2007) on current facility status
SESAR/SFEAR

Resulting CSNI action

1) Maintain the PANDA facility in the NEA joint project SETH-2 was planned.
thermal-hydraulics andcontainment areas in the near term.
2) Monitor the LSTF/ROSA, PACTEL,
RD-14M, ATLAS, PKL and PSB-VVER
thermal-hydraulics facilities in thelong term.

Maintained facilities addressing issues ofbroad international interest through theestablishment of international projects.
Some of these projects have been
completed (SETH, PSB-VVER, ROSA-1 and2), while others are still ongoing (PKL-1 to
4, ATLAS-1 and 2). Some PKL tests werecomplemented by experiments performedin the PMK, ROCOM and (more recently)
PACTEL facilities.

3) In the fuel area, PHEBUS was
Supported (and continued to support) the
identified as being at short-term risk CABRI International Project. The Halden
while Halden, NSRR, CABRI, NRU and project will be supported until 2020.
MIR were identified as facilities tomonitor in the long term.
4) Numerous containment andInitiated joint projects involving QUENCH,
severe accident facilities were
VERDON, CHROMIA and THAI facilities.
identified as being at short-term
risk, requiring long-termmonitoring. These included CTF and
LSCF in Canada, THAI, QUENCH andCOMET facilities in Germany,
CHROMIA platform and VERDON
facilities in France, and the MCCI
facility in the United States.

16

Impact on facilities since 2007
Facility has remained operational, with asubsequent project, HYMERES-1, completed and
HYMERES-2 underway. New joint project will beplanned before HYMERES-2 is completed in 2021.
RD-14M has closed following a national decision.
The other facilities remain open. PKL and LSTF/
ROSA are currently identified to be at short-term
risk, with proposals for NEA joint projects expected
to be proposed in the near future.

National decisions resulted in PHEBUS, Halden and
NRU being shut down. CABRI has been the subjectof an NEA joint project since 2000. Both CABRI andMIR are planning joint projects under the new NEA
FIDES network. The NSRR is available and not
identified as being at risk.
CTF, LSCF and COMET facilities closed based onnational decisions. VERDON, CHROMIA, QUENCH
and THAI facilities continue to be supported bynational programmes and NEA joint projects. All
are currently identified as being at short-term riskor requiring monitoring in the long term and are
the subject of new NEA joint projects (THEMIS for
the THAI facility, QUENCH-ATF for the QUENCH,ESTER for the CHROMIA facility). For the MCCI
facility, a new joint project, ROSAU, makes use ofsome of the expertise and components employedby the MCCI, although a new facility will be used.

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1. INTRODUCTION

References
NEA (2001), Nuclear Safety Research in OECD Countries: Major Facilities and Programmes at Risk, OECD
Publishing, Paris, www.oecd-nea.org/SESAR.
NEA (2007), “Nuclear Safety Research in OECD Countries: Support Facilities for Existing andAdvanced Reactors (SFEAR)”, NEA/CSNI/R(2007)6, OECD Publishing, Paris, www.oecd-nea.org/
SESAR-SFEAR.

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2. OVERVIEW OF REACTOR DESIGNS AND SAFETY ISSUES

2. Overview of reactor designs and safety issues

This section contains short overviews of each of the major reactor types within the scope of thisreport. These overviews are for the purpose of familiarising the reader with the basic
characteristics of the designs as well as the major safety features and issues. The reactor designs
included in this section are:•boiling water reactors (BWRs);•

pressurised water reactors (PWRs);•

water-water energetic reactors (VVERs);•

pressurised heavy water reactors (PHWRs) and advanced PHWRs (APHWRs);•

advanced light water reactors (ALWRs);•

gas-cooled reactors;•

sodium fast reactors (SFRs);•

molten salt reactors (MSRs);•

small modular reactors (SMRs).

2.1. Boiling water reactors
2.1.1. Introduction
The direct-cycle BWR nuclear system is a steam generation and steam utilisation system
consisting of a nuclear core located inside a reactor pressure vessel and a conventional turbinegenerator and feedwater supply system. Associated with the nuclear core are auxiliary systems
to accommodate the operational and safety requirements and necessary controls andinstrumentation. Water is circulated through the reactor core, producing saturated steam which
is separated from the recirculating water, dried in the top of the vessel, and directed to thesteam turbine-generator. The turbine employs a conventional regenerative cycle withcondenser deaeration and condensate demineralisation.
The steam produced by the nuclear core is radioactive. The radioactivity is primarily due to
N-16, a very short-lived isotope (7 second half-life) so that the radioactivity of the steam fromthe reactor vessel is mostly present during power generation. However, other radioactive
material (e.g. from fuel cladding failures) can also be entrained in the steam, increasing its
radioactivity and primary coolant system contamination.
BWR core designs and containment designs can vary depending on the age of the design
and product type. However, they all have some common characteristics:•

The almost universal use of recirculation inside the reactor vessel (using either jet pumps
or centrifugal pumps) to increase water flow through the core (and thus control boilingand power level).

•

The use of a pressure suppression type containment, whereby steam released from thereactor coolant system would be condensed by being directed to a pool of water (called
the pressure suppression pool), thus allowing a smaller containment building.

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2. OVERVIEW OF REACTOR DESIGNS AND SAFETY ISSUES•The use of control rod drives that enter through the bottom of the reactor vessel, thus
allowing for installation of steam dryer and moisture separator equipment above thecore outlet. Such a control rod location also allows removal of the upper reactor vessel
head for refuelling without disturbing the control rod drives.

•

Automatic depressurisation capability for the reactor coolant system to allow water
injection at low pressure.

•

No use of boron in the primary coolant in normal operation.

The next section describes the main design features and safety issues associated with BWRs.

2.1.2. Boiling water reactor design features
The nuclear core consists of fuel assemblies and control blades contained within the reactor
pressure vessel and cooled by a recirculating water system. The fuel elements of an assembly
are located within a box made of Zircaloy and the cross-shaped control blades are located within
the gap between fuel assembly boxes. The recirculating water system consists of the feedwaterflow and flow internal to the reactor vessel (recirculation flow), which is the result of pumps
which increase flow internal to the reactor vessel. These pumps generate about two thirds ofthe flow within the reactor vessel. The pumps can be mechanical or jet pumps internal to thereactor vessel. BWR power level is maintained or adjusted by positioning control rods up anddown within the core. The BWR core power level is further adjustable by changing therecirculation flow rate through the core without changing control rod position. This feature
helps achieve load-following capability for the BWR.
The BWR employs bottom-entry control rods and bottom-mounted control rod drives,
which allow refuelling without removal of control rods and drives, and allow drive testing with
an open vessel prior to initial fuel loading or at each refuelling operation.
BWRs operate at constant steam pressure (approximately 70 bars or 1 000 pounds per
square inch) at the corresponding saturation temperature. They employ moisture separatorsand steam dryers to enhance the quality of the steam entering the turbine.
The BWR reactor is housed in a reactor building structure, which includes two main
structures: the shield building and the containment system. The containment portion of thereactor building is divided into two main compartments called the drywell and wetwell.
Components located within the drywell include, but are not limited to, the reactor vessel, thereactor water recirculation system, the main steam lines, the main steam line safety/relief
valves and discharge piping, control rod drives and piping, piping and valves associated withthe reactor vessel, nuclear system instrumentation, and heating and ventilation. Components
located outside the drywell, but inside the containment vessel include, but are not limited to,the control rod drive hydraulic modules, standby liquid control system components, reactor
water cleanup system heat exchangers, auxiliary system piping, refuelling bridge, polar crane,
nuclear systems instrumentation heating and ventilating, and the pressure suppression pool.
The containment is a steel leakage barrier which prevents significant fission productrelease to the outer shield building in the event of an accident. The containment, including all
penetrations and welded attachments, acts as an independent structural component within thereactor building for the maximum temperature and pressure conditions that can occur as theresult of a loss-of-coolant accident (LOCA), and accommodates reactor blowdown through thesafety/relief valve discharge piping to the suppression pool.
The containment is subdivided into drywell and wetwell or suppression pool. The
suppression pool is a toroidal pool of demineralised water within the containment boundaryand connected to the drywell by condensation pipes. The suppression pool provides: a means
to condense any steam released in the drywell area during a LOCA; a heat sink for the reactor
core isolation cooling system during hot standby operation until the decay heat can be removedby the residual heat removal heat exchangers; a heat sink for venting the nuclear systemsafety/relief valves; a source of water for emergency core cooling; and a source of water to thecontainment spray system.

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2. OVERVIEW OF REACTOR DESIGNS AND SAFETY ISSUESSurrounding the containment is the shield building. The shield building is often a cylindricalshell of reinforced concrete. It completely encloses the reference free-standing containment. The
primary function of the shield building, the “secondary fission product barrier”, is to further limit
nuclear radiation to the environment in the event of an accident involving the release of fission
products. The structure also protects the containment from adverse atmospheric conditions andexternal threats, such as missiles. It contains the spent fuel pool.
The annulus between the shield building and the containment provides a plenum for thecollection and filtration of fission product leakage from the containment that may occur
following a design-basis accident. The annulus is normally kept at a negative pressure relative
to atmospheric pressure so any leakage through the shield building or containment is into thisspace. Under accident conditions, the ventilation exhaust from this space is automaticallydiverted through the filtered standby gas treatment system before release to the environment.
A number of safety systems are provided on BWRs to respond to loss-of-coolant, loss-ofpower and reactivity insertion events. These consist of:•a fast-acting SCRAM system;•a standby liquid control system for emergency boron injection into the core in the eventof a failure to SCRAM;•

high- and low-pressure coolant injection systems in the event of a LOCA;•

primary coolant system depressurisation capability;•

backup power supplies.

2.1.3. Boiling water reactor safety issues
Design, operation and research associated with BWRs since the mid-1950s have generated
information used to address many safety issues. These issues include:•

anticipated transient without SCRAM;•

shroud, feedwater nozzle and recirculation pipe cracking;•station blackout;•core spray distribution;•

suppression pool dynamics;•

stability;•

severe accident concerns:–

in-vessel melt progression and melt retention;–

in- and ex-vessel fuel coolant interaction;–molten core concrete interaction (MCCI)/containment shell melt through (e.g. Mark I
containment);–

ex-vessel core debris coolability;–

containment cooling/integrity/venting;–combustible gas control;–

source term.

However, these and other issues have been re-evaluated due to the development of new
designs, operating experience feedback (lessons learnt) or industry initiatives to extend plant
lifetime, raise power levels, increase fuel burnup levels or increase operating cycle length. Theseissues include:•

plant ageing, including materials cracking/corrosion;•

power uprates;•

high-burnup/MOX fuel;

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2. OVERVIEW OF REACTOR DESIGNS AND SAFETY ISSUES•

installation of digital instrumentation and control (I&C);•

emergency core cooling system (ECCS) sump strainer clogging;•stability;•

impact on human actions and reliability;•

flow-induced vibration;•

cracking of reactor internals.

These issues have been, or are being, resolved through generic or plant-specific reviews withsignificant input from safety research programmes. Some of the above issues are unique to BWRs;
others are shared with PWRs. Accident management programmes have been developed to preventor mitigate many of the severe accident concerns mentioned.
Although many of the basic safety issues are the same, research to address the above issues
may be different due to the differences in design and operation between BWRs and PWRs.
Accordingly, these differences need to be considered when assessing facility and programmeneeds.
Recently, more emphasis has been placed on risk-informed regulation in some countries
and in more detailed assessments of accident scenarios and accident management actions. This
has led to new issues research needs, such as:•

redefinition of large-break LOCA;•

break location and orientation.

2.2. Pressurised water reactors
2.2.1. Introduction
The pressurised water reactor (PWR) consists of a high-pressure reactor vessel (about 140 bar or
2 200 pounds per square inch) with anywhere from 2 to 4 coolant loops. Each coolant loop has areactor coolant pump and a steam generator where heat is transferred from the reactor coolant
to generate steam in separate secondary loops. The secondary loops carry steam from the steam
generators to the turbine and also pump feedwater back into the steam generators. The steam
generators produce saturated steam at approximately 70 bar or 1 000 pounds per square inch.
PWR core and plant designs can vary depending on the product type (2-, 3- or 4-loop plant).
However, they all have some common characteristics:•

The use of large dry containment buildings, although there are some PWRs with subatmospheric containment buildings, ice-condenser pressure suppression containment
systems or double wall designs with a partial vacuum in the annulus.

•

The use of soluble boron in the reactor coolant to help control reactivity and achieve cold
shutdown.

•

The use of control rod drives that enter through the top reactor pressure vessel (RPV) head.

•

The use of an electrically heated pressuriser to maintain reactor coolant pressure.

The next section describes the main design features and safety issues associated with PWRs.

2.2.2. Pressurised water reactor design features
The reactor core is of the multi-region type. All fuel assemblies are mechanically identical,
although the design (e.g. grids, nozzles) and the fuel enrichment is not the same in all of theassemblies. All fuel assemblies are without assembly boxes. In the typical initial core loading,
three fuel enrichments are generally used. Fuel assemblies with the highest enrichments areplaced in the core periphery, or outer region, and the two groups of lower enrichment fuelassemblies are arranged in a selected pattern in the central region. In subsequent refuellings,

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one part of the fuel is discharged (generally one-fourth to one-third) and fresh fuel is loaded
into the outer region of the core. The remaining fuel is arranged in the central part of the corein such a manner as to achieve optimum power distribution.
High-pressure water circulates through the reactor core to remove the heat generated in
the fuel. The heated water exits from the reactor vessel and passes via the coolant loop piping
to the steam generators. There it gives up its heat to the secondary coolant (feedwater) togenerate steam for the turbine generator. The primary coolant loop is closed when the water ispumped back to the reactor vessel. The entire reactor coolant system is composed of leak-tightcomponents to ensure that all radioactivity is confined to the system.
The reactor system containment building is usually a reinforced concrete or steel shellpressure vessel. The contained volume and design pressure of the vessel are sufficient towithstand and contain the contents of the reactor coolant system in the unlikely event of a
LOCA or a main steam line break. The containment building houses the reactor and reactor
coolant system, including the steam generators, reactor coolant pumps, pressuriser, piping andthe safety injection equipment. The reactor coolant system is arranged with the reactor vessel
adjacent to and below the fuel transfer canal to permit complete underwater fuel handling. The
fuel storage building located immediately adjacent to the containment building has underwater
fuel storage facilities, which are connected to the containment refuelling canal by a fuel transfer
tube and a mechanised fuel transfer dolly.
The reactivity of the reactor is controlled by the temperature coefficient of reactivity; bycontrol rod cluster motion, which is required to follow load transients and for startup andshutdown; and by a soluble neutron absorber, boron, in the form of boric acid, which is inserted
during cold shutdown, partially removed at startup, and adjusted in concentration during core
lifetime to compensate for such effects as fuel consumption and accumulation of fission productpoisons, which determine the core reactivity and tend to slow the nuclear chain reaction.
Rod cluster control assemblies are used for reactor control and consist of clusters ofcylindrical absorber rods. The absorber rods move within guide tubes in certain fuel assemblies.
Above the core, each cluster of absorber rods is attached to a spider connector and drive shaft,which is raised and lowered by a drive mechanism mounted on the reactor vessel head.
Downward trip of the rod cluster control is by gravity. A number of safety systems are provided
on PWRs to respond to loss-of-coolant, loss-of-electric power and reactivity initiated events.
These consist of:•a fast-acting SCRAM system;•boron injection capability;•

high-pressure and low-pressure coolant injection systems;•

an auxiliary feedwater system for decay heat removal through the steam generators;•

backup power supplies;•in containment spray systems to condense steam and scrub fission products from thecontainment atmosphere (although not all PWRs have this feature);•containment cooling systems.

2.2.3. Pressurised water reactor safety issues
Design, operation and research associated with PWRs have generated information to address
many safety issues, including:•anticipated transient without SCRAM;•

loss-of-coolant accidents;•

loss-of-feedwater transients;•

steam line breaks;•

station blackout;

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severe accident concerns:–

core debris criticality under reflood conditions;–in-vessel melt retention by late reflooding;–high-pressure vessel failure and direct containment heating due to RPV bottom head
failure;–core concrete interaction (MCCI);–

ex-vessel core debris coolability or external cooling of the RPV;–

in- and ex-vessel fuel coolant interaction;–

combustible gas control;–

containment cooling/integrity/venting;–source term issue.

As is the case for BWRs, PWR safety issues are routinely re-evaluated in response to industry
initiatives such as raising power or fuel burnup levels, increasing operating cycle lengths or
extending plant lifetime. These issues have included:•

reactivity initiated events (e.g. boron dilution);•plant ageing, including materials cracking/corrosion;•

power uprates;•

high-burnup fuel/MOX fuel;•installation of digital I&C;•

ECCS sump strainer clogging;•

impact on human actions and reliability;•

steam generator tube rupture.

Some of these issues are shared with BWRs. In addition, like BWRs, emphasis on riskinformed regulation in some countries and more detailed assessments of accident scenarios
and accident management actions has led to new issues needing research, such as:•

redefinition of large-break LOCA;•

break location and orientation.

Many of these issues have been largely resolved today due to the huge research efforts inthe past decade and due to accident management programmes, which have been developed toprevent or mitigate many of the severe accident concerns mentioned. Finally, Fukushima
evaluations have required consideration of design-basis extension and beyond design-basis
accidents in PWR safety issues, some of which are discussed in Chapter 3.

2.3. Water-water energetic reactors
2.3.1. Introduction
Nuclear power plants with VVER reactors of Soviet origin are operated in the Russian Federationand several other European countries. Some of these are NEA member countries such as theCzech Republic, Finland, Hungary and the Slovak Republic where nuclear power plants with
VVER-440/213 and VVER-1000/320 reactors are in operation; older reactors of generation
VVER-440/230 are already shutdown. VVER reactors are classified as a specific type of PWR
reactor. However, there are some significant differences between the VVER and other types of
PWR, both in terms of design and materials used. Distinguishing features of the VVER include:
use of horizontal steam generators; use of hexagonal fuel assemblies; avoidance of bottom
penetrations in the reactor vessel; use of high-capacity pressurisers. The review of their designand safety features shows that the main concept of these reactors is similar to PWR units
designed at the same time in other countries.

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A next-generation of VVER-1000 reactors with larger reactor power and different passive
safety features has already been designed and some plants are under construction in Russia,
Belarus, the People’s Republic of China (hereafter “China”) and India. Drawing on thesignificant body of experience gained with the well-established VVER-1000/V-320, the AES-91(or VVER-1000/V-428) was developed by Saint-Petersburg Atomenergoproekt and the AES-92(or VVER-1000/V-412 and 466) by Moscow Atomenergoproekt. Along with upgraded technology
and improved economics, these designs deployed the concept of beyond design-basis accident
management based on a balanced combination of passive and active safety systems. Only
small modifications were made in the basic power production systems – reactor, primarycooling circuit and turbine cycle. The main changes were in the safety systems and plant layout.
The VVER-1200 AES-2006 design is the latest evolution in the long line of VVER plants. It meets
all the international safety requirements for Gen III+ nuclear power plants. The first AES-2006
units are under construction in Russia (Rosatom Overseas, 2018).
The latest project is the VVER-TOI (typical, optimised, with enhanced information) project tocreate a standardised VVER power plant, optimised in terms of technology and economics. It isbeing developed by Moscow Atomenergoproekt and is based on the AES-2006/V-392M design. It
represents a further evolution of the VVER-1200 design and is designated V-510. This design has
an upgraded pressure vessel, increased power to 3 300 MWt and 1 255-1 300 MWe gross (nominally
1 300 MWe), improved core design to increase cooling reliability, further the development ofpassive safety with a 72-hour grace period requiring no operator intervention after shutdown,lower construction and operating costs, and a 40-month construction time. It will use a low-speed
turbine-generator (WWN, 2019).

2.3.2. Water-water energetic reactor design features
The VVER is a pressurised water reactor, employing light water as a coolant and moderator. In
the design safety philosophy of the early VVER-440/230, preventive features dominated overmitigating actions, which led to certain inherent safety features, such as low power density, large
coolant volumes and cracking resistance of the primary circuit, as well as low impact ofequipment failures due to a large number of primary loops with isolation valves. However, there
were also a number of deficiencies that are related to the differences in engineering design
solutions, shortcomings in engineered safety features such as insufficient emergency corecooling system, missing containment, quality of manufacturing and reliability of equipment. One
weld of the reactor pressure vessel also proved to be prone to radiation embrittlement. Many ofthese positive and negative features were inherited by the next VVER generations, but the designorganisations started to improve safety in line with western safety standards. Back-fitting of theexisting reactors has been intensive during the recent decade and the process is not muchdifferent from that which is going on in plants built to earlier safety standards all over the world.

2.3.3. Water-water energetic reactor safety issues
As a general statement, it may be concluded that the majority of safety issues of VVERs are thesame or very similar to those of other PWRs. This includes severe accident issues.
The safety of nuclear power plants with VVER reactors was reviewed in the framework of
the International Atomic Energy Agency’s (IAEA) Extrabudgetary Programme (EBP) on the safetyof VVER and high-power channel-type reactor (RBMK) nuclear power plants during the period
1990-98. The programme addressed the safety issues and ranked them according to their safetysignificance into four categories in the areas of reactor core, component integrity, systems,
instrumentation and control, electrical power supply, containment, internal and external
hazards, accident analysis, operating procedures, management, plant operation, radiation
protection, training, and emergency planning.
Regarding the safety ranking of the more modern reactor type VVER-440/213, no safetyissues of the highest category were identified. High safety concerns include such issues as
insufficient qualification of equipment for anticipated ambient and seismic conditions, seismic
safety in general, strength of some structural elements of the bubbler condenser, deficiencies
of in-service inspection of reactor coolant system, ECCS clogging under LOCA, layout of the

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2. OVERVIEW OF REACTOR DESIGNS AND SAFETY ISSUES

emergency feedwater system, fire protection and possible multiple failures of safety-related
systems in high-energy pipe breaks at certain locations. These safety issues have since been
resolved at all the plants via retrofitting programmes.
The VVER-1000 concept may be considered much closer to the other PWRs. Its power ishigher and safety margins smaller than in the VVER-440 concepts, but in its safety philosophy,
defence in depth has been taken into account from the beginning. The IAEA safety review for the
standard series VVER-1000/320 did not identify any safety issues of the highest category. Safetyissues for the standard series included qualification of equipment, control rod insertion reliability,
reactor pressure vessel embrittlement and monitoring, non-destructive testing, steam generator
collector integrity, steam and feedwater piping integrity, steam generator safety and relief valves
qualification for water flow, reactor vessel head leak monitoring, emergency battery discharge
time and fire prevention. For the moment, these safety issues have been resolved.
The majority of safety issues, especially of the first-generation VVER plants, have been
identified as deviations from current standards and practices, which have evolved since these
nuclear power plants were designed. In all countries, extensive back-fitting and upgrading
programmes have been performed, which have resolved a great majority of the remaining
safety issues; older plants of VVER-440/230 generation have been shut down.
The particular problem of these older VVERs was that their safety analysis was not validated
via experimental facilities. It was for this reason that the experimental validation of computercodes and system behaviour was of great importance. As an example, the successful experimental
validation of the bubble condenser system of VVER-440/213 was done, and the experimental
validation of thermal-hydraulic computer codes for VVER-1000/320 at the PSB-VVER facility, as
well.
In the future, it may be expected that the lifetime of the first generation of VVERs will be
extended. This will lead to having to address many of the same issues as for PWRs, such as:•

materials behaviour;•

increased inspections;•component replacement or refurbishment.

Among other things, this most probably will include increased use of digital automation,
which may bring about new safety issues, but most probably will not be specific to VVERs. The
next generations of VVER reactors are now under construction.

2.4. Pressurised heavy water reactors and advanced PHWRs
2.4.1. Introduction
A PHWR is a nuclear reactor commonly using natural uranium or slightly enriched uranium as
fuel, that uses heavy water (deuterium oxide, D2O) as its coolant and neutron moderator. The
heavy water coolant is kept under pressure, allowing it to be heated to higher temperatures
without boiling, much as in a PWR. While heavy water is significantly more expensive thanordinary light water, it creates greatly enhanced neutron economy, allowing the reactor tooperate without fuel-enrichment facilities (offsetting the additional expense of the heavy water)and enhancing the ability of the reactor to make use of alternate fuel cycles.
Two generations of PHWRs of German design are under operation in Argentina. Atucha 1
entered commercial operation in 1974, and Atucha II in 2014. The Siemens design of the Atucha
PHWR units is unique to Argentina. Both are PHWR with a different power employing a mixtureof natural uranium and enriched uranium (0.85% of 235U), and use heavy water for cooling andneutron moderation. The nuclear area is based on the prototype reactor MZFR 56 MW(e), which
was placed in operation at the Nuclear Research Center in Karlsruhe (Germany) in 1966. The
secondary and auxiliary systems as well as the containment and other buildings were designed
in a similar way as Konvoi PWR plants from Germany (around 1979). A main difference to the
CANDU design is the vertical arrangement of the individual fuel channels in the reactor corelike in a PWR. Day-by-day fuel loading is possible as in other PHWRs.

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2. OVERVIEW OF REACTOR DESIGNS AND SAFETY ISSUESThe CANDU (Canada Deuterium Uranium) is a Canadian PHWR design that has also been
deployed worldwide. The acronym refers to its deuterium oxide (heavy water) moderator andits use of (originally, natural) uranium fuel. CANDU reactors were first developed in the late
1950s and 1960s by a partnership between Atomic Energy of Canada Limited (AECL), the HydroElectric Power Commission of Ontario, Canadian General Electric and other companies.
PHWRs of CANDU design are unique for containing the nuclear fuel and coolant in an arrayof horizontal fuel channels, rather than a pressure vessel like in the other type of PWR. Beyond
the headers that distribute coolant to individual feeders for the fuel channels, the remainder ofthe reactor coolant system is similar to a PWR with reactor coolant pumps, steam generators,
etc. To moderate the reactor, the fuel channels are surrounded by low-pressure heavy water ina cylindrical calandria vessel. This moderator water is kept below 100oC, and serves as a backup
heat sink in the event that primary and emergency cooling are lost. In addition, the calandria
vessel is contained within a shield tank or vault, filled with light water to serve as a biologicalshield that provides an additional heat sink in a severe core damage accident.

2.4.2. Pressurised heavy water reactor CANDU design features
The fuel for a PHWR consists of a 0.5 m long bundle of fuel elements (28, 37 or 43, depending on
the reactor), with typically 12 bundles in a fuel channel. Similar to LWRs, the fuel is UO2, usually
natural uranium with some newer fuel designs considering using slight enrichment. The central
element in some fuel designs can contain a small amount of burnable neutron poison (Dy) to
reduce void reactivity. The fuel cladding, end caps, appendages and end plates are all madefrom Zircaloy.
PHWRs can be refuelled on-power through the use of fuelling machines that connect to theends of a fuel channel. Once connected, the short fuel bundles can be repositioned in thechannel to optimise fuel utilisation, and used fuel is replaced with fresh fuel. Because a PHWR
can be refuelled on-power, there is not much excess reactivity in the core, and there is norequirement to poison the coolant to reduce reactivity with the introduction of fresh fuel.
Therefore, reactivity excursions such as boron dilution are not a concern. On the other hand,attention is paid to ensuring adequate protection against fuel-handling accidents.
The current generation of PHWRs use natural uranium fuel and heavy water coolant,
leading to a positive void coefficient. This is accommodated by employing two independent fastacting shutdown systems. The first is spring-assisted shut-off rods that drive down between thefuel channels in the moderator. The second is liquid poison (Gd) injection into the moderator.
Both shutdown systems put neutron absorbing material directly into the low-pressure
moderator, and are therefore not subject to high pressure, nor jamming due to fuel damage, inthe event of an accident.
Similar to all water cooled reactors, PHWRs use an ECCS to provide backup cooling in a
LOCA. There are typically three modes of operation: 1) high-pressure injection;2) intermediate/low-pressure injection; 3) long-term recovery and recirculation. Emergency
core cooling is accompanied by venting steam from the secondary side to “crash cool” thesteam generators and reduce the primary side pressure below the ECCS injection pressure.
There are currently three major types of containment: 1) single-unit containment; 2) multiple
unit containment (incorporating a common vacuum building); and 3) a double containment
system used in Indian PHWRs. The single-unit containment consists of a cylindrical, pre-stressed,
post-tensioned concrete building with a concrete dome. The building has an epoxy lining to
reduce leakage. Short-term pressure rises are mitigated with a dousing system, while local air
coolers are used to provide long-term pressure control and heat removal. Hydrogen igniters andrecombiners prevent build-up of hydrogen to explosive levels.
In a multi-unit vacuum containment, four or eight reactors, each with its own individual
containment, are connected to a vacuum building by large-scale ducting. In the event of a LOCA,
self-actuating valves connect the vacuum building to the ducting. Effluent is then drawn fromthe reactor building to the vacuum building, reducing the pressure. Dousing is used to condense
steam in the vacuum building and to wash out soluble fission products. In the longer term, anemergency filtered air discharge system is used to control pressure, while filtering out fission
products. Multi-unit containments also feature hydrogen recombiners.

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Current Indian PHWRs use a double concrete containment. The inner containment is acylinder and dome of pre-stressed concrete, with an epoxy lining for leak tightness. The outer
containment is a cylinder and dome of reinforced concrete. The intervening space is maintained
at a negative pressure with a purging arrangement. A suppression pool between drywell andwetwell volumes in containment is used to limit peak pressures. The suppression pool also
provides a source of long-term, low-pressure emergency core cooling. Local air coolers also
provide pressure control and heat removal, and there is a filtered system for controlled gas
discharge in the longer term.

2.4.3. Pressurised heavy water reactor CANDU safety issues
The main residual safety issues for operating PHWR CANDU reactors revolve around improved
understanding of phenomena, and reduced uncertainties in safety code predictions. Information
has been generated to address the following generic safety issues:•

limited core damage accident (damage is contained within the fuel channels);•

combustible gas control;•core cooling in the absence of forced flow;•pressure tube failure with consequential loss of moderator;•

void reactivity uncertainty allowance;•

moderator sub-cooling requirements;•

flow distribution in headers during a LOCA.

Issues that continue to be evaluated, particularly with feedback from operating experience,with initiatives to extend plant lifetime and with development of new designs, include:•

plant ageing, including materials cracking and corrosion;•molten fuel-moderator interaction;•

channel voiding during a LOCA;•

flow distributions between channels and header effects;•

severe accidents:–

core disassembly;–

source term;–

in-vessel retention.

Many severe accident issues and phenomena are similar to those for LWRs and thus much of
the LWR severe accident research and strategies for issue resolution can be applied to PHWRs.
However, the use of pressure tubes as a pressure boundary, the horizontal channels, and theexistence of a water-filled calandria vessel and end-shield cooling tank make accident progression
in-core very different than that of an LWR. Specialised facilities have been built to address thesechallenges.

2.5. Advanced light water reactors
2.5.1. Introduction
ALWR designs constitute improvements of current generation pressurised water reactors (PWRs)
and boiling water reactors (BWRs). For the purpose of this study, the ALWR designs considered
are those being developed for deployment in the next five to ten years, which will likely exhibit
some of the following design features:

28•

longer design life (up to 60 years);•advanced materials more resistant to corrosion and cracking;

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2. OVERVIEW OF REACTOR DESIGNS AND SAFETY ISSUES•

advanced fuel designs;•

longer operating cycles;•

the use of passive safety systems for emergency core and containment cooling and for
decay heat removal;•

more automated controls and safety systems, including the use of digital technology;•

less reliance on operator action (and fewer staff).

Although LWRs with other design features (e.g. pre-stressed concrete reactor vessel, thoriumfuel) have been considered in the past, it is assumed that the most likely LWRs will employ steel
RPVs, use UO2 or possibly MOX fuel, and operate at conditions similar to present-day LWRs.

2.5.2. Advanced light water reactor safety issues
Key safety issues associated with ALWRs are not substantially different from current LWR plants,
but the design features to cope with them need to be assessed. Of particular interest will be:•advanced fuel transient performance;•

advanced materials performance;•passive safety feature performance;•

use of digital technology and the impact on human performance.

2.6. Advanced pressurised heavy water reactors
2.6.1. Introduction
Advanced PHWR designs based on the CANDU 6 design are also being developed to improve thesafety and economics of the PHWR. Key features of the Enhanced CANDU (EC6) include:•target life up to 60 years, >90% capacity factor;•

modern steam turbines with higher efficiency and output;•

increased safety and operating margins;•

additional accident resistance and core damage prevention features;•

addition of a reserve water system for passive accident mitigation;•

improved plant security and physical protection.

An additional advancement in the PHWR design is the Advanced Fuel CANDU Reactor (AFCR),
which allows countries to use recycled uranium from the spent fuel of PWR plants to improveuranium resource utilisation. In addition, the AFCR can use thorium-based fuels, which will allow
countries with indigenous thorium resources to use them as a near-term energy strategy tominimise dependence on uranium imports. The use of depleted uranium and enriched uranium
to give a natural uranium equivalent (NUE) based fuel has been successfully demonstrated in
China.
India is also developing an advanced heavy water reactor (AHWR) concept that aims to meetthe objectives of using thorium fuel cycles for commercial power generation. The AHWR is avertical pressure tube type reactor cooled by boiling light water under natural circulation.
A unique feature of this design is a large tank of water on top of the primary containment vessel,
called the gravity-driven water pool. This reservoir is designed to perform several passive safetyfunctions.
The overall design of the AHWR is to utilise large amounts of thorium and the thorium cycle.
The reactor design incorporates advanced technologies and several passive safety features,such as core heat removal through natural circulation; direct injection of ECCS water in fuel;and the availability of a large inventory of borated water in overhead gravity-driven water pool
to facilitate sustenance of core decay heat removal. The ECCS injection and containment cooling
can act (SCRAM) without invoking any active systems or operator action.

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The reactor physics design is tuned to maximise the use of thorium-based fuel through theuse of PuO2-ThO2 MOX and ThO2-233UO2 MOX in different pins of the same fuel cluster, and theuse of a heterogeneous moderator consisting of amorphous carbon (in the fuel bundles) andheavy water in 80-20% volume ratio. The core configuration lends itself to considerable
flexibility and several feasible solutions, including those not requiring the use of amorphous
carbon-based reflectors, are possible without any changes in reactor structure.

2.6.2. Advanced pressurised heavy water reactor safety issues
As is the case for ALWRs, the safety issues for AHWRs are expected to be similar to those of theirpredecessor Generation II designs. However, the advanced reactor designs have incorporated
passive safety systems and additional design changes to reduce core damage probability. Thesechanges, and their effectiveness at addressing current safety issues, will require assessment.

2.7. Gas-cooled reactors
2.7.1. Introduction
The origins of commercial gas-cooled reactors are found in the graphite-moderated carbon
dioxide cooled “Magnox” reactors developed in the early 1950s in the United Kingdom and France.
The high-temperature aspect, which is the high-temperature gas reactor (HTGR) concept, dates
back in the United States to the late 1950s, when the design of the fully ceramic core and the useof the helium gas for cooling were pioneered by General Atomics. This development effort resulted
in the 40 MWe Peach Bottom 1 HTGR and the 330 MWe Fort St. Vrain HTGR, which adopted theblock-type core. Also in the late 1950s, Germany began designing the pebble bed type of HTGR.
Two HTGRs were constructed in Germany, the experimental 15 MWe AVR and the 300 MWe
THTR300. All of the HTGRs mentioned above have been decommissioned.
A Task Group on Advanced Reactor Experimental Facilities (TAREF) evaluated the facilities
required for gas-cooled reactors in 2009 (NEA, 2009) and provided a high-level description of two
different gas-cooled reactor concepts. This information is summarised below.

2.7.2. Design features of gas-cooled reactors
Design features of the high-temperature reactor
The typical high-temperature reactor (HTR) design features include the following:

30•

High-performance coated fuel particles (CFPs) able to contain fission products for the fullrange of operating and postulated accident conditions, with a very low fuel failure fraction
and subsequent release of fission products. The CFPs are embedded in either a rod compact
inserted into a stacked prismatic block or a spherical compact, referred to as a pebble.

•

An inert single-phase high-pressure coolant (helium).

•

A graphite-moderated core with the characteristics of low power density, large heat
capacity, high effective core thermal conductivity and large thermal margins to fuel failure.

•

Negative fuel and moderator temperature coefficients of reactivity, which, along with thenegative reactivity feedback of the fission product xenon-135, are sufficient to shut down
the reactor during loss-of-forced circulation (LOFC) events. This aspect stabilises powercontrol feedback for most reactivity insertion events (for both startup and power
operation) for the entire fuel life cycle and for all applicable temperature ranges.

•

A design-basis accident decay heat removal system, typically a passive system utilising
natural convection-driven processes (the reactor cavity cooling system).

•

A confinement-style reactor building structure to accommodate depressurisations may
be used instead of a leak-tight sealed containment.

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2. OVERVIEW OF REACTOR DESIGNS AND SAFETY ISSUES•

The balance of plant consists of an electrical power generation unit (most often a gas
turbine) and, in some cases, a high-temperature process heat component potentiallyused for the production of hydrogen.

Design features of the gas fast reactor
The gas fast reactor (GFR) system features a high-temperature, helium-cooled, fast-spectrum
reactor with a direct-cycle helium turbine or an indirect cycle using an intermediate heatexchanger for electricity production. It uses a closed fuel cycle. The GFR combines the advantagesof fast-spectrum systems (long-term resource sustainability in terms of use of uranium andwaste minimisation, through fuel reprocessing, recycling and burning of long-lived actinides)with those of the high-temperature reactors (high thermal cycle efficiency and possibly hydrogenproduction), and of the direct-cycle energy conversion option. Its development approach is torely on technologies already used for the HTR but with significant advances, in order to reach theobjectives stated above.
The main GFR design specifications as derived from the general objectives of Generation IV
systems are:•use of gas as a coolant as a means of reaching high temperatures;•

economic competitiveness by means of simplicity, compactness and efficiency;•

a robust safety demonstration, based on probabilistic safety assessment and defence in
depth principles, and including severe accident management;•

fast neutron spectrum core with a zero (self-breeding) or positive breeding gain, with noor very limited use of fertile blankets in order to:•–

generate as much fissile material as it consumes, with an optimal use of uranium;–

have a fuel cycle fed with only depleted or natural uranium;–

achieve homogeneous recycling of all actinides in order to have no separation of
plutonium from other actinides (proliferation resistance).

core plutonium inventory not exceeding 10 tonnes/GWe, in order to have a realistic
reactor fleet deployment (in a few decades) and high fuel burnup.

In contrast to the HTR, GFR cores have a low thermal inertia, with a special design feature
to overcome this apparent unfavourable feature. These include:•

A fuel element based on refractory materials and high thermal conductivity, with theability to ensure radioactive material confinement up to very high temperatures.

•

A primary circuit design based on upward core cooling and a moderate pressure drop for
all the primary components and circuit involved in accident scenarios. A gas-tightenvelope encloses the primary circuit to limit the loss of pressure in the case of a primaryloss of coolant. Maintaining high helium density allows the decay heat removal system
to rely on moderate pumping power and even on passive natural convection in somesituations.

The fuel element is able to withstand high operating temperatures and transients associatedwith the poor heat capacity of the gas coolant. The main temperature limits are the following:•

an operating temperature around 1 000°C that provides a sufficiently ample margin to
failure;•a boundary temperature of 1 600°C, below which fission product release is prevented;•

an upper temperature of 2 000°C, below which the core geometry can safely be cooled
down.

The Generation IV objective of ultimate waste minimisation, proliferation resistance andnatural resources optimisation (zero or positive breeding gain) are to be achieved by having no
fertile blanket and multi-pass recycling of all actinides without separation.

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Two primary fuel concepts are proposed to fulfil the above requirements: a ceramic platetype fuel element and a ceramic pin-type fuel element. The reference material for the structure
is reinforced ceramic, a silicon carbide (SiC) composite matrix ceramic. The fuel compound ismade of pellets of mixed uranium-plutonium-minor actinide carbide. A leak-tight barrier madeof a refractory metal or of Si-based multi-layer ceramics is added to prevent fission productdiffusion through the clad.
The reactor pressure vessel is a large metallic structure with a thickness of 20 cm in the beltline region. The material selected, a martensitic 9Cr-1 Mo steel (industrial grade T91, containing
9% by mass chromium and 1% by mass molybdenum) undergoes negligible creep at operating
temperature (400°C). The reference material for the internals is either 9Cr-1Mo or stainless steel,typically SS316LN. The global primary arrangement is based on three main loops (3×800 MWth),each fitted with one intermediate heat exchanger blower unit, enclosed in a single vessel.
The shutdown system has two redundant and passive shutdown systems (no power supply,
gravity drop of absorber elements). Each main control rod and shutdown device and diversified
shutdown device is individually driven, considering two independent groups each connected to a
dedicated group of the instrumentation and control support system. Specific loops for emergency
decay heat removal are directly connected to the pressure vessel and are equipped with heatexchangers and blowers.
The gas-tight envelope is designed to provide a sufficient pressure in the case of a large gas
leak from the primary system. It consists of a metallic structure, initially filled with nitrogenslightly over atmospheric pressure to reduce the possibility of air ingress.

2.7.3. Safety Issues of gas-cooled reactors
High-temperature reactor
The basic safety design approach for high-temperature reactors (HTRs) is different from mostcurrently operating and advanced LWRs, in that HTRs rely on the retention of fission products
in high-integrity ceramic CFPs in a relatively chemically inert environment to withstandaccidents without fuel damage or fission product release. HTRs are designed with passive heatremoval systems and inherent negative reactivity to limit fuel temperatures and maintain fuelparticle integrity. However, with these novel design features and characteristics, HTR designers
must provide proof of the safety performance of the equipment, including the CFPs; integrity of
the reactor vessel; and supporting safety-related structures, systems and components. Fission
product release and transport behaviour must be well understood and analysis tools must be
validated against an adequate database if a vented confinement is to become an acceptablefeature of an HTR.
The major HTR safety issues of concern that were identified and categorised as highimportance combined with medium to low knowledge can be summarised as follows:•core coolant bypass flows (normal operation);•

power/flux profiles (normal operation);•outlet plenum flows (normal operation);•reactivity feedback coefficients (normal operation and accidents);•

emissivity aspects for the vessel and the reactor cavity cooling system during
depressurised loss-of-coolant (D-LOFC);•

reactor vessel cavity air circulation and heat transfer (D-LOFC);•

convection/radiation heating of upper vessel area during pressurised loss-of-coolant(P-LOFC) accidents.

Gas fast reactor
A specific characteristic of GFR systems is the lack of thermal inertia of the primary system,
combined with the pressurised gas and the associated risk of leakage. From a thermal-hydraulicpoint of view, P-LOFC and D-LOFC are two main categories of transients that strongly influence32

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2. OVERVIEW OF REACTOR DESIGNS AND SAFETY ISSUES

the design of the system. Safety of the reactor mainly relies on active systems, and progressionof a transient depends on the characteristics of several critical components. Hence, thebehaviour of these safety critical components is a relevant GFR safety issue. In a fast neutron
environment, the GFR core physics is sensitive to several reactivity effects that require specificattention as related to, for example, water ingress or reactivity effects resulting from accidental
conditions. In this context, there will be a need for accurate nuclear data for GFR materials at
representative neutron spectrum conditions.
A key conclusion from the gas-cooled reactor evaluation was that existing and planned
facilities in member countries cover all technical areas of concern and most of the safety issues
identified in these areas. Hence, there is no apparent need for a facility to be built (beyond what
is currently planned in member countries).
Based on the responses received, the highest ranked facilities were identified and are shownin Table 2.1. Detailed descriptions of each of the facilities can be found in the TAREF report (NEA,2009).

Table 2.1. TAREF gas-cooled reactor summary ratings
Accident andthermal fluids
Czech Republic

Fission producttransport

High-temperaturematerials

Graphite andceramics

HTHL

HTHL

HTHL
HEDYT

France*HEDYT
ENIGMA

MERARG

Germany

HELOKAA2

THAIHigh Power Laser
Lab

Italy

HE-FUS3

Japan

HTTR

HTTR

HTTR

ATR

ORNL Materials Lab
INL High Temp Test
Lab

United States

Fuel

PLINIUS

NSRR
MIT
HFIR

ACRR
ATR
MIT

* For the longer term (2020 and beyond), the French gas fast reactor demonstration reactor ALLEGRO should also be considered.

2.8. Sodium fast reactors
The TAREF also conducted a study of sodium fast reactors (SFRs; NEA, 2011). The SFR system
features a fast-spectrum, sodium-cooled reactor and a closed fuel cycle for efficient managementof actinides and conversion of fertile uranium. This section provides a summary of the 2011 TAREF
report.

2.8.1. Design features of sodium fast reactors
The fuel cycle employs a full actinide recycle with two major options. One is an intermediatesize (150-600 MWe) sodium-cooled reactor with uranium-plutonium-minor-actinide-zirconium
metal alloy fuel, supported by a fuel cycle based on pyrometallurgical processing in facilities
integrated with the reactor. The second is a medium to large (500-1 500 MWe) sodium-cooled
reactor with mixed uranium-plutonium oxide fuel, supported by a fuel cycle based on advanced
aqueous processing at a central location serving a number of reactors. The outlet temperature
is approximately 550°C for both.
The SFR is designed for management of high-level wastes and, in particular, managementof plutonium and other actinides. Important safety features of the system include a longthermal response time, a large margin to coolant boiling, a primary system that operates nearatmospheric pressure, and an intermediate sodium system between the radioactive sodium inthe primary system and the water and steam in the power plant. With innovations to reduce
capital cost, the SFR can serve markets for electricity.
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2. OVERVIEW OF REACTOR DESIGNS AND SAFETY ISSUESThe SFR's fast spectrum also makes it possible to use available fissile and fertile materials(including depleted uranium) considerably more efficiently than thermal spectrum reactors
with once-through fuel cycles.
The SFR system uses liquid sodium as a coolant, allowing high-power density with low
coolant volume. The primary system operates at near-atmospheric pressure with outlet
temperatures between about 500°C and 550°C. Austenitic and ferritic steel structural materials
can be used, with a large margin (about 400°C) to coolant boiling. Despite these desirable
features, one key characteristic of the various SFR designs that warrants special mention. Thisis the positive sodium void reactivity, except for very small cores, and the possibility of corere-criticality in the event of a core damage event.
Two main types of designs exist, the loop design, wherein the primary coolant and the heatexchangers are outside of the RPV, and the pool-type reactor, in which the coolant and heatexchangers reside inside the RPV. The loop design has advantages in ease of maintenance andthe ability to have a significant height difference between the coolant and the heat exchangers,
to promote natural circulation. Its disadvantages are that a loss-of-coolant could result in
sodium fires. The pool-type reactor has advantages because it has a secondary vessel to containany loss of coolant, nearly eliminating core uncover risks, and a large thermal inertia which
mitigates accident consequences by absorbing excess core heat. Disadvantages are primarily
related to ease of access to components residing within the pool.
Key results from the TAREF assessment are summarised below.

2.8.2. Safety issues of sodium fast reactors
The TAREF group members agreed that for new SFR projects, the most important and top-tier
R&D safety needs concern the technical areas with the following priority order:•

fuel safety and severe accident issues are of prime interest due to the lack of knowledge
on new pin design and materials;•

thermo-fluids and reactor physics issues are of second priority, as one can live with thecurrent knowledge when considering some margins to cover uncertainties;•

sodium risks and structural integrity issues may be considered a third priority, as they
are more design-dependent.

Key to resolving the safety issues, the group stressed the need for fuel pin irradiation
capabilities under representative conditions of fast neutron flux. They identified that a large
body of work had been undertaken during the period 1970-95, followed by a slowdown, and thata large number of facilities operating in the past with sodium as coolant are no longer available.
A number of facilities appropriate to address SFR safety issues were evaluated. They aresummarised below, almost verbatim from the TAREF report and identified in Table 2.2, alongwith facilities that may be available to address them in the short, medium and long term.

34•

The Indian FBTR fast reactor can be a valuable resource for irradiation of SFR fuel pinsand new materials data; the American reactor ACRR (Sandia) would address issues
related to fuel safety and severe accidents under specific conditions (provided there isconfirmation of its availability for testing in the short term).

•

The German KASOLA (Karlsruhe Institute of Technology) facility would provide data forthe thermo-fluids issues in relation with computational fluid dynamics modelling
approaches.

•

The Japanese SWAT-1R-3R facility (Japan Atomic Energy Agency – JAEA) can be appropriate
for studying sodium-water interaction in steam generator units; the Indian SFTF facility
can be valuable for addressing several issues related to sodium fires; the SURTSEY facility(US Department of Energy – DOE) can be relevant to studies on sodium fires and sodiumwater interaction in steam generators.

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2. OVERVIEW OF REACTOR DESIGNS AND SAFETY ISSUES•The JOYO fast neutron reactor (JAEA, Japan) was identified as suitable to address fuelsafety issues related to new fuel pin design (fuel pin performance and new materials
database under irradiation, margin to fuel melting, impact of use of minor actinides) andsome other issues; however, since it is under safety review by the nuclear regulation
authority to ensure that new requirements prompted by the Fukushima accident are
respected, uncertainty exists as to when operation will be permitted.

•

Severe accident issues can only be addressed in a comprehensive way for the medium
term and beyond due to the lack of available facilities for simulation of representativetransient conditions in the short term with irradiated fuel pins. IGR (Kazakh facility used
for JAEA programmes), which is addressing fresh fuel (controlled fuel relocation, debris
bed formation), may be a suitable solution in the medium term as plans are underconsideration for it to handle irradiated fuel. The VULCANO (Alternative Energies andAtomic Energy Commission, CEA, France) can also help for severe accident issues,
provided it is refurbished for sodium use. The TREAT experimental reactor (US DOE) was
also considered in the medium term for its relevance to severe accident issues (pastexperimental programmes simulating fast power transients).

•

The MASURCA (CEA, France) was identified as being suitable for core physics issues forproviding improved nuclear data of core materials (in relation with high burn-up level,use of minor actinides) and associated uncertainties; however, a decision was taken to
shut down this facility subsequent to issuing the TAREF report.

•

The CABRI experimental reactor (operated by the CEA) was recognised by members as themost appropriate facility to address irradiated fuel behaviour under incidental andaccidental conditions (fuel safety issues such as margins to fuel melting and deterministic
pin failure, severe accident issues such as consequences of various accidents leading tofuel melting, with associated consequences and risk of critical events and energy release).
However, the sodium loop in the facility has now been decommissioned.

•

In the case of innovative design for secondary circuits, the LIFUS5 Italian facility(National Agency for New Technologies, Energy and Sustainable Economic Development,
ENEA) would address sodium interaction with alternative coolant species.

Table 2.2. Highest ranked issues and facilities for sodium fast reactors
Thermo-fluids

Fuel safety

Physics

Severe accident

Flow regime transitions,transport properties,
channel flow distribution,
sodium boiling
KASOLA, NADYNE

Fuel pin performance understeady-state conditions(irradiation, cladding)
FBTR, JOYO, ASTRID, JHR

Doppler, fuelexpansion reactivity
feedback

Unprotected loss of flow withsubsequent transient overpower
accident and uncontrolled passage
of gas bubble accident
MELT, IGR, ACRR, CAFÉ, TREAT,

Coolant structureinteraction KASOLA, TTS,
PLANDTL

Margin to fuel melting
JOYO, TREAT, JHR

Reactivity feedback
due to sodium
voiding

Local blockage accident
TREAT

New fuel pin designs andmaterials (fuel, cladding)
FBTR, ACRR, JOYO

Fuel coolant interaction
MELT, IGR, ACRR, MCCI, SURTSEY,VULCANO, TREAT

Use of minor actinides
FBTR, ACRR, JOYO, TREAT

Post-accident decay heat removal
MELT, IGR, ACRR, MCCI, VULCANO
Release of fission products (source
term uncertainty)
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2.9. Molten salt reactors
The molten salt reactor (MSR) is a concept that dates back to the 1950s, when the design was
first proposed as the propulsion system of a nuclear-powered aircraft at the Oak Ridge National
Laboratory. It is now one of the reactor concepts identified by the GEN IV International Forumas a candidate for co-operative development.
The GEN IV initiative, in conjunction with national reactor development drivers, revitalised
safety research interest in the concept (Serp et al., 2013; Elsheikh, 2014; Watt-Logic, 2017), whichcan:•minimise weapons useable material in storage;•

minimise need for high-level waste repository space;•increase the proliferation resistance of nuclear energy;•

make beneficial use of spent fuel from LWRs;•

increase resource utilisation.

2.9.1. Design features of molten salt reactors
MSRs are a class of nuclear fission reactors in which the primary coolant, or even the fuel itself,
is a molten salt mixture. Two concepts are available. In the first, fissile material is dissolved in
the molten salt. In the second, the molten salt serves as the low-pressure coolant to a coatedparticle fuelled core similar to that employed in HTRs. The solid fuel variant is typically referred
to as a fluoride salt-cooled high-temperature reactor (FHR).
Since the 1940s, several MSR concepts have been proposed all over the world using different
fuel (uranium [U], plutonium [Pu], thorium [Th]) and salt compositions (chlorides, fluorides).
These research projects attempted to find the optimum solution for the fuel salt composition
suitable simultaneously for optimising:•neutronic properties (neutron moderation, breeding ratio, fissile inventory);•

operating temperature (melting temperature, radiation stability, transport properties);•

actinide and fission products solubility in the considered molten salt to guarantee ahomogeneous core composition;•materials compatibility and salt chemistry control;•

processing and low-waste options.

Design concepts for MSR abound, but some common themes can be identified. First, MSRs
run at much higher temperatures (up to 700-750°C) than LWRs and operate at near-atmosphericpressure. Molten salts offer attractive characteristics as coolants, especially their high volumetricheat capacities and high boiling points. In liquid fuelled MSR designs, the nuclear fuel is dissolvedin the molten fluoride salt coolant as fissile elements such as UF4, PuF3, minor actinides fluorides
and/or fertile elements such as ThF4 depending on the desired application (breeder reactor,
actinide burner, etc.). Liquid fuelled MSRs are often associated with the 233U-232Th fuel cycle which,by using abundant thorium as a fertile element, enables breeding with a thermal spectrum, andis often considered as more convenient than the U-Pu fuel cycle in order to minimise thegeneration of highly radiotoxic transuranic elements.
Examples of different concepts demonstrated already include:•36

In the first phase of the Molten Salt Reactor Experiment (MSRE) at Oakridge from 1965 to
1968, 235U tetrafluoride (UF4) enriched to 33% was dissolved in molten lithium, beryllium
and zirconium fluorides at 600-700°C which flowed through a graphite moderator atambient pressure, with the fuel comprising about 1% of the fluid. The second phase of
the project (1968-69) used 233U, making MSRE the first reactor to use 233U. This programme
paved the way for building an MSR breeder utilising thorium, which would operate in thethermal (slow) neutron spectrum.

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2. OVERVIEW OF REACTOR DESIGNS AND SAFETY ISSUES•In the United Kingdom, plans for a 2.5 GW lead-cooled fast-spectrum MSR with theplutonium fuel dissolved in a molten chloride salt were only recently declassified (EPD,2015). Theoretical work on the concept was conducted between 1964 and 1966, while
experimental work was ongoing between 1968 and 1973.

There are a host of MSR designs currently under development, including several at theconceptual designs highlighted in Watt-Logic (2017). Examples, which vary widely with respect
to technology readiness levels include:•

In Europe, concepts have been focused on fast-spectrum concepts with or without
thorium support (e.g. molten salt fast reactor [MSFR] and molten salt actinide recycler
and transmuter [MOSART] led by the Kurchatove Institute in Russia (Ignatiev et al., 2015),
which have been recognised within the GEN IV International Forum as long-term
alternatives to solid-fuelled fast neutron reactors with attractive features (strongly
negative feedback coefficients, and smaller fissile inventory, simplified fuel cycle).
Several considerations have demonstrated the potential of the MSFR and MOSART as
systems with flexible configurations and fuel cycle scenarios which can operate withintechnical limits with different loadings and make up based on transuranic elements(TRUs) from spent LWR fuel with MA/TRU ratios up to 0.45 as special actinide
transmuters, as self-sustainable system or even as a breeder.

•

In the United States, Massachusetts-based Transatomic Power Corp is planning todevelop a single-fluid MSR with a zirconium hydride moderator and a lithium fluoride –uranium fluoride salt, which can hold about 27 times as much uranium as an LWR.

•

Newly incorporated Alpha-Tech Research Corp in Utah (United States) is developing a
30 MW thorium test reactor to produce medical isotopes, and has a consortium of seven
Utah counties considering supporting the project.

•

Flibe Energy, based in Alabama (United States), is studying a thorium-fuelled, graphitemoderated thermal reactor concept based on the 1970s MSRE. It uses lithium fluoride/
beryllium fluoride (FLiBe) salt as its primary coolant in both circuits. It starts with a pilot
plant, followed by a 100 MW demonstration reactor entering operation in 2023, andultimately a commercial plant in the 250-1 000 MW range ready by 2027.

•

In the United States ThorCon International is developing small modular reactors based on
the MSRE concept and believes it can build a full-scale 250 MW prototype in as little as four
years, providing a suitable test site can be found. The company is collaborating with theIndonesian government launching a molten salt thorium reactor that is expected to be
operating by 2025 (Dalton, 2017).

•

In 2011, the Chinese Academy of Sciences announced its intention to commercialise athorium-based MSR in 20 years (Mitchell, 2016) (non-thorium MSRs and solid fuelthorium reactors are also being developed).

•

The Shanghai Institute of Applied Physics (SINAP) has two streams of thorium MSR
development (Yu and Xu, 2016; Wang, 2016): solid fuel, in pebbles or prisms/blocks witha once-through fuel cycle, and liquid fuel (dissolved in FLiBe coolant) with reprocessing
and recycle. A third stream of fast reactors to consume actinides from LWRs is planned.

•

Canada’s Terrestrial Energy is developing a modular liquid fuel MSR using low-enricheduranium. Each unit is envisaged to have a 190 MW capacity, and be deliverable in four
years. A feasibility study is underway to explore siting the world’s first commercial unit
at Canadian Nuclear Laboratories.

•

Thorium Power Canada claims to have a construction-ready, modular thorium reactor.
A 10 MW plant is planned for Copiapa in Chile, while a 25 MW demonstration plant isslated for Indonesia.

•

UK-based Moltex Energy’s stable salt reactor is a conceptual reactor design with nopumps that relies on convection from static vertical fuel tubes in the core to convey heatto the steam generators. The reactors are designed to be modular in construction, withunit sizes from 150 MW.

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Several universities and companies propose reactor concepts of molten salt reactors inJapan, including 10 MW mini thorium molten salt power plants for supplying IT servers
and electric vehicle charging stations, and integral molten salt fast reactor (IMSFR)
proposed as a highly efficient transmutations system for tansuranium elements(Yamawaki and Koyama, 2015).

•

Seaborg Technologies in Denmark is also developing modular thorium MSRs, and isplanning a 50 MWt pilot unit with a view to 250 MWt commercial modular units fuelledby spent LWR fuel and thorium.

•

In 2017, NRG (Netherlands) announced the use of thorium reactors in collaboration with
the EU. The Salt Irradiation Experiment (“Salient”), based at the high flux reactor in Petten,
will begin with experiments to melt thorium fuel and then bombard it with neutrons totransmute the thorium into 233U, which can sustain the chain reaction needed to generateenergy. If this is successful, the next step will be to study tough, temperature-resistantmetal alloys and other materials that can survive the high heat and corrosive conditions
inside a molten salt reactor. Further investigations into handling of waste materials will
follow.

2.9.2. Safety issues of molten salt reactors
MSRs have some unique characteristics offering a potentially safer, more efficient and sustainable
form of nuclear power associated with online fuel processing. For example:•

MSRs do not operate under high pressure, and are not cooled by water, making steam
explosion impossible.

•

For the case in which the fuel is dissolved in a molten salt, in case of reactor damage, thefuel salt can be drained into sub-critical storage tanks.

There are several unique safety and operational challenges associated with MSRs, however:38•

The risks of corrosion by the impurities (oxygen and water primarily) dissolved into a
molten salt coolant or by the fission products present in fuel salt. Advanced materials
investigations for MSFR and FHR designs are currently underway in different countries,
but qualification efforts for special materials involved in the MSR design are still needed.
Experience gained at the ORNL in the 1960s and 1970s remains a source of knowledge
that demonstrated many aspects of the technology; however, new designs with different
salt mixtures present challenges.

•

The sometimes conflicting requirements for corrosion resistance and limiting radiation
damage. Nickel steels such as Hasteloy, while superior for molten salt corrosion resistance,
undergo radiation-induced embrittlement.

•

Beryllium is a proposed component of some fuel designs; it is highly toxic.

•

Normal operation produces small amounts of fission product tellurium, which deposits
on surfaces and leads to intergranular cracking.

•Use of lithium (Li) creates some issues. 7Li must be used, since 6Li is a neutron poison.
This necessitates isotopic enrichment of the fuel salts to remove 6Li.

•

Fission gas tritium is produced from irradiation of 7Li directly, and indirectly from 6He.
Hence, tritium management strategies must be adopted. The PHWR community has
significant experience with tritium management; however, the complication of muchhigher operating temperatures allows rapid diffusion of tritium through the walls or the
primary heat exchanger to the coolant salt and into the steam system. This could bedischarged during a steam system blowdown. Investigations of the use of additional
molten salt eutectic mixtures in a separate loop for tritium removal in MSR somewhat
challenges current technologies.

•

Waste disposal is complicated by fission products being in a water-soluble form as
fluoride salts. These are not as amenable to long-term storage as insoluble waste forms.

•

Even in insoluble waste forms, irradiation (self) of immobilised waste can produce UF6
and fluorine gas.

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2.10. Small modular reactors
Since 2010 there has been a flurry of activity in the development of small modular reactors.
A recent post by the World Nuclear Association states:
There is strong interest in small and simpler units for generating electricity from
nuclear power, and for process heat. This interest in small and medium nuclear
power reactors is driven both by a desire to reduce the impact of capital costs andto provide power away from large grid systems. The technologies involved are
numerous and very diverse. (WNA, 2021)
As nuclear power generation has become established since the 1950s, the size of reactor
units has grown from 60 MWe to more than 1 600 MWe, with corresponding economies of scalein operation. At the same time, there have been many hundreds of smaller power reactors built
for naval use (up to 190 MW thermal) and as neutron sources yielding enormous expertise in
the engineering of small power units. The IAEA defines “small” as under 300 MWe, and up to
about 700 MWe as “medium” – including many operational units from the 20th century. Together,
they are now referred to by the IAEA as small and medium reactors (SMRs). However, “SMR” isused more commonly as an acronym for “small modular reactor”, designed for serial
construction and collectively to comprise a large nuclear power plant. (In this report, the use ofdiverse pre-fabricated modules to expedite the construction of a single large reactor is not
relevant.) A subcategory of very small reactors – vSMRs – is proposed for units under about
15 MWe, especially for remote communities.

2.10.1. Design features and safety issues of small modular reactors
SMR designs cover a range of concepts, many of which were described in this chapter, and mostof which have advanced safety features. The safety issues, therefore, are the same as those
identified for the larger scale reactors. There are a few unique concerns that arise as the resultof the use of SMRs in remote areas, however:•

The remote nature of some installation sites requires more extensive evaluation for
implementation of emergency measures in the event of an accident.

•

There are security concerns for remote operation, due to difficulties in ascertaining that
fuel diversion is not occurring.

•

Digital security/safety solutions may be difficult to access without appropriate IT
infrastructure.

•

Many proposed installations in the north are in areas that are environmentallyvulnerable. Environmental concerns range from damage to permafrost coverage, lack ofaccess to secondary heat sinks that conventional reactors rely on, enhanced radionuclide
migration to the biosphere from routine or accidental releases, and ease of establishing
rigorous environmental monitoring.

References
Dalton, D. (2017), “ThorCon molten salt reactor would be viable for Indonesia, study
concludes”, NUCNET, www.nucnet.org/news/thorcon-molten-salt-reactor-would-beviable-for-indonesia-study-concludes.
Elsheikh, B.M. (2013), “Safety assessment of molten salt reactors in comparison with light water
reactors”, Journal of Radiation Research and Applied Sciences, Vol. 6/2, pp. 63-70, https://doi.org/
10.1016/j.jrras.2013.10.008.
EPD (2015), MSR Review: Feasibility of Developing a Pilot Scale Molten Salt Reactor in the UK, Energy
Process Developments Ltd., London, www.energyprocessdevelopments.com/uploads/
EPD%20MSR%20Review%20Feasibility%20Study%20July%202015%201.02.pdf.

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2. OVERVIEW OF REACTOR DESIGNS AND SAFETY ISSUES

Ignatiev, V. et al. (2015), “MARS: Story on molten salt actinide recycler and transmuterdevelopment by Rosatom in co-operation with Euratom”, www.oecd-nea.org/iempt13.
Mitchell, M. (2016), “China’s incipient LFTR”, Adam Smith Institute, blog, www.adamsmith.org/
blog/chinas-incipient-lftr.
NEA (2011), “Experimental Facilities for Sodium Fast Reactor Safety Studies”, OECD Publishing,Paris, www.oecd-nea.org/jcms/pl_18956.
NEA (2009), “Experimental Facilities for Gas Cooled Reactor Safety Studies”, OECD Publishing,Paris, www.oecd-nea.org/TAREF2009.
nextBIG Future (2016), “17 countries cooperating on molten salt nuclear reactor design anddevelopment”, nextBIG Future, www.nextbigfuture.com/2016/11/17-countries-cooperatingon-molten-salt.html.
Rosatom Overseas (2018), The VVER Today: Evolution, Design, Safety, Rosatom Overseas, Moscow,
www.rosatom.ru/upload/iblock/0be/0be1220af25741375138ecd1afb18743.pdf.
Serp, J. et al. (2014), “The molten salt reactor (MSR) in generation IV: Overview and perspectives”,
Progress in Nuclear Energy, Vol. 77, pp. 308-319, https://doi.org/10.1016/j.pnucene.2014.02.014.
Wang, B. (2016), “Update on the liquid fluoride thorium reactor projects in China and the USA”,
nextBIG Future, www.nextbigfuture.com/2016/10/update-on-liquid-fluoride-thorium.html.
Watt-Logic (2017), “New project re-ignites European interest in thorium”, Watt-Logic,
http://watt-logic.com/2017/09/10/thorium.
WNN (2019), “Russia’s VVER-TOI reactor certified by European utilities”, World Nuclear News,
http://world-nuclear-news.org/Articles/Russia-s-VVER-VOI-reactor-certified-by-European-ut.
WNA (2021), “Small nuclear power reactors”, World Nuclear Association, www.worldnuclear.org/information-library/nuclear-fuel-cycle/nuclear-power-reactors/small-nuclearpower-reactors.aspx.
Yamawaki, M. and T. Koyama (2015), “Reducing the burden of high level radioactive waste withtransmutation-proposal of integral molten salt fast reactor (IMSFR)”, 3rd Japan-China
Academic Symposium on Nuclear Fuel Cycle.
Yu, X. and H. Xu (2016), “Update on SINAP TMSR research” presentation and the MSR Workshop,
ONRL, 4-5 October 2016, https://public.ornl.gov/conferences/MSR2016/docs/Presentations/MSR2016-day1-15-Hongjie-Xu-Update-on-SINAP-TMSR-Research.pdf.

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3. TECHNICAL ISSUES AND ASSOCIATED FACILITIES

3. Technical issues and associated facilities

This chapter describes the safety research issues currently being investigated, or identified as
needing investigation, to support the continued operation of current plants and/or thedevelopment of future plants.
The first section discusses issues and facilities unique to the nuclear industry and addresses
those technical areas where experimental data specific to the nuclear industry may be essential
for addressing the safety issues. The technical areas addressed are:•thermal-hydraulics;•

nuclear fuel;•reactor physics;•

severe accidents and containment phenomena;•

integrity of equipment and structures.

It lists the safety research issues currently associated with each technical area. In general, thesafety research issues are relevant to all reactors, but those that are specific to one type of reactor
are noted. The reactor types addressed in this section include boiling water reactors (BWRs),
pressurised water reactors (PWRs), water-water energetic reactors (VVERs), pressurised heavy
water reactors (PHWRs), advanced light water reactors (ALWRs) and advanced pressurised heavy
water reactors (APHWRs). Other reactor types with issues corresponding to the above technicalareas are addressed briefly in the second section.
A table for each technical area provides a list of the safety issues that currently may require
some degree of additional research to improve the state of knowledge and support issue
resolution. The table lists the facilities currently available that are conducting or have thepotential to conduct research relevant to each issue.
The previous version of this document included a ranking of the importance of each safetyissue, followed by a ranking of facilities on the basis of their importance to resolving the issue.
Facilities at risk were considered on the basis of their versatility to address more than one issue,
their uniqueness and the cost of their operation or replacement. This general approach was
followed in the current assessment, although a detailed ranking was not done in the same
systematic manner, largely because the previous thorough assessment established theimportance of many of the facilities to resolving safety issues, and the cost of their replacement
and operation. In most cases, the issues faced today are very similar to those faced at the end of
the 2000s. Some have been largely resolved (e.g. boron dilution and wear) and have decreased inimportance, but regulatory drivers, life extension of ageing nuclear power plants and advances inreactor design have expanded the number of issues. The Fukushima accident has also identified
issues previously not foreseen. Care has been taken to identify only those facilities that might
require significant investment (>EUR 5 million) or medium investment for very unique facilities(>EUR 2 million). Furthermore, except where identified, it is assumed that all of the issues have ahigh to medium safety relevance, and that additional research is required to resolve them.
The technical areas addressed in Section 3.2 on issues and facilities not unique to the nuclear
industry are:•human and organisational factors;•

plant control and monitoring;•

cybersecurity;•

external events;

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3. TECHNICAL ISSUES AND ASSOCIATED FACILITIES•

fire assessment.

Each of these technical areas also includes a description of associated safety research issues
and facilities that have been identified to resolve the issues. In general, data needs to resolvethese issues can be met using facilities more widely available because of use by a broader
industry base, because they are not unique to the nuclear industry. No recommendations weremade in the previous report for the NEA Committee on the Safety of Nuclear Installations (CSNI)
to consider related to preservation of these facilities, which were listed for information only.
The current report does identify a unique fire assessment capability that is at risk, however.

3.1. Issues and facilities unique to the nuclear industry
3.1.2. Thermal-hydraulics
Introduction
Thermal-hydraulics is one of the fundamental disciplines for the design and operation of watercooled nuclear reactors. Achieving large core power densities requires a deep understanding ofthermal-hydraulics. Thermal-hydraulics became one of the main nuclear safety disciplines whenpostulated accidents like the loss-of-coolant accident (LOCA) and other thermal-hydraulictransients were identified as the dominant safety concern for light water reactors (LWRs). As fullscale experimentation was not feasible in most situations, significant computational
developments had to be undertaken to be able to properly simulate such transients, as needed forthe safety case of these reactors. Numerous national and international experimental programmes
provided the data necessary for understanding the phenomena and simulating them.
The CSNI has always considered thermal-hydraulic code validation as well as theexperimental database needed for such validation with great attention. The previous Senior
Expert Group on Safety Research (SESAR) reports give an overview of the large number of
separate-effect and integral test programmes that have been carried out in the past (NEA, 1993;
1996; 2001; 2007). The results from these programmes provide a sound basis for model validationof traditional system codes, whereas they are insufficient for multidimensional and, especially,for computational fluid dynamics (CFD) codes.
Extensive research programmes in thermal-hydraulics were carried out from the 1970s to
the 1990s and contributed to confirming the safety of existing reactors. Once these objectives
were achieved, the necessity for some of these activities diminished, as did the available national
funding streams. As a result, several large-scale facilities and even laboratories all over the world
have been closed or have largely reduced their activities. On the one hand, the availability of datafrom those research programmes for future use became difficult or questionable; on the other
hand, the expertise of experimentalists leading those programmes is diminishing.
In the early 2000s, the CSNI took the initiative to support safety-relevant thermal-hydraulicfacilities that were in danger of closure. This was done through the establishment of international
projects addressing issues of broad international interest and centred on the technical capabilities
of selected facilities. Some of these projects have been completed (SETH, PSB-VVER, ROSA-1 and 2),
while others are still ongoing (PKL-1 to 4, ATLAS-1 and 2). From PKL-2 onwards, the PKL tests werecomplemented by experiments performed in the PMK (Hungary), ROCOM (Germany) and, more
recently, PACTEL (Finland) facilities.
The SETH project focused on the capabilities of the PKL and PANDA facilities, which wererecommended for international consideration (NEA, 2009a). The PKL experiments addressed theissue of potential boron dilution accidents in PWR reactors; this subject figured prominently on
the programme of the follow-up PKL projects as well. The PANDA experiments provided dataon containment three-dimensional (3D) gas flow and distribution that are important for code
prediction capability improvements, accident management and design of mitigating measures.
The PSB-VVER project had the objective of providing unique experimental data needed for thevalidation of thermal-hydraulic codes used for the safety assessment of VVER-1000 reactors. In
the framework of the ROSA-1 and ROSA-2 projects, different issues were investigated, viz.
temperature stratification and mixing, water hammer, natural circulation in the primary circuit

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and different LOCA cases (pressure vessel break, intermediate breaks and steam generator tube
ruptures [SGTR]).
From the beginning, the different phases of the PKL project addressed the consequences ofinherent boron dilution and proposed countermeasures to avoid critical situations, both during
small-break LOCA and accidents in cold shutdown conditions. Cool down under asymmetricnatural circulation conditions and fast cool down transients following a main steam line break
were also evaluated (PKL-2). After Fukushima, beyond design-basis accidents became of primaryinterest. Accident scenarios such as station blackout (SBO), LOCA with additional system
failures leading to significant core heat-up, became the central topics of the OECD/PKL-3 project.
At the same time, the efficiency of using core exit temperature measurements to diagnose
severe core heat-up was also investigated. The PKL-4 project, currently underway, is performing
LOCA-related basic parameter studies (quench front propagation, swell level investigation) forcode model development and validation, but is also expected to come to a final conclusion on
inherent boron dilution and on boron precipitation following large-break LOCA. The PKL tests
have been complemented by experiments performed in other facilities with the aim toinvestigate the system behaviour of different type of plants (PMK for VVERs) or the scale effect(PACTEL), but also more detailed 3D processes (ROCOM).
The first OECD ATLAS project started in 2014 and addressed thermal-hydraulic safety issues
and accident management issues relevant for water reactors, by means of experiments in the
Korean ATLAS integral effect test facility. Prolonged SBO with active or passive secondary
cooling, small-break LOCA during SBO and total loss of feedwater assuming additional failures
were investigated and, to address the scaling issue, two counterpart tests with ROSA were
performed. ATLAS-2, initiated in 2017, is investigating safety key issues. Specifically, it will
evaluate: long-term coolability with a failure of the residual heat removal system; passive coremakeup with either a hybrid safety injection tank or passive emergency core cooling system
during SBO; intermediate break of LOCA either at the pressuriser surge line or the direct vessel
injection line; design extension condition scenarios such as steam line break followed by SGTR;
and scaling issues by conducting a counterpart test for the previous reactor pressure vessel head
break LOCA conducted at LSTF. Strong co-operation has developed among the ROSA, PKL and
ATLAS projects in the past decade, materialising in counterpart tests and joint analyticalworkshops, the latter reflecting the strong commitment of the participants to perform code
validation in parallel to the running projects. A computer code benchmark has always been apart of these projects assuring a fruitful co-operation between experimentalists and analysts.

Scope
As the scenarios of primary concern shifted from the large-break LOCA to small and intermediate
breaks as well as to other incidents (e.g. boron dilution), the thermal-hydraulic research effort
shifted accordingly to cover the more complex phenomena associated with this category ofaccidents. Improved computational tools were also developed to properly handle these. Although
reactivity-related accidents and transients were, of course, considered from the beginning of the
deployment of LWRs, increasing emphasis has been put on the accidents having a neutronic
origin or a strong neutronic aspect. Accordingly, it has been realised that multidimensional,
coupled thermal-hydraulic/neutronic computations were needed to reduce the conservatism of
the earlier simpler analyses and/or to simulate properly some complex situations. Althoughmany of the existing facilities are not sufficiently instrumented to be used to validate finely
detailed analysis tools (e.g. CFD codes), they are included in this section for completeness.
The new concerns that regulators faced after the Chernobyl accident, more generally
related to the understanding and simulation of situations and phenomena in reactors designedin the former Eastern bloc countries, provided additional needs for research and development.
The emergence of advanced LWRs having passive safety systems opened another new area ofless known phenomena and situations that had to be addressed.

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3. TECHNICAL ISSUES AND ASSOCIATED FACILITIESThe Fukushima accident raised a number of questions, but these concerned mainly issues
or conditions which could lead to severe accidents. However, it also pointed out the potential
vulnerability of spent fuel pools (SFP). In order to address this question, a joint activity by theWorking Group on Fuel Safety (WGFS) and Working Group on Analysis and Management of
Accidents (WGAMA) produced a status report on SFPs, followed by a phenomena identification
and ranking table to identify the most important phenomena (NEA, 2015b; 2017). One of theconclusions was that SFP behaviour is significantly influenced by 3D effects, which need further
experimental support for model development. The Fukushima accident also highlighted theneed for understanding the reliability of cooling systems connected with the primary loop
(e.g. the isolation condenser and the reactor core isolation cooling) and improved modelling via
coupled thermal-hydraulic and mechanical modelling. Finally, Fukushima reinforced thealready ongoing efforts to evaluate the consequences of beyond design-basis accidents: as aresult, design extension conditions have to be taken into consideration in several countries inthe design of nuclear power plants, and has improved accident management concepts. This has
prompted the need for better modelling tools, particularly for new builds and reactor designs,
as well as for less explored accident sequences.
System codes have reached maturity over the past two decades, but the demand to reduce
safety margins, for example due to power upgrading of existing reactors, may also require
refinement and further validation of existing analytical tools and additional experimental data.
The better understanding of 3D processes within the primary system also calls for more precise
experimental data for code validation purposes; this is also underscored by an ongoing WGAMA
activity: Status on Simulation Capability of 3D System-scale T-H Analysis Codes (3DSYSTH).
Uncertainty and sensitivity assessments have become increasingly important in thermalhydraulic analysis. In the last decade, a number of CSNI activities have addressed this issue(BEMUSE, PREMIUM, SAPIUM) with goals of reviewing available methods, assessing theircapabilities and identifying further needs. It is important to note that “best estimate plus
uncertainty” (BEPU) modelling requires, in many cases, more reliable test data, including reduced
or well-understood measurement uncertainties.
CFD has made good progress in the last decade supported by computer development, and theuse of CFD tools for nuclear safety applications has expanded considerably. The growing interestin CFD applications is also reflected by the high popularity of WGAMA actions in the field, like theseries of CFD4NRS workshops and of benchmarks. The move towards uncertainty quantificationof CFD results represents an important milestone in the area. All these developments have led tosingle-phase CFD being increasingly used in support of safety studies.
The Scaling State of the Art Report (S-SOAR), issued in 2017 (Bestion et al., 2017), constitutes acornerstone for harmonising the knowledge and the understanding in relation to a vital issue
in nuclear reactor safety. Scaling activities were considered essential in the 2001 SESAR (NEA,
2009a) report and scaling is part of most of the thermal-hydraulic issues identified in the 2007
SESAR report (NEA, 2007).
Based on the most important safety issues and a review of available experimental data tocover them, a list of prioritised thermal-hydraulic issues was proposed in 2007 (NEA, 2007). It is
worthwhile to review them in order to measure the progress made in the last decade.
In a number of cases, CSNI projects or activities were initiated to address the issues:

44•boron dilution has been investigated within several PKL projects, producing useful
information both for safety evaluation and for code validation;•

mixing in cold legs and downcomer was extensively investigated in ROSA and ROCOM(as part of the PKL project);•a WGAMA activity (3D-SYSTH) is presently reviewing the needs as required by 3D system
codes;•

shutdown accidents have been one of the important items in the ROSA, PKL and ATLAS
projects;•

the ECCS strainer clogging issue was investigated by a CSNI Task Group (NEA, 2013);

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3. TECHNICAL ISSUES AND ASSOCIATED FACILITIES•a CFD benchmark was performed to study the maturity of CFD codes for the analysis ofthermal cycling (NEA, 2011c);•a recently started WGAMA activity will produce a status report in the field of passive
system performance;•a joint WGFS/WGAMA activity produced a status report on vulnerabilities of spent fuelpools.

It is expected that when these activities are completed, decisions can be taken as to whether
to close these issues or to make recommendations for further actions.
No CSNI activities have been launched in the past decade on BWR stability or PHWR thermalhydraulics. However, extensive studies have been performed for PHWRs by the CANDU Owner’sGroup and the Atomic Energy of Canada Limited for a series of postulated accident scenarios
ranging from design basis to station blackout events using RD-14M. In addition, moderator
cooling and flow experiments were performed to evaluate moderator analyses tools, and a seriesof experiments were performed to derive analysis parameters for critical heat flux and post
dryout behaviour for various CANDU bundle types, with various degrees of diametral creep andaxial power profiles. Freon experiments were also performed to support the high-temperatureexperiments in water, and a header facility was built to evaluate the 3D water flow through a
CANDU core. Fukushima has also prompted evaluation of spent fuel cooling in CANDU designs.
In spite of these continuing developments, often conducted internationally, a number ofissues still require some attention. New issues will certainly also arise in relation to the designand safety analysis of future reactor systems and the need for design-basis extension and severe
accident analyses. For example, as current plants continue to make operational changes(e.g. power uprates), analysis will be needed to assess changes in safety margins and plant
response to off-normal conditions. Also, the increasing use of risk-informed regulation will
require better tools and data. The following sections summarise the issues of current andnear-term interest in the thermal-hydraulics area.
The remaining issues from 2007 are considered to still be pertinent, even if a slight shift intheir content may be observed.

Description
Table 3.1 shows the list of current thermal-hydraulic safety issues that has been derived.
It includes safety-related phenomena, accident scenarios and parameters. The list is based on
the corresponding table of the previous SESAR report (NEA, 2011c), on a recent list of important
thermal-hydraulic phenomena (Aksan, D’Auria and Glaeser, 2018; D’Auria et al., 2017) and on
discussions of the members of the present SESAR group. The first column lists the high-priority
issues; however, no priority is assigned to the row number. The second column provides a
description of each issue and the third the motivation for inclusion in the present list.
Identification of the facilities to resolve the issues is given in column 4 and additional notes areprovided in the last column.

3.1.2. Nuclear fuelIntroduction
The fuel of a nuclear reactor is tied directly to the economic performance, investment protection
and safety of the nuclear power plant. Fuel that performs well means fewer plant shutdowns, lessradioactive contamination, less radiation exposure to operations and maintenance personnel, andless potential for releases of radioactive material offsite. The role of fuel performance in nuclear
power plant safety can vary depending on plant design and technology. Plants are generally
designed to prevent fuel damage for events that are expected to occur one or more times during
the life of the plant and for more rare events, varying degrees of fuel damage may be allowed.

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3. TECHNICAL ISSUES AND ASSOCIATED FACILITIES

Table 3.1. Current thermal-hydraulic safety issues
Issue

Description

1) Anticipatedtransient without
SCRAM (ATWS)

Facilities

Notes

Core heat transfer is
Each anticipated operational
occurrence (AOO) may cause an concerned. The
ATWS. ATWS typically evolve at corresponding
thermal-hydraulichigh pressure and high linear
phenomenon is “nuclear
power. Situations of interest
include void collapse following thermal-hydraulicfeedback”. Control rod
pressure increase in boilingwater reactors and may include ejection ATWS (seeboiling transition during AOO. Issue 13) is not considered.
See also Issue 14.

SPES, PMK, HIDRA. Facilities
characterised by full power(and full linear power) in
addition to full height andfull pressure scaling areneeded to simulateconditions expected after
ATWS.

An attempt could be madeto ask Russian scientists
about the PSB facility inElectrogorsk and to use LOFTATWS experiments.

2) Natural
circulation (NC)

NC is established when a heat
In the case of boron
sink and source co-exist at
dilution, re-criticality
different elevations. Sub-topics might be of concern
include: stratification; flow
during a few days after thestability; reliability of passive
start of each fuel cycle
systems; application of(high boron concentration
computational fluid dynamics
in the primary system).
(CFD); boron dilution; CANDU
Need to model parallel
NC between core and steam
U-tubes to catch flow
generator (SG); 3D effects
reversal and stability
including flow distribution and phenomena. NC is a key
mixing in CL and inphenomenon in case of
downcomer; over-spilling effect steam generator tube
or siphon condensation; reflux ruptures in Section 3.1.4.
condensation;
non-condensable gas effect;
horizontal heated channels andstratification of two-phase flow;
NC in the presence of multiple
interacting (passive) systems.

PKL, ATLAS, PACTEL, LSTF,
FINCLS. Full height scaling ispreferred for NC studies
either in pressurised water
reactor (including AP-1000)
and in boiling water reactor
conditions.

Several facilities are suitablefor NC-related investigations.

3) Interaction
primary-system(PS)/containment(CO)PS and CO interact at thermalhydraulic level in case of mostof the accident scenarios. The
occurrence of two-phasecritical flow (TPCF) at the break(e.g. in case of a loss-of-coolant
accident) keeps PS and CO
decoupled. However, whenconditions for TPCF disappear,a tight interaction isestablished. The conditions of
TPCF disappearance is ofinterest, as well as long-termcooling, including sumprecirculation and potential for
clogging and pump cavitation
due to debris. TPCF throughvalves remains of concern since
the writing of the CCVM on
Separate Effects Test Facilities(non-pipe breaks). Additional
interest derives from thepossible occurrence ofsupersonic conditions indownstream valves discharging
in pools. Additional importancefor this issue is expected invarious SMR and AP-1000
accident scenarios.

46

Motivation

Few test data are available(somewhat ignored),e.g. from the Bubble
Condenser facility in
Russia (TACIS/PHARE
project) and the multiapplication small lightwater reactor (MASLWR)
ITF at Oregon State
University (United States).
Interaction tests with PS
and CO are planned in
ATLAS, in which a newcontainment is underconstruction.

MASLWR (United States),
The Bubble Condenser
ATLAS-CUBE (Korea),
facility in Russia could beinteractions between PS and considered.
CO are of interest in
experiments investigating
two-phase critical flow andtransition to Bernoulli flow.
Condensation, spray, stored
energy distribution ofinternal structures,compartment arrangement
are relevant phenomena andshould be instrumented.
Reverse flow from CO to PS
in the long-term coolingphase, resulting in clogging
and N2 ingress also ofinterest.

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3. TECHNICAL ISSUES AND ASSOCIATED FACILITIES

Table 3.1. Current thermal-hydraulic safety issues (cont’d)Issue

Description

Motivation

Facilities

Notes

4) Coefficients for Pressure drops at geometricdiscontinuity (direct flow andpressure drop at
flow reversal) substantiallygeometric
discontinuity (KPD) affect any accident scenarios.
Under many circumstances,
errors in KPD result ininaccuracies in calculations
which are incorrectly attributed
to different origins and which
may lead to modification ofexisting code models which are
not responsible for thoseinaccuracies.

Limited data availability:
CFD application might beused to derive suitablevalues for KPD, but errors
are not negligible andderivations arecumbersome in two-phaseconditions. Experimental
data may be affected bylarge errors due to mixingof reversible andirreversible contributions.

PKL with high-quality
instrumentation (pressuredrop) is relevant. Reversible
and irreversible pressure
drop are of interest. Specificinstrumentation needed (atleast in one or a few
geometric discontinuity
locations).

All facilities have thepotential to be relevant inthis connection as long as
suitable instrumentation isavailable.

5) Main coolant
pump (MCP) trip
performance(mostly in
two-phase
conditions)

MCP affects LOCA and nonLOCA scenarios during thecoast-down phase and longterm NC cooling because ofpressure drops inside complex
geometries (KPD) and TPCF,
which are difficult tocharacterise experimentallyand may not obey known
scaling law.

Four-quadrant head
curves used in computercodes feature large
uncertainties, especially in
two-phase conditions.

Reactor Coolant Pump Test
(RCPT) Facility in Korea.
Large-scale and two-phaseconditions (including
suitable range of parametersinvestigated); separate
effects tests also relevant.

Other component
performances may constitutean issue, e.g. separator,
dryers, jet pumps, etc.

6) Shutdown
transient otherthan de-borated
water injection
and residual heatremoval failures

During shutdown, the reactor ispartially unprotected(e.g. removal of parts of PS) andmore exposed to operator
actions. Failure of the residual
heat removal system (RHRS)
can challenge the core
coolability. In the mid-loop
operation, air ingress into the
RHRS pump due to the hot-leg
water level decrease can resultin failure.

A variety of situations mayneed investigation:
experimental data areneeded and modellingcapabilities need to be
demonstrated. Some test
data are available from
LSTF, ATLAS and PKL.

LSTF (Japan), ATLAS (Korea),
PKL (Germany). Low
pressure, openings in the
primary loop and natural
circulation are key words forthe experiments.

7) Coolant flow
distribution
through horizontal
channels inpressure tube
reactors

In a PHWR, coolant is
distributed to the fuel channels
via individual feeder pipes froma distribution header. The
distribution of the PHWR
primary coolant through thechannels as a function ofageing (diametral creep), axial
power profile, bundle
deformation and header
distribution system needs to beassessed for a wide variety ofpossible scenarios andanticipated operational
occurrences.

8) Fluid structureinteraction (FSI)

Also including thermal cycling
leading to thermal-fatigue.
Thermal-hydraulic oscillation
may also trigger FSI, e.g. in caseof fuel rods or SG tubes.
Pressure wave propagation
phenomena may also beaffected by FSI, including loadon internals following a LOCA.
Turbulence and swirl formation
at elbow, bend, orifice, etc. may
affect erosion-corrosion of pipe
wall.

Full-scale, instrumented
water loops operated at
PHWR temperature andpressure, full-scale PHWR
freon loop full bundle
CANDU header facility
1/8 scale (Canada).

Motivation is related tonumerous problems withflow-induced vibration of
nuclear power plants. SG
tubes, fuel rod (or fuelassembly) and RPVinternals are of maininterest. In facilities,
well-defined boundary
conditions and simplified
geometry facilitating easymeshing is needed forcode (mainly CFD)
validation.

Small-scale burst test facility. The issue is more relevant forstructural integrity, see(Canada), MSUB facility
Table 3.6, Item 8. Use of freon(Canada).
to induce vibrations in steam
generator tube arrays.

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3. TECHNICAL ISSUES AND ASSOCIATED FACILITIES

Table 3.1. Current thermal-hydraulic safety issues (cont’d)Issue

Description

Motivation

Facilities

Notes

9) Passive system
performance

The installations of passive
systems poses new challenges
to safety, which are strongly
system dependent: leakages,
minor installations errors forhorizontal pipes, heat losses,
small amounts of
non-condensable gases.
Activity connected withreliability of passive systems.

Various sensitivity studies
are necessary in facilities
to confirm the reliability ofpassive systems (flowinstability, water hammer,
condensation, boiling,
common head effects,
heat exchanger inclinationof effects, and so on). Issue
as NC and KPD are also
applicable here.

PACTEL (Finland), LSTF
(Japan), INKA (Germany),
ATLAS-PAFS (Korea), PASCAL,FESTA (Korea).

Full height and full pressure
facilities are relevant, but
reduced height facilities arecomplementary.

10) Power
excursion duringloss-of-coolant
accident

Also covering header flow
distribution and moderator
sub-cooling and sub-cooling
distribution phenomena.

Associated with positive
void coefficient.

OMEGA loop (France).

OMEGA loop is a separate
effects test facility where
power supply (other thandecay power) during a LOCA
is investigated relevant to
CANDU, PHWR of Atucha andRBMK.

For simulation,
multidimensional analysis
tools and improvement of
subchannel codes areneeded. Experiments
focusing on
multidimensional
behaviour, and spacer
effects are needed to
improve primarily CFD
1-phase modelling of
mixing in-core.

MCT (Korea, Moderator
Facility). Stern Labs steamwater experiments, and CNL
Freon experiments (Canada),
study effects of spacers andappendages on flow anddeparture from nucleateboiling on PHWR horizontal
bundle.

“Old” data available (partly)from UPTF (Germany), SCTF
(United States) and CCTF
(Japan) large-scale facility.

11) 3D global void, The trend towards higher
power densities means that intemperature andopen cores, departure fromflow distribution
nucleate boiling in the
maximum loaded fuel bundle
and single rod may be
prevented only due to thecross-flow connections inside
the core. Especially for boilingwater reactors, new coregeometries such as partial
length rod and swirl-type
spacer significantly affect thecore heat transfer.

MEDEA, ASPIC, MIDITH
12) Spent fuel pool SFP behaviour, and deborated Both identified key AS(France).
became of interest after(SFP) accidents
water injection are of interest.
Because of large FP inventory in Fukushima. Phenomena
are still unclear during loss
the SFP, its cooling should beof cooling of SFP:
maintained in all possible
accident conditions, including convection, boiling in thepool, uncovering process
those with an uncovered rod,of fuel assemblies, steam
where H2 production, spray
aspiration, criticality due tocooling and 3D pool fluid
circulation are of interest. Water boiling, flow reversal, flow
injection using the fire engine is blockage, NC reduction byrecovery of cooling system,
also now considered as anspray cooling, CCFL, void
important AM measure when
distribution inside the fuelplant-wide damage occurs.
assembly, oxidation by air
A variety of situations involveand steam.
injection of deborated water.
13) Reactivity
insertion accident(RIA) including
control rod
ejection (CRE). Also
included under
fuel.

48

The cladding temperature
evolution plays an importantrole since it strongly affects themechanical strength of thecladding. The knowledge oftransient clad-to-coolant heattransfer coefficients are
uncertain and need to beverified by experiments. Suchheat transfer is also dependent
on the transient power change.
Sub-cooled boiling and boilingtransition are of interest for
CRE. Transient void formation is
also important to assess thevoid feedback effect.

For high-burnup fuels, the CABRI, BR-2, NSRR. Facilities
to simulate RIA conditions inenthalpy limit and PCMI
pool and/or convective
failure threshold are
reduced, resulting in safety boiling with various power
pulse variations are
margin reduction. Apartfrom fuel behaviour, from necessary to validate
the viewpoint of thermal- thermal-hydraulics models
hydraulics, applicability of and to reduce conservatism
the conventional post-CHF of the current regulatory
practice.
(critical heat flux) heattransfer models (CHF,
transition boiling and film
boiling) to the RIA
conditions is of interest.
Transient boiling data are
not sufficient in case of
CRE.

All part of the DENOPI
programme being executedby the IRSN.

There are facilities withsimulated rods, but theothers are research reactors
using fuel rods. There are
already some data from the
now closed PATRICIA.

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3. TECHNICAL ISSUES AND ASSOCIATED FACILITIES

Table 3.1. Current thermal-hydraulic safety issues (cont’d)Issue

Description

Motivation

Facilities

Notes

14) Heat transfer athigh temperature
with radiation(HTR)

Combination of convection andradiation (HTR) in the presenceof H2 production possibly withpartly damaged fuel (ballooned
or embrittled clad) is safety
relevant.

Experimental data not
sufficient. Ranges ofvariations of this key
parameter different from
Issue 15.

High Temperature Fuel
Channels Laboratories(Canada).

Revision of the acceptancecriteria for AOO wasproposed by the Atomic
Energy Society of Japan
in 2003. Investigation of
ballooning (including fuelrelocation and consequent
thermal power effect) withsimultaneous H2
production. The rewetting
mechanism at the quench
front and its effects on theheat transfer are still
unclear. Currently,
empirical correlations areused for the heat transfer
prediction, but new LOCA
rulemaking and adoptionof design extension
conditions need
integrated analysis of fuelperformance and thermalhydraulics. Fully coupledfuel behaviour andthermal-hydraulicexperiments are required.

QUENCH (Germany),
QUEST (Korea), ATHER(Korea), HIDRA (Japan),
RBHT (United States),
COAL (France).

At high temperature, radiative
heat transfer becomes more
dominant than convection witha magnitude dependent onemissivity of the oxide layer ofthe fuel rod or pressure tube.
Canada has a facility to measureemissivity from pressure tubes
as a function of oxide thickness.
QUENCH is used to study
reflood/quenching at partly
damaged fuel elements.
COAL focuses on the coolabilityof a partially deformed fuelassembly, particularly in theballooned area.
HIDRA has test sections of a fulllength 4×4 bundle and a shortlength 3×3 length bundle.
Studies focus on post-boilingtransition heat transfer.
The QUEST facility utilises IR
thermometry and a totalreflection visualisation method
to obtain temperature
distribution of the cladding.
RBHT consists of 7×7 full-lengthrods with a highly detailed
thermocouple distribution tomeasure quench profile anddroplet size measurements up
and down-stream of the spacer
grid.

15) Reflood/
Well-established phenomenon
quench front
occurring with the transition
propagation (QFP) from film to nucleate boiling orforced convection at low
pressure under large-break
LOCA conditions. Combinationof QFP with situation of
ballooning, H2 production, crudformation, etc. pose safetyconcerns. Core heat transfer
behaviour during AOO is ofinterest for boiling water
reactors in terms of theacceptance criteria of the safety
analysis, where the boilingtransition (BT) is allowed aslong as the fuel integrity is
maintained. Models to predictthe maximum temperature andduration of BT should bevalidated with the data, which
are not sufficient, especially forrewetting propagation.

The largest quantity of radioactive material in the plant is contained in the fuel in the form offission products and higher actinides. These build up over the lifetime of the fuel and, to varying
degrees, have the potential to be released from the fuel in the event of fuel damage. As such, theamount, timing and nature of fuel damage during accident conditions determine the amount,
timing and nature of fission products available for release in the plant and potentially outside theplant. These fission products and actinides have the potential to be released from the reactor core
into the reactor coolant system, containment and, ultimately, the environment, and affect thedesign and qualification of plant safety systems, site suitability and emergency preparedness.
After the shutdown of major facilities – Osiris (France), PHEBUS (France), Halden (Norway), theNational Research Universal (NRU, Canada) and the Japan Material Testing Reactor (JMTR, Japan)– there is a lack experimental facilities necessary for testing fuel performance and for safetyassessment. This is occurring at a time when a new generation of fuels are under development,
and at which there are open issues for current fuels which will still be in operation for decades.

Scope
The scope of this chapter includes safety issues associated with the performance of LWR(including VVER), ALWR and PHWR fuel. The fuel type is quite large, ranging from the “oldgeneration” of fuels, still in operation (UO2, UO2-Gd, MOX), with zirconium-based claddings and forwhich there are still pending issues, to accident-tolerant fuels, which include “evolutionary-fuels”(doped pellets with Cr coating cladding, for example) and “advanced-fuels (SiC/SiC cladding forexample).

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3. TECHNICAL ISSUES AND ASSOCIATED FACILITIESDescription
Fuel designs vary by reactor type and technology. This section gives a summary of the fuel byreactor type.
LWRs (including VVERs): For LWRs, the uranium enrichment ranges from 2% to nearly 5%,although in some countries LWR recycled mixed oxide fuel is being used in portions of the core.
Burnable poison (such as Gd2O3) may also be present in the fuel to compensate for excess
reactivity at the beginning of life. The UO2 pellets (or mixed oxide pellets) are housed inzirconium alloy tubes (called cladding) 12-14 feet long. These tubes are then assembled into tube
bundles of varying sizes, which are then inserted into the reactor core in a vertical orientation.
The fuel design (U enrichment, cladding material, service condition, etc.) determines how longthe fuel can stay in the reactor and still be able to perform satisfactorily under steady-state andaccident conditions. For currently operating LWRs, the accidents that pose the greatest potential
for fuel damage are large reactivity insertions, loss-of-coolant events, and events that cause dryout or departure from nucleate boiling on the cladding. The behaviour of the fuel under these
conditions will be affected by the fuel burnup level (which affects internal fuel pin pressure),
the condition of the fuel cladding (e.g. corrosion, oxidation and embrittlement) and the locationin the core. For economic reasons, increases in burnup levels are being considered in many
countries. To support higher burnups, new cladding materials are being developed that exhibit
reduced oxidation characteristics. The safety issues associated with current LWR fuel are all
related to deciding where to establish safety limits based on how fuel performance under
accident conditions changes with changes in burnup, cladding material and service condition.
ALWRs: For ALWRs, it is expected that high-burnup levels will be desired for economic andpossibly security reasons (e.g. proliferation resistance). To support such fuel designs, burnablepoisons and additional advances in cladding materials will likely be needed. Ensuring that these
new fuel designs achieve an acceptable level of safety will require testing and analysis toconfirm fuel performance, validate analysis tools and establish safety limits. Such testing will
range from analyses and unirradiated clad testing to more extensive testing, depending on thefuel design and the types of accidents that may be postulated.
PHWRs: PHWR fuel has many of the same basic characteristics as LWR fuel. The majority of
the fuel is UO2 in composition, but some PHWRs use recycled fuel where the plutonium oxide wasnever separated from the uranium oxide. There are also reactors utilising thorium oxide to flatten
the power profile of the cores on a regular basis. PHWRs utilise a shorter fuel pin length (1.5 feetversus 12-14 feet), smaller pin bundles, low enrichment or natural uranium, and horizontal
orientation in pressure tube core geometry. Generally, PHWR fuel is designed for lower burnups
and higher linear powers than LWR fuel; however, the safety issues and performance concerns
are essentially the same. PHWRs operate using an online refuelling system and thus the fuelhandling (and potential refuelling accidents) of PHWR fuel is very different than that for LWRs.
With the exception of a new requirement for some countries to evaluate the impact of fuelfailures on transient behaviour, and concerns about fuel integrity during interim storage, thefuel issues listed in Table 3.2 for existing fuel designs are the same as those compiled during
the last writing of this report and all issues remain relevant today. A 2012 report on fuel safety
criteria (NEA, 2012) states:“Complete or sufficient information is not available for a number of issues discussed
in this report. These include CRUD deposition, cladding oxidation and hydriding, rodinternal gas pressure, pellet-cladding and thermal-mechanical loads, fuel melting,fuel fragmentation, cladding embrittlement, gap activity, radioactive source term,high burn-up, mixed oxide fuel, slow or incomplete control rod insertion, axial offset
anomaly, cladding elongation and cladding stability. Under the auspices of the OECD
Nuclear Energy Agency, active research is being conducted in many of these areas
through programmes including the Halden Reactor Project in Norway, the Studsvik
Cladding Integrity Project in Sweden, and the CABRI International Project in France.
These issues have been and will continue to be addressed by the Working Group on
Fuel Safety, as directed by the NEA Committee on the Safety of Nuclear Installations.”

50

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3. TECHNICAL ISSUES AND ASSOCIATED FACILITIES

Furthermore, a more recent state of the art report on LOCA written by the Working Groupon Fuel Safety (NEA, forthcoming) has made several recommendations, summarised below:•

Ductility-based LOCA criteria derived from ring compression tests of double-sided
oxidised specimens have been proposed by the US Nuclear Regulatory Commission toensure long-term coolability of the core, whereas fracture-based LOCA criteria derived
from integral rod testing have been chosen by the French regulators to ensure that thefuel rods are not excessively embrittled and maintain long-term coolability. Bothapproaches take into account the burnup effect through the level of hydrogen that has
been picked up by the cladding prior to the transient. The significant differences between
experimental techniques used in different countries to establish LOCA safety criteria callfor further discussions between experimentalists and regulators. Development ofstandard LOCA criteria procedures could be a future objective.

•

Fuel fragmentation, relocation and dispersal during a LOCA, observed in several in-pileand out-of-pile tests, needs further investigations. The effects of pellet burnup, thepresence of additives in the pellet, cladding corrosion state, linear power during normal
operation and axial gas transport phenomena should be addressed in new experiments.
Fuel fragmentation, relocation and dispersal is more important today than breakaway
oxidation of the cladding, and thus its impact on cladding temperature should be better
assessed. As-fabricated cladding was used in most of the oxidation tests, but specific tests
have been performed to demonstrate that pre-hydrided/pre-oxidised and sometimes
thinned claddings are sufficiently prototypical. As such, a significant amount of new data
has been accumulated on the effects of irradiation, alloying elements, pre-oxidising andpre-hydriding of the cladding, steam purity (addition of nitrogen or air), and high external
steam pressure. Based on these findings, new licensing frameworks have been defined inmost member countries. Any new cladding material has to be studied accordingly.

•

It is important to continue to develop predictive computer codes associated withuncertainty analysis methods, to capture quantitatively the increased understanding of
the phenomenology. A specific effort should focus on multi-rod configurations to predict
core coolability.

Finally, since the last SESAR report was written, significant attention has been given
worldwide to the development of accident-tolerant fuels, and advanced reactor designs,
necessitating an additional need for irradiation testing of potential fuel designs from performanceand safety perspectives. A thorough State of the Art Report on Accident Tolerant Fuels was publishedby the NEA in 2018 (NEA, 2018b) and versatile irradiation options for evaluation of such fuels andconcepts were identified. Most of the facilities identified in the last SESAR report were listed, with
BR-2, ATR, HFIR, CABRI, Halden, Hanaro, NSRR and HFR remaining on the list as available reactors
at the time. Added to the list are the TREAT facility, recently refurbished, the LVR-15 research
reactor in the Czech Republic, the Jules Horowitz Reactor in France (under construction) and the
China Mianyang Research Reactor (CMRR, People’s Republic of China). NRU and Halden were bothshut down in 2018, and OSIRIS and JMTR were shut down in 2016.
It is encouraging to note that testing of entire assemblies of accident-tolerant fuels in an
operating power reactor has recently been announced. Advanced fuel assemblies featuring
chromia-doped fuel pellets and chromium-coated fuel cladding developed by Areva NP under
the US Department of Energy’s (DOE) Enhanced Accident Tolerant Fuel (EATF) programme, havebeen loaded into the Vogtle 2 plant in Georgia, which commenced operation on April 3 after arefuelling shutdown.
Table 3.2 lists fuel issues that are still considered to be of high or medium importance, alongwith the facilities available to address these issues. A comprehensive listing of research reactors
worldwide capable of addressing these issues can be found in Table 3.3.

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3. TECHNICAL ISSUES AND ASSOCIATED FACILITIES

Table 3.2. Current nuclear fuel issues and facilities
Issues, scenarios,
phenomena

52

Description

Facilities

Response to lossof-coolant
accidents (LOCAs)

As fuel designs change (minor evolution or advanced technologies
for fuel pellets and claddings), the response under LOCA conditions
needs to be investigated to support development of appropriate
criteria that ensure coolable geometry is maintained during and after
design-basis LOCA events. Experimental data are needed, consistent
with the design basis, and perhaps beyond the design-basis LOCA.
Also, small- and intermediate-break LOCAs can lead to clad
embrittlement. Some experimental data on small-break LOCAs areuseful to confirm fuel clad condition after such events and such
experiments are underway. Fuel fragmentation, relocation anddispersal inside the primary circuit, cooling of area with ballooning
fuels have to be included in research investigations.

ANL Hot cells (United States)
CABRI and CEA Hot Cells(France)
JAEA and Hot Cells (Japan)
MIR and Hot Cells (Russia)
RIAR and Hot Cells (Russia)
Studsvik Hot cells (Sweden)
TREAT and Hot Cells(United States)

Response toreactivity insertion
accident

As fuel designs change (minor evolution or advanced technologies forfuel pellets and claddings), the response to reactivity insertion accidents
needs to be investigated. Failure modes and criteria associated withnew cladding materials and new fuel are currently being investigatedthrough experimental programmes, consistent with design-basis andbeyond design-basis events. Additional focus may be needed on fuelfragmentation and dispersal after boiling crisis.

BIGR (Russia)
CABRI and CEA Hot Cells(France)
NSRR and Hot Cells (Japan)
TREAT (United States)

Response to power
oscillation events

As fuel designs change to achieve higher burnup (e.g. through theuse of new cladding material), the response to power oscillation
events (such as could occur due to an anticipated transient withoutsram or a loss of stability) needs to be determined. Experimental data
are needed to establish failure modes and limits; however,conducting experiments simulating these conditions is difficult.

CABRI (France)
NSRR (Japan)

Fuel performance
under steady-state
conditions

Fuel performance under steady-state conditions can affect coolant
circulating activity, which impacts the dose to operating personnel.
The performance of new and changes in fuel and clad properties
during operation can also affect the ability of the fuel to withstanddesign-basis accidents. Therefore, the performance of new fuel understeady-state conditions is important to ensuring the safety ofoperating personnel and to understanding and predicting fuelperformance during design-basis accidents (i.e. the condition of thefuel and cladding resulting from steady-state operation represents
the initial conditions for transients).

ATR (United States)BR-2 (Belgium)
Hanaro (Korea)
HFR (Netherlands)

New materials
properties (fuelproperty database)

Improving plant performance in both existing and future plants will
likely include improving fuel performance with respect to higher
burnup and power levels. These improvements will require new
cladding materials and the use of burnable poisons. The performanceof these new materials and poisons will need verification to ensure
their safety and to establish a fuel property database for safety
analyses.

ANL (United States)BR-2 (Belgium)
Hanaro (Korea)
HFR (Netherlands)
TREAT (United States)

Response of
Defective rods now have to be taken into account in safety
defective rods toassessments. Experimental data are needed, consistent with thetransient (reactivity design basis.
insertion accident,
steam generator
tube ruptures)

CABRI (France)
NSRR (Japan)
TREAT (United States)

Fuel integrity
during interim
storage

ANL Hot cells (United States)
CEA Hot Cells (France)
CNL Hot Cells (Canada)
IFE Hot Cells (Norway)
JAEA Hot Cells (Japan)
Russia Hot Cells (Russia)
Studsvik Hot Cells (Sweden)

As final disposal of spent fuels delays, interim storages under dry
condition and wet condition are expected to increase. Behaviour offuel under the interim storage conditions needs to be investigated tosupport the regulatory requirements that ensure no release ofradioactive materials from interim storage cask. The expected issues
are generation of hydrogen gas by radiation decomposition under
wet condition, cladding failure due to thermal creep and hydride
reorientation under dry condition, etc.

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4.0E+11
Pulse: 1.0E+17

4.0E+14

6.0E+147.0E+13

2.5E+15

1.0E+15

Maximum thermal
flux (n/cm2/s)
E<0.625eV

2.0E+141.0E+14

1.7E+141.0E+15
6.0E+14(E>1MeV)

5.0E+142.0E+14(E>1MeV)

Maximum fast
flux (n/cm2/s)
E>0.1MeV

1959
Refurbished 2018

1967/
≥2065

1966/
≥2056

1975/
≥2050

1965/
≥2050

1967/
≥2040

Initial criticality/
design life

7.2 cm dia.
61 cm height(4.3E+14)

13.7 cm dia.
122 cm height(1.0E+15)

Largest thermal
flux test volume(n/cm2/s)

0 loops
10 in-core positions
7 reflector positions
18 beam ports
Static capsules only

0 loops
3 in-core positions
12/3 reflector/pool
positions
6 beam ports
Static capsules only

8.89 cm dia.
73.7 cm height(4.0E+14)

13.6 cm dia.
61 cm height(6.0E+14)

1 loop4.57 cm dia.
3 in-core positions
55.9 cm height(3.6E+13)
9 reflector positions
9 beam ports
In-core flow loops at PWRor BWR conditions, HTGR
materials loop up to
1 600°C, gas-filled static
capsule with
instrumentation
available

0 loops
37 in-core positions
42 reflector positions
4 beam ports
Irradiation temperature
up to 1 200°C

6 loops
0 channels
47 in-core positions
24/36 reflector/pool
positions
0 beam ports
PWR loops

Irradiation
capabilities/test
conditions

8.89 cm dia.
73.7 cm height(2.0E+14)

13.6 cm dia.
61 cm height(6.0E+13)

4.57 cm dia.
55.9 cm height(1.2E+14)

7.2 cm dia.
61 cm height(1.3E+13)

13.7 cm dia.
122 cm height(5.0E+14)

Largest fast flux
test volume(n/cm2/s)

Source: IAEA Research Reactor Database.

Note: PWR: pressurised water reactor; BWR: boiling water reactor; GCR: gas-cooled reactor; SFR: sodium fast reactor; VVER: water-water energised reactor.

Graphite, pulse

0.08Light water tank 10

University ofMissouri Research
Reactor (MURR)
United States

Transient Reactor
Test Facility (TREAT)
United States

Light water tank 6
with heavy
water outertank

Massachusetts
Institute ofTechnology Reactor– II (MITR-II)
United States

20

Light water tank 85

High Flux Isotope
Reactor (HFIR)
United States

US National Bureau Heavy waterof Standards Reactor tank
(NBSR)
United States

Light water tank 250

Advanced Test
Reactor (ATR)
United States

Thermalpower (MW)Type

ReactorTable 3.3. Fuels and materials reactors worldwide

Yes

No

No

No

Yes

Yes

Fuel
irradiation?

Yes

No

No

No

No

Yes, limited

Transients?

3. TECHNICAL ISSUES AND ASSOCIATED FACILITIES

53

54
2.7E+145.7E+15Light water tank 45

140

60

Fast, sodium
cooledFast breeder,
sodium cooled

JOYO
Japan

BOR-60Russia (RIAR)

3.7E+15
5.0E+14(E>1MeV)

4.0E+15

5.1E+142.2E+14(E>1 MeV)

7.0E+14

7.34E+13
Pulse: 5.87E+16

1.0E+15
5.5E+14(E>1 MeV)>3.0E14

Maximum fast
flux (n/cm2/s)
E>0.1MeV

0 loops
19 in-core positions
12 reflector positions
0 rabbits
12 beam ports
PWR, BWR, GCR

1968 (operating
license until 2020)

0 loops
11 channels
15 in-core positions
10 reflector positions
0 beam ports
SFR

1977
0 loopsCurrently shutdown, 5 channels
restart in 2021
21 in-core positions
1 reflector positions
SFR

1961

1 loop underconstruction
80 in-core channels
0 rabbits
0 beam port
15.7 MPa, 340°C
PWR

1 pressurised water loop
(PWR, BWR, VVER)
1 channel

1963
Totally refurbishedin 2017
1961/ projected life
is 2026 or beyond

1 corrosion loop
10 in-core positions
26 reflector positions
0 rabbits
0 beam ports
PWR, BWR, GCR, SFR

Irradiation
capabilities/test
conditions

Expected in 2022
(50 years)

Expected in 2020

Initial criticality/
design life

60 cm height(2.9E+14)

90 cm height
8.0 cm dia.
20 cm dia.

4.4 cm width, 45
cm height

60 cm height
Fuel bundle-sizedcapsules

60 cm height(1.8E+14)

10 cm in-core
position
20 cm dia.
reflector position

5 cm dia.
12 cm length>3.0E+14

8 x 8 cm section
with 65 cm length>1.0E+14
10 cm in-core
position
20 cm dia.
reflector position

Largest fast flux
test volume(n/cm2/s)

Largest thermal
flux test volume(n/cm2/s)

Source: IAEA Research Reactor Database.

Note: PWR: pressurised water reactor; BWR: boiling water reactor; GCR: gas-cooled reactor; SFR: sodium fast reactor; VVER: water-water energised reactor.

2.0E+141.0E+15High Flux Reactor(HFR)
Netherlands

2.65E+13
Pulse: 2.12E+16

5.5E+14Light water tank 100

100

3.0E+14

Belgium Reactor-2(BR-2)
Belgium

Tank in pool

Jules Horowitz
ReactorFrance30

Maximum thermal
flux (n/cm2/s)
E<0.625eV

Light water pool 25
Pulse 20 GW

Light waterRA-10
Argentina

Thermalpower (MW)CABRIFranceType

ReactorTable 3.3. Fuels and materials reactors worldwide (continued)

Yes

Yes

Yes

Yes

Yes

Fuel
irradiation?

Yes

Yes

Yes

Yes

?

Transients?

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4.8E+15

1.5E+142.6E+14Light water pool 0.3

Light watertank, heavy
water reflector

Air cooled

Fast, power,
sodium cooledLight water tank 10

Light water
TRIGA dual
cores

Nuclear Safety
Research Reactor(NSRR)Japan

PIK
RussiaBIGR
RussiaMultipurpose Fast
Research Reactor(MBIR)
Russia

LVR-15 REZ
Czech Republic

TRIGA-II
Romania

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5.0E+14

1.8E+143.0E+14

5.3E158.0E+14

Fluence bypulse: 9.6E+14(n/cm²)

1.0E+14

6.0E+14(E>1MeV)

2.0E+15

Maximum fast
flux (n/cm2/s)
E>0.1MeV

1980

1957

BOR-60 replacement-targetCommissioning in
2020/
50 years

1977

2011
Full power 2019?

1975

1966/
>2020

1961

Initial criticality/
design life

1 loop8 channels
6 in-core channels
2 reflector channels
15.5 MPa, 280-300°C

4 loops
16 channels
2 in-core channels
2 reflector channels

3 loops (1 sodium loop)
14 channels
3 in-core channels
Inlet: 330-600°C
Outlet up to 850°C

1 loop2 channels

2 loops
17 channels
1 in-core channel
6 reflector channels

0 loop1 in-core position

7 loops
13 channels
11 in-core channels
20 MPa, 300°C, up to
27.8 kg/s

2 loops
1 channel
6 in-core positions
30 reflector positions
0 beam ports

Irradiation
capabilities/test
conditions

7.2 cm width
12.0 cm dia.
55 cm height

10 cm dia.
4.8E+1538 cm height

12 cm dia.

6.8 cm dia.

Largest thermal
flux test volume(n/cm2/s)

7.2 cm width
12.0 cm dia.
5.0E+15

4.1 cm dia.
8.0E+1438 cm height

12 cm dia.

6.8 cm dia.

Largest fast flux
test volume(n/cm2/s)

Source : IAEA Research Reactor Database.

Note: PWR: pressurised water reactor; BWR: boiling water reactor; GCR: gas-cooled reactor; SFR: sodium fast reactor; VVER: water-water energised reactor.

14

150

500
(Pulse 70 GW)

100

1.9E+12100

Light watercooled,beryllium
moderated

MIR-M1
Russia (RIAR)

5.0E+15

100

Light waterpressure vessel,
trap-typeHigh Flux Research
Reactor (SM-3)
Russia (RIAR)

Maximum thermal
flux (n/cm2/s)
E<0.625eV

Type

Reactor

Thermalpower (MW)Table 3.3. Fuels and materials reactors worldwide (continued)

Yes

Yes

Yes

No

Yes

Yes

Yes

Yes

Fuel
irradiation?

No

No

?

Yes

?

Yes

Yes

No

Transients?

3. TECHNICAL ISSUES AND ASSOCIATED FACILITIES

55

56

Light water

MARIA POLAND
NCBJ, Świerk
1E+14

4.5E+131E-9

1.7E+15

6.0E+142.0E+14

Maximum fast
flux (n/cm2/s)
E>0.1MeV

1974

1985

2014

1979

2010

1995

Initial criticality/
design life

Largest thermal
flux test volume(n/cm2/s)

6 horizontal channels
13 vertical channels

1 loop22 channels
5 in-core channels
2 MW pressurised water
test loop

1 loop11 channels
7 in-core channels>11 MPa, >300°C, ~8 kg/s

>1 loop
31 channels
4 in-core channels
9 beam ports

3 E+141 loop7.4 cm dia.
39 channels
1.95E+14,
E>0.82Me
7 in-core channels
25 reflector channels
10 MPa, 290°C at outlet,
26.5 kg/s (CANDU mode)

Irradiation
capabilities/test
conditions
7.4 cm dia.
4.30E+14

Largest fast flux
test volume(n/cm2/s)

Source: IAEA Research Reactor Database.

Note: PWR: pressurised water reactor; BWR: boiling water reactor; GCR: gas-cooled reactor; SFR: sodium fast reactor; VVER: water-water energised reactor.

3.5E+14

1.8E+14

Heavy water

DHRUVA
India

30

2.4E-14

CMMR
China

100

Light water tank 125

High Flux
Engineering Test
Reactor (HFETR)
China
6.2E+148.0E+14Light water tank 60

Maximum thermal
flux (n/cm2/s)
E<0.625eV

China Advanced
Research Reactor(CARR)
China

Thermalpower (MW)
4.5E+14TypeHigh Flux AdvanceLight water pool 30
Neutron Application
Reactor (HANARO)
Korea

ReactorTable 3.3. Fuels and materials reactors worldwide (continued)

Yes

Yes

Yes

Fuel
irradiation?

No

No

Transients?

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3. TECHNICAL ISSUES AND ASSOCIATED FACILITIES

3.1.3. Reactor physics
Introduction
Reactor physics issues are important to improve plants’ performance through power uprates,
higher burnup fuel, longer operating cycles, etc. Core configurations are becoming increasingly
heterogeneous in composition and distribution of power generation. This makes prediction ofcore behaviour and of safety parameters, such as reactivity coefficients, that dictate transient
behaviour more difficult. Experimental validation of neutronics methods is needed. In addition,
the use of advanced computational methods (e.g. 3D neutronics) to better refine safety analyses
and safety margins has emphasised the need for more detailed reactor physics data andexperimental confirmation of analytical methods. Also, thermal-hydraulic and neutronic codes
are being coupled to address issues such as boron dilution and ATWS and to analyse PHWR
pressure tube reactors.
The NEA Nuclear Science Committee (NSC) has activities in the reactor physics area and isalso concerned about the status of key facilities. Accordingly, this section was written in
co-operation with the NSC. However, to ensure an integrated approach to the preservation ofcritical facilities in this area, the NSC will take the lead to monitor facility status and recommend
appropriate actions for consideration by the NSC and the CSNI.

Scope
The scope of work in reactor physics covers the current and future needs of nuclear power
plants such as PWRs, VVER, BWRs, gas-cooled (thermal) reactors, PHWR reactors, gas-cooled(fast) reactors, liquid metal fast reactors, and molten salt reactors. The latter three are not thefocus of this report, yet experiments and research on these systems contribute to a wider andmore comprehensive validation of the models and computer codes used and to their further
development.
This section addresses reactor physics data (cross-sections, neutron spectra, reactivity
coefficients, etc.) and facilities to measure data for code assessment. Table 3.4 lists the issues
and facilities. Since the last SFEAR report, the PROTEUS, MINERVE and TCA facilities havestopped operation. PROTEUS (Switzerland) was a flexible facility that was capable ofrepresenting numerous types of reactors, such as LWRs, PHWRs and pebble bed modular
reactors, among others. The shutdown of MINERVE (France) meant that the number of facilities
with pile oscillators continued to diminish. Historically, pile oscillator experiments have beenused to accurately determine the reactivity worth of small samples to support applications suchas burnup credit and minor actinide burning; numerous comparisons had been done between
the DIMPLE reactor and MINERVE in the past through the CERES programme. The TCA supported
criticality assessments of potential fuel and materials for high-burnup configurations. VENUS
was converted from an LWR to an ADS fast reactor facility with lead/bismuth as the coolant, but
is returning to a configuration that can support LWR applications.
The International Atomic Energy Agency (IAEA) Research Reactor Database lists 238 operational
and temporarily shut down reactors as of May 2019. Many of these reactors are not suitable toadequately test the current reactor physics issues identified. Facilities which may have some
capabilities include Giacint (Belarus), IPEN/MB-01 (Brazil), Astra (Russia), CROCUS (Switzerland),
ZED-2 (Canada), STACY (Japan, temporarily shut down) and RB (Serbia, temporarily shut down).
Measurements from these facilities have been used as the basis for validation of criticality andreactor physics codes internationally. The status of the above facilities should be monitored
closely.

Description
Work in the area of reactor physics is of particular importance for the continued developmentof nuclear power. Key areas include:•

reactor core and fuel cycle physics issues at very high burnup and for enrichments higher
than currently used in LWRs;•

minor actinides recycling in LWRs;

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3. TECHNICAL ISSUES AND ASSOCIATED FACILITIES•

physics related to plutonium management in the medium term (before GEN IV systems
are deployed);•effects of radiation on reactor internals and the reactor vessel at high fluence fromcurrent operation and extended plant lifetime.

Support facilities for providing the data required for resolving these issues continue to be
essential. Integral data collected from past experiments carried out on now dismantled or still
existing facilities are not sufficient to cover the need of the evolutionary and next-generation
power systems. Specific new experiments are required, many of which can be covered byexisting facilities, provided they are maintained and refurbished.
The experimental facilities, research reactors and tests in power reactors need to cover themeasurement of the following parameters in critical and sub-critical configurations:•

neutron multiplication and K-effective;•

buckling and extrapolation length;•

spectral characteristics;•

reactivity effects;•

reactivity coefficients;•

kinetics measurements;•

reaction-rate distributions;•power distributions;•nuclide composition;•

shielding.

Computational models and codes have to cover core physics, coupled neutronics/thermalhydraulics, radiation shielding, criticality safety, physics of the fuel cycle, materials activation,
decay heating and energy deposition. The necessary basis in integral experimental data for model
development and validation must be available, maintained and expanded to meet requirementsfrom advanced reactors.
The NSC together with the OECD/NEA Data Bank, in collaboration with member countries
and other specialised institutions have developed databases with evaluated and qualified
experimental data shared internationally in addition to a large set of computer codes covering
the different needs in nuclear applications modelling. The databases cover:•basic nuclear and chemical thermodynamics data;•radiation shielding a dosimetry experiments (SINBAD);•

criticality experiments (ICSBE);•

reactor core and lattice experiments (IRPhE);•

data from coupled neutronics/thermal-hydraulics experiments and reactor operation;•

fuel behaviour experiments (IFPE);•

isotopic concentrations of spent nuclear fuel (SFCOMPO).

Basic data needs, such as improved capture cross-sections of certain absorbers – hafnium,
erbium and gadolinium – improved scattering cross-sections of oxygen, as well as improvementof yields of fission product isotopes in the fission of most heavy isotopes and decay schemes andenergy yields of radioactive isotopes, are required. In general, higher than current resolution
cross-section measurements from thermal energies to several MeV are required for a number ofimportant isotopes.
Such data will be useful in the evaluation of the accuracy of methods and codes throughverification, validation and qualification studies, and the measurements made in criticalfacilities and irradiation measurements in reactors play an essential role in the qualification
studies. The interpretation of experiments is a driving force for the continuous improvement ofcomputational methods and nuclear data.

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3. TECHNICAL ISSUES AND ASSOCIATED FACILITIES

More information on the activities and R&D needs identified by the NSC is provided in NEA(2003). An OECD expert group released a report in 2009 on the research and test facilities requiredin nuclear science and technology (NEA, 2009b).

Table 3.4. Current reactor physics issues
Issues and relevant
reactors

Description

Facility1) MOX fuel data:
PWR, BWR, VVER,
ALWR, PHWR

Reactor physics data to support the use of MOX fuel in current and future
reactors is essential to ensure safe operation. Such data include
cross-sections and their uncertainties, delayed neutron generation,
power distributions, decay heat production and power, temperature andvoid coefficients. This issue relates to the use of weapons-grade
plutonium (Pu) in LWRs, PHWRs and VVERs as well as the use of PWR
recycle Pu. Advanced fuel cycles involving Pu and other actinides are also
being studied for use in LWRs and PHWRs.

VENUS (Belgium)
ZED-2 (Canada)
CABRI (France)*
TREAT (United States)*

2) High-burnup fueldata:
PWR, BWR, VVER,
ALWR, APHWR

Reactor physics data to support the use of high-burnup fuel in current
and future reactors are essential to ensure safe operation. Such data
include cross-sections and their uncertainties, delayed neutron
generation, power distribution and power, temperature, and void
coefficients.

VENUS (Belgium)
ZED-2 (Canada)

3) Coolant void
coefficient:
PHWR, APHWR

Loss-of-coolant accident conditions can cause voiding in some PHWR
coolant channels. This voiding may lead to positive reactivity input prior
to reactor shutdown. The timing and degree of voiding (and thesubsequent reactivity effect) need to be understood and included in
PHWR safety analysis.

VENUS (Belgium)
ZED-2 (Canada)
KUCA (Japan)

4) Neutron flux andspectra:
PWR, BWR, VVER,
ALWR, PHWR,
APHWR

The neutron flux and spectra on the reactor pressure vessel (RPV) internal VENUS (Belgium)
structures and the RPV wall are critical to determining material
ZED-2 (Canada)
embrittlement, component lifetime and the potential for RPV failure due LR-0 (Czech Republic)
to pressurised thermal shock. Such data are especially critical to plants
seeking extended lifetime or those being designed for long lifetimes. Thisissue also applies to the ageing of pressure tubes in PHWRs.

5) Shielding:
PWR, BWR, VVER,
ALWR, PHWR,
APHWR

The ability of materials inside the reactor vessel to shield key
components from irradiation-induced damage is key to understanding
their lifetime and ability to withstand transients. Also, protecting
operating personnel and predicting the environment in which
equipment must function depends on predicting shielding performance.

LR-0 (Czech Republic)

6) Moderator
coefficients:
PHWR, APHWR

The coolant and moderator in PHWRs are separate. Thus, the impact onreactivity of changes in the heavy water moderator temperature, densityand poison concentration must be included in the safety analysis.

VENUS (Belgium)
ZED-2 (Canada)
KUCA (Japan)* Use for temperature coefficient studies.
Note: PWR: pressurised water reactor; BWR: boiling water reactor; VVER: water-water energetic reactor; ALWR: advanced light water
reactors; PHWR: pressurised heavy water reactors; APHWR: advanced pressurised heavy water reactors.
Source: NEA (2009).

3.2. Severe accident and containment phenomena
3.2.1. Introduction
Severe accidents (SA) – often typically understood as (reactor) core melt accidents – are generally
considered to be events beyond the traditional design basis of currently operating nuclear power
plants. Today (after Fukushima), the terminology “severe accidents” is used in a more general
sense and includes fuel melt accidents in the SFP as well. SAs are strongly linked with various
phenomena inside the containment (thermal-hydraulics, combustible gas behaviour, fission
product behaviour, molten core concrete interaction, etc.), as the containment is the last barrier
against fission product release into the environment in a SA. Therefore, this section covers bothsevere accidents and containment-related phenomena and research issues.
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Accident management which deals with the prevention or mitigation of SAs – both in thereactor core as well as in the SFP under all plant operating states – also focuses on maintaining
containment integrity, as the largest contributor to reducing risk to the public from the operationof nuclear power plants. Many experimental series supported the development of accident
management strategies or specific systems/components. Examples include experiments tounderstand the quenching of an overheated or partially molten core, experiments to assess theperformance of passive autocatalytic recombiners for removal of combustible gases from thecontainment, experiments to qualify filtered containment venting systems, and assessments of
the influence of engineered safety features (spray air coolers) on accident progression. Althoughgenerally not considered during initial licensing of most operating nuclear power plants, SAs
have been assessed through specific plant reviews, generic analysis and the development ofsevere accident management programmes. Severe accident management programmes are, tosome extent, plant design-specific, and have been significantly improved and extended after
Fukushima, based on the latest stress test results and findings of SA research and analyses.
SAs involve an initiating transient, such as a loss-of-coolant accident or a loss of SFP heatremoval accident, accompanied by the postulated failures of multiple safety systems, thus
compromising the capability to shut down the reactor or maintain adequate cooling of the fuelin the core or the SFP. These failures can result in damage and melting of the fuel, leading to
the release of significant amounts of radioactivity from the reactor core into the containment
or from the SFP into the surrounding building. In a core melt accident, the containment may
also be postulated to fail or to be bypassed (e.g. through steam generator tube failure in a PWR)
under certain circumstances, resulting in a major radioactive release to the environment. A fuelmelt accident in the SFP may also result in a major radioactive release to the environment, as
the SFPs in most nuclear power plants are typically in the “less protected” reactor building.
For many years, important national and international programmes have been undertakenin the field of severe accidents and their results have been shared through international“networks.” Many research programmes on various severe accident phenomena have been
carried out since the severe accidents in Three Mile Island and Chernobyl and contributed to
improve the safety of existing reactors. Once the objectives were achieved, interest andfinancing towards the related activities dropped down. Several important facilities all over theworld have been closed down, the most prominent in the severe accident area being the PHEBUS
FP facility at the IRSN. It is expected that at closure of the cited activities the question will beanswered whether the issue/facility can be closed or what further actions need to be initiated.
As a result, new research activities may be launched. The availability of data from those earlier
research programmes for future use is an important topic, as is maintaining the expertise ofexperimentalists/organisation in leading those large experimental programmes.
Several severe accident-related activities are still continuing or are being extended,
especially after the Fukushima accident. The CSNI is still playing a major role in organising andadministering co-operative research programmes in the area of severe accidents.
Earlier experimental programmes included: 1

1.

60

•

PHEBUS FP was conducted in France by the IRSN in the framework of EC programmes.
These in-pile experiments to study key physical phenomena associated with a severe
accident in PWRs specifically encompassed fuel rod degradation, release to and transportof radioactive materials in the primary system and the containment, and their physicochemical behaviour. Results provide an experimental basis of vital importance tounderstanding fuel degradation and FP behaviour.

•RASPLAV/MASCA (conducted in Russia to assess the molten corium pool behaviour andthermal load on the RPV lower head).

•

SNL-LHF/OLHF (conducted in the United States to assess the mechanical behaviour of
the RPV lower head until failure under pressurised severe accident conditions).

NEA joint projects: www.oecd-nea.org/jointproj.
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3. TECHNICAL ISSUES AND ASSOCIATED FACILITIES•MCCI (conducted in the United States to assess ex-vessel molten core concrete interaction
and core debris coolability using the ANL test facility).

•RASPLAV/MASCA (conducted in Russia to measure the physical properties of molten corematerial within the lower head of the RPV on small-scale facilities).

•

SERENA (a programme assessing the state of knowledge related to fuel coolant interactions
in- and ex-vessel using the KROTOS [France] and TROI [Korea] facilities).

More recent and/or ongoing experimental programmes include:•

SESAR Thermal-hydraulics (SETH-1) made use of the PANDA (PSI Switzerland) and PKL(Areva, Germany). SETH-2 used the PANDA and MISTRA (CEA, France) facilities andconsisted of thermal-hydraulic experiments in support of accident management.

•

The HYMERES (Hydrogen Mitigation Experiments for Reactor Safety) Project made use ofboth PANDA and MISTRA and provided data for detailed analysis of the release anddistribution of hydrogen in containment during severe accidents, the effect of activationof containment components (e.g. spray, cooler, heat sources – simulating the thermal
effect of PAR), and hydrogen distribution issues related to the thermal stratification in
BWR suppression pools.

•

THAI (Thermal-hydraulics, Hydrogen, Aerosols and Iodine Project conducted at Becker
Technologies, Eschborn [Germany] using the THAI/THAI+ facility, contributed significantly
to hydrogen and fission product-related issues in a water cooled reactor containment inaccident conditions).

•

ISTP (International Source Term Program) by the IRSN in the framework of EC
programmes: conducted at the CEA [France] using the VERDON facility to complete theexisting database on FP release for high-burnup UO2 and mixed oxide fuel; ruthenium
release under oxidant conditions was also investigated, conducted at the IRSN using theCHIP facility to develop chemical models [thermodynamic and kinetic description] of theiodine, caesium, boron, hydrogen and oxygen chemical system to improve predictionsof gaseous iodine release fractions at the RCS break.

•

BIP (Behaviour of Iodine Project) conducted separate-effect tests at the CNL (Canada) andmodelling studies of iodine behaviour in a nuclear reactor containment building following
a severe accident.

•

The STEM (Source Term Evaluation and Mitigation) project conducted at the IRSN (France)
to improve the general evaluation of the fission product source term.

•

The SFP (Sandia Fuel Project) used a 17x17 full-length PWR bundle to do experiments on
ignition of Zirconium alloy in a prototypical PWR fuel assembly in an SFP during acomplete drain-down.

Several of these projects provided data which are used for the development and validationof CFD codes, too.
The CSNI also sponsored efforts through senior expert groups to assess different research
aspects and open topics. Examples are the Senior Expert group on Severe Accident Management,
which reported SAMI published 2000) (NEA, 2011b).
Within Europe, a Network of Excellence SARNET (Severe Accident Research Network) was
organised to co-ordinate European research on severe accidents in nuclear power plants. After
2013, it continued in Technical Area No. 2 of the NUGENIA association. Within SARNET, the
SARP activity was related to define SA-related research priorities for future research (KleinHessling et al., 2014). An update of the earlier findings was recently made (Manara et al., 2019).
The latest activities and projects under the NUGENIA framework are summarised in
van Dorsselaere et al. (2017). One of them is the ALISA (Access to Large Infrastructures for Severe
Accidents) project (Miassoedov et al., 2018), a four-year project that started in mid-2014 andwhich was led by the Karlsruhe Institute of Technology. It addressed the transnational accessto large research infrastructures for optimal use of the R&D resources in Europe and China in
the field of SA analysis for existing and future power plants. To optimise the use of the resources,

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the project provided access to experimental platforms in Europe to Chinese research institutes
and access to Chinese experimental platforms for European research institutes. Activities
focused on large-scale experiments under prototypical conditions for SA issues in LWRs, suchas coolability of a degraded core, corium coolability in the reactor pressure vessel, possible melt
dispersion to the reactor cavity, and hydrogen mixing and combustion in the containment.
After the Fukushima accident, the CSNI organised an SA-related analytical project, BSAF
Phase I (NEA, 2015a) and II (Benchmark Study of the Accident at the Fukushima Daiichi Nuclear
Power Station Project to improve severe accident codes and to analyse the accident progression
and current status of units 1 to 3 of the Fukushima Daiichi). In 2015, the US DOE sponsored several
efforts to gain consensus among US experts related to gaps in knowledge about severe accident
progression (Farmer et al., 2015) and information needs from forensics examinations of thedamaged reactors at Fukushima Daiichi (Rempe et al., 2015). Finally, the CSNI sponsored anotherso-called Senior Expert Group SAREF (NEA, 2016) (Senior Expert Group on Safety Research
Opportunities Post-Fukushima), which addressed activities related to safety research knowledge
gaps and data needs from Fukushima Daiichi decommissioning. The report was published in 2016
and contains useful information on past SA research and experimental facilities.
All of these programmes have contributed to increase the knowledge about severe accident
phenomena, to the resolution of open questions related to severe accidents, and have identified
potential for further improvement of SA management measures to successfully terminate or
mitigate the accident progression. They have also served to maintain some of the key facilities
from premature shutdown, but not all of them. However, important research issues remain,
especially with regard to the late severe accident phase prior to and after RPV failure, to severe
accidents in BWR plants in general and with regard to some aspects of fission product behaviour,which need to be studied to support the continued safe operation of nuclear power plants, toreduce the risk of severe accidents through severe accident management programmes, as wellas to support the licensing of new nuclear power plant designs.
System and integral codes that simulate main relevant phenomena of SAs have reachedmaturity over the past two decades. Owing to continuous development of computing capability,
the application of the CFD methodology can now be used for safety evaluation, especially in thearea of containment thermal-hydraulics/phenomena, less in the SA domain in general.
Thermal-hydraulic behaviour inside the containment is generally 3D, for which analyses
using system or integral codes is not always the best solution. However, further efforts are neededfor efficient computation and enhancements of capability analysis for two-phase flow, gas
distribution, hydrogen combustion and so on. Experimental data with sufficient instrumentation
are required for the validation of the CFD codes.

3.2.2. Scope
Based on the most important safety issues and a review of the available experimental data tocover them, the first Senior Expert Group on SFEAR published a list of prioritised severe accident
issues (NEA, 2011a). It is worthwhile to review them in order to measure the progress made in
the last decade. Information from the Senior Expert Group SAREF (NEA, 2016), SARNET/SARP andfollow-up activities (Manara et al., 2019; Klein-Hessling et al., 2014), and the US DOE (Farmer etal., 2015; Rempe et al., 2015) on SA-related research priorities for future SA research have beenused to update the severe accident research topics presented in the last SFEAR report to someextent. The issues in this section are listed in Table 3.5 and arranged still according to the main
phases of progression of a severe accident in the reactor core and the phenomena present ineach of those phases; SAs in SFP are added as a separate topic in addition to the same issue
mentioned under the “thermal-hydraulics area” in Section 3.1.1:

62•in-vessel phenomena (e.g. core heat-up, reflooding and quenching, clad/fuel melting andrelocation, combustible gas generation);•ex-vessel phenomena (e.g. reactor vessel failure, core concrete interaction, debris cooling,
combustible gas phenomena);

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3. TECHNICAL ISSUES AND ASSOCIATED FACILITIES•

source term/fission product behaviour (e.g. quantity, chemical form, transport andtiming of fission product release from the fuel [in- and ex-vessel], behaviour in RCS andcontainment);•

containment integrity (e.g. capability of containment to withstand severe accident
conditions caused by combustible gas burning, decay heat, molten core concrete attack,as well as containment bypass scenarios);•

phenomena of SAs in an SFP (e.g. loss of fuel cooling, fuel heat-up and melting).

3.2.3. Description
The severe accident issues and phenomena that could benefit from additional, ongoing or planned
research are related to reducing the remaining uncertainties in SA progression, especially in thelate phase prior to and after RPV failure and with regard to BWR, to the improvement of mitigation
measures, and to the understanding of the safety implications caused by changes in plant design
or operating characteristics (e.g. high-burnup or MOX fuel). The prevention and mitigation ofsevere accidents remain an important objective for the continued safe operation of nuclear power
plants; the resolution of severe accident issues is tightly coupled to it.
To reduce uncertainties and close knowledge gaps, current research should be conducted at
sufficient scale to investigate the important phenomena and use real materials, wheneverpossible. The Senior Expert Group SAREF (NEA, 2016), SARNET/SARP and follow-up activities(Manara et al., 2019; Klein-Hessling et al., 2014), and US DOE (Farmer et al., 2015; Rempe et al., 2015)
on SA-related research priorities concluded that although the early phase of in-vessel core melt
progression is fairly well understood, significant uncertainties remain in the later phase, including:•melt relocation within the RPV, especially for BWR;•

late-phase core degradation, melt behaviour in the lower plenum and hydrogen generation,
closely linked to accident management issues of in-vessel melt retention;•

RPV failure mode, especially for BWRs with many/large penetrations, and the questionof whether or not molten core material will remain in-vessel after RPV failure;•

consequences of molten core material release from reactor vessel and retention/hold up
by external structures below the RPV, especially for BWR;•

melt spreading even under water and melt coolability respectively termination of MCCI,and to some extent combustible gas generation;•

FP release, transport and behaviour in the containment, especially FP retention by poolscrubbing, effects of water chemistry on source term generation, linked to conditions
which could lead to containment failure or bypass.

Improvement and determination of the best accident management strategies for preserving
RPV and containment integrity and reducing the amount of radioactive material available forrelease to the atmosphere will continue and benefit from ongoing research.
Severe accident progression in BWRs is generally less well understood due to a lack ofexperimental data as most of the experiments have been conducted for PWRs. PHWRs have
similar SA issues as LWRs, however. The core melt progression in a pressure tube reactor
presents additional challenges associated with propagation of pressure tube failure, fuel coolant
or fuel-moderator interaction, the potential to overpressurise the calandria and cause expulsionof the moderator through calandria overpressure relief ducts and additional pressure tube
rupture. Finally, severe accidents in SFP with air/steam access to the overheating/melting fuelis another open issue.

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Table 3.5. Severe accident and containment-related issues – past and current
IssuesDescription

Facilities

Notes

In-vessel phenomena
1) Pre-core melt
conditions

Understanding the conditions that can lead to coremelt and the thermal-hydraulic conditions of thecore prior to core melt (see Section 3.1.1) is essential
to understand whether or not the implementation of
accident management strategies will be successful incore melt prevention. Knowledge of pre-core melt
thermal-hydraulic conditions will also help to refine
accident management strategies so as to understandand be prepared for the outcome of actions taken by
the operator. For CANDU-type reactors, theevaluation at calandria vessel subshell, penetrations
and end-shield cooling system was an issue.

PHWR
CANDU: Fuel
ChannelSafety Lab(Canada)Several intermediate-scale studies
evaluating critical heat flux in calandria
vessel in PHWRs (CNL, Canada) havebeen done. Experiments can also bedone to assess the integrity of fuelchannels of PHWRs.

2) Coolability ofoverheated core/debris bed andquenching

The effects of adding water to an overheated or
partly molten core as well as to a debris bed isinvestigated in order to understand the thermalhydraulics phenomena, hydrogen generation andcooling (quenching), and to access whether accident
management measures to prevent core melting or tostop melt progression are efficient. The issue of
quenching of an overheated fuel bundle is closely
coupled with Issue 15 of Section 3.1.1.

QUENCH, FZK(Germany)

Reflood/quenching of an overheated fuelbundle. Plans for operation until 2022.

PEARL(France)

Dedicated to debris bed reflooding. Its
main features are: water injection flow rate
up to 50 m3/h, pressure up to 10 bar; initialdebris bed temperature up to 900°C;
debris bed of 500 kg, heated by induction.
Currently used for the French PROGRES
programme (up to 2021), and in parallel
for the European IVMR programme. Use of
PEARL facility in NEA projects may beenvisaged in the near future.
CORDEB can obtain material properties
using real materials Used in the H2020
IVMR frame of the European project(2015-19).

CORDEB(Russia)3) In-vessel melt
progression

In-vessel melt progression includes melt relocation in
the core and to the lower plenum of the reactor
pressure vessel (RPV) and determines the heat loadon the RPV wall during a core melt accident. The typeof fuel (UO2 or MOX), cladding material, burnup andother factors which affect the composition of themelt are also important in this determination. The
amount, composition, rate and timing of a core melt
are important to determining the effectiveness ofaccident management measures and the ability of
the RPV or reactor calandria to maintain its integrity.

4) In-vessel fuelMolten fuel relocating within the reactor andcoolant interaction contacting reactor coolant or moderator (PHWR,(FCI)
APHWR) may cause the rapid generation of steam,and this is an important component of the load on
the RPV or calandria.

5) Effect of air oncore melt
progression

64Core melt accidents where air is present in the RPV(such as during refuelling) could behave differently
than those where no air is present. This could include
the dynamics of the melt progression and the FP
release.

CNL melt
facilities(Canada)Cold-crucible experiments for
prototypical CANDU corium melts andsimulant melt experiments to evaluate
convection within the melt.

LIVE-FZK(Germany)

Uses simulant material; studies the latephase of core degradation, onset of
melting, and the formation and stabilityof melt pools in the RPV lower plenum.
Facility still exists, but currently not used.

PEARL(France)

Description provided under Issue 2.

KROTOS (invessel FCI)
(France)
TROI (exvessel FCI)
(Korea)FCI tests using prototypic materials.
Operation within OECD SERENA projects
completed. Facilities still exist.

MFMI(Canada)CANDU prototypical experiments withmolten corium was constructed to study
such extremely violent boilingphenomena and vapour generation
during an energetic interaction between
molten fuel and water. Facilities still exist.
Can conduct hot cell experiments with
irradiated fuel.

VERDON(LECA-STAR)(France)
CNL Hot Cells(Canada)Can conduct hot cell experiments with
irradiated fuel under various
experimental conditions.

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Table 3.5. Severe accident and containment-related issues – past and current (cont’d)IssuesDescription

Facilities

Notes

6) Effect of highburnup and MOX
fuel

The use of high-burnup or MOX fuel could change
the dynamics of melt progression and fission productrelease. Data on these effects are needed to properly
assess consequences and risk from accident
sequences involving high-burnup or MOX fuel.

VERDON(France)

Can conduct hot cell experiments with
irradiated fuel.

CNL Hot Cells(Canada)Can conduct hot cell experiments with
irradiated fuel under various
experimental conditions.

7) RPV pressure

None
Depressurising the primary coolant system isimportant during the in-vessel melt progression phaseto reduce stress or facilitate water injection into the
RPV. Accordingly, if the design does not have thecapability to depressurise the primary system, it isimportant to understand the effect of high pressure on
the RPV and other RCS components’ integrity and thesubsequent effect on core melt progression. This is
primarily an analysis issue.

8) Maintaining RPV
integrity by invessel melt
retention

Maintaining the integrity of the RPV or reactor
calandria vessel is important to terminating andconfining a core melt accident, thus eliminating
ex-vessel severe accident phenomena and theirchallenge to containment integrity. Cooling the RPVor reactor calandria both internally and/or externallyare potential strategies for maintaining RPV integrityin the event of a core melt accident. However, higher
core power densities will make it more difficult tomaintain RPV integrity by external cooling due to thehigher heat flux on the RPV. Knowledge of RPV
integrity as a function of heat flux is important inassessing the success of accident management
strategies.

LIVE-FZK(Germany)

Description provided under Issue 3.

Calandria
VesselIntegrity
Experiments(Canada)Measurement of critical heat flux (CHF)
on a scaled calandria vessel with externalcooling and internal heating of the shell.
Measurement of corium convection in acalandria vessel corium convection uses
simulants with crust thickness
measurements.

IVR2D (China)

The 2D IVR facility, REVECT-II at the CNPRI,
is designed to investigate the external
reactor vessel cooling of the lower plenumof the RPV in order to achieve in-vessel
melt retention. Experiments focus on theinfluence of heat flux profiles either in a
homogenous oxide pool or in a two-layer
stratified pool in the RPV lower plenum onnatural convection and CHF.

CORDEB(Russia)

Can obtain material properties using realmaterials. Currently used in the H2020
IVMR project.

IVR3D (China)

IVR3D test facility at the CNPRI studies 3D
vessel external cooling under integrated
reactor component mode. The aim is tocompare CHF under different coolingchannel geometries and two-phase
coolant flow patterns. The research
programme includes: investigation of
two-phase coolant circulation process ofnatural convection, venting andcondensation; and determination of the
CHF in 3D geometry.

COPRA (China) This facility at XJTU is designed to study
the natural convection heat transfer incorium pools with high Rayleighnumbers up to 10^16. The test vessel is a
two-dimensional ¼ circular slice with an
inner radius of 2.2 m, it simulates fullscale for the CNNC’s ACP-1000.

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Table 3.5. Severe accident and containment-related issues – past and current (cont’d)IssuesDescription

Facilities

Notes

9) Pressure tube
integrity

Maintaining the integrity of the pressure tubes in apressure tube reactor (PHWR) is important for
maintaining cooling of the fuel in the tube andpreventing overpressurisation and failure of thecalandria due to high-pressure water injection
and/or molten fuel injection and FCI.

High
Temperature
Fuel Channel
Lab (Canada)

Fuel channel thermal-mechanicalbehaviour (pressure tube and calandriatube) at high-temperature conditions.

B) Ex-vessel phenomena
10) Ex-vessel melt
progression anddebris coolability11) Core concreteinteraction

The amount, rate, timing and spreading of molten
core material released following RPV failure areimportant to determine the ability of the concretebase-mat to maintain its integrity and the ability ofan overlying pool of water or base-mat coolingsystem to cool the debris and terminate the coreconcrete reaction (i.e. ex-vessel melt coolability).
Debris coolability can be affected by the amount ofwater overlying the core debris and the porosity ofthe debris or the strength of the crust formed on top
of molten core debris. Obtaining the properties of
the crust and underlying debris is important tounderstanding debris coolability and its
uncertainties.

VULCANODry melt
spreadingCan use prototypic materials.

PULiMS (KTH
Sweden)Melt spreading under water.

Support of the development of new mitigation
PLINIUS-2
concepts (Gen 2 or Gen 3 reactors) with conditions as (France)
close as possible to the reactor case. First test isplanned in 2021.

The CEA built a new large mass
prototypic corium experimental
platform, PLINIUS-2, to support
consolidation of modelling andvalidation of corium behaviour with
prototypical corium.

When molten core material leaves the RPV, it cancome into contact with concrete of the reactor
cavity. Depending on the amount and depth of thematerial and the composition of the concrete,various amounts of combustible and noncombustible gas will be released into thecontainment, raising its pressure. These gases canalso be a source of additional energy if they ignite,
causing additional pressure and temperature rises. If
not stopped, the core concrete interaction can also
penetrate the reactor containment base-mat, thus
failing containment. Understanding the rate andamount of gas generated from core concreteinteractions is important to understanding thepotential for containment failure and for success of
mitigation strategies and, in the case of new plantdesigns, aid in selecting materials and configurations
to minimise core concrete interactions.

MCCI(UnitedStates)

Large-scale test (1 m2) with realmaterials, simulated decay heat and withor without overlaying water cooling.
Facility used with OECD project.

MOCKA(Germany)

Core concrete interaction using simulant
corium and decay heat simulation,modelling of concrete with reinforcing
bars.

ARTEMIS(France)

Uses simulant material.

KROTOS (invessel FCI)
TROI (exvessel FCI)

FCI tests using prototypic materials.
Operation within OECD SERENA projects
completed. Facilities still exist.

12) Ex-vessel fuelUpon failure of the reactor pressure vessel, molten
coolant interaction core material may fall or be ejected into water if thereactor cavity has been partially or fully flooded.
Such contact with water has the potential to cause
rapid steam generation and, depending on theamount, rate, fragmentation and mixing of themolten material, release a large amount of energythat should be taken into account when assessing
the structural integrity of containment.

66

MCCI (United Large-scale test (1 m²) with real materials,
States)simulated decay heat and with orPrimarily MCCI without overlaying water cooling.

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Table 3.5. Severe accident and containment-related issues – past and current (cont’d)IssuesDescription

Facilities

NotesC) Source term/fission product behaviour13) Fission product The amount, composition, chemistry and timing offission product release from the fuel through thechemistry andreactor coolant system and into containment results
releasein the source term available for release to theenvironment. This affects the onsite and offsite
consequence analysis and protective actions, which
need to be planned for. The source term is also
affected by the type of fuel (UO2 or MOX) and burnup
level. In addition, the effectiveness of source term
attenuation measures (e.g. sprays, water chemistry,
filters, pool scrubbing) needs to be understood.

14) Postcontainment
failure FP releaseto theenvironment

VERDON
Fission product behaviour in(France)containment.
THAI (Germany) Aerosol and iodine behaviour incontainment.
ARTIST(Switzerland)

The ARTIST Facility at PSI was a uniquesteam generator configuration: bundle,
separator and dryer, used to study steam
generator tube rupture accidents and FP
retention. Facility is no longer in use.
Chemistry of iodine under irradiation

EPICUR(CHROMIA
Platform,
France)STARTTransport of ruthenium in thermo(CHROMIA
hydraulic and chemical conditions
Platform,
representative of those of the primaryFrance)system.
CHIP (CHROMIA Transport of iodine in thermo-hydraulicPlatform,and chemical conditions representative
France)of those of the primary system.
CNL Hot cell
Fission product release from spent fuel,Facility
including leaching characteristics.
(Canada)Containment failure can lead to additional FP release
due to revolatilisation under depressurised conditions
and/or due to air ingress.
Understanding these phenomena is necessary forpredicting the consequences and risk from accident
sequences that fail and depressurise the containment.

D) Containment phenomena
15) Containment
integrity/venting

Understanding the conditions (e.g. pressure,temperature and equipment failure) which could leadto containment failure is important. Therefore,knowledge of the integrated effects of design basis
and severe accident loads and the use of safetysystems is necessary input to containment design. Thisissue provides input for structural analysis andcontainment failure modes testing (see Section 3.1.5).
Preventing containment overpressure failure bycontainment venting is one of the AM measures
implemented during a severe accident. Sincecontainment might have a significant amount ofhydrogen, venting might cause large-scale hydrogen
combustion with an unacceptable risk.
To reduce such a risk, the AM measure should bebased on understanding of thermal-hydraulicbehaviours during venting, and effects of geometryof the system and operational conditions. Several
containment research facilities exist with different
purposes.
Containment venting specific issues are notsufficiently established due to lack of related studies.
For example, prediction of hydrogen concentration
at the intake of the venting system is difficult whenhydrogen is non-uniformly distributed in thecontainment. This is primarily an analysis issue.

PANDA(Switzerland)

Large-scale tests in multi-compartment
configurations for decay heat removalfrom containment; originally designed to
study BWR behaviour.

THAI(Germany)

Two-vessel facility built to study thermalhydraulics, hydrogen, aerosol and iodine
related containment issues.

MISTRA(France)

100 m3 PWR containment (0.1 linear
scale) using helium as a simulant of H2.
Flexible free and compartmented
volumes. Capability for steam and gas
injection. Spray system, 3D
instrumentation.
7 m3 facility to simulate H2 (using helium)
mixing under severe accident conditions,
including steam, sprays and aerosols.
1 700 m3 containment thermalhydraulics facility, use of helium as asimulant.

TOSQAN(France)

LSCF (LargeScale
Containment
Facility,
Canada)CIGMA (Japan) Facility for investigations of LWR
containment thermal-hydraulics, built in2015 for experiments addressing
containment responses, separate effects
and accident management.

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3. TECHNICAL ISSUES AND ASSOCIATED FACILITIES

Table 3.5. Severe accident and containment-related issues – past and current (cont’d)IssuesDescription

Facilities

Notes16) Containment
bypass:Overheating andfailing steam
generator tubes

During core heat-up under severe accident conditions,significant amounts of heat are transferred by natural
circulation of steam and hydrogen to the hot legs,
surge line and steam generators. This may cause one
or more steam generator tubes (SGTR) to fail prior to
failure of the hot leg, surge line or vessel lower head, orlead to a failure in a connected system outside thecontainment (ISLOCA). During these conditions, fission
products could be released by the containment bypass
scenario. Understanding the behaviour of steam
generator tubes is important for preventing
containment bypass scenarios. Many phenomena areto be studied such as aerosol retention inside/outside
tube in steam generator under the wet/dry conditions,
iodine retention in pool, etc.

AEOLUS(KAERI)A scaled-down model of the SG in
OPR1000 and APR1400 that can perform
experiments of aerosol retentions in the
break vicinity.

ARTIST(Switzerland)

The ARTIST Facility at PSI was a unique SG
configuration: bundle, separator and dryer,
used to study SGTR accidents and FP
retention. Facility is no longer in use.

17) Combustible
gas control

Combustible gas (H2 and CO) are generated fromcore oxidation during the in- or ex-vessel phase of asevere accident, from metal-water reactions orcore-concrete interaction after RPV failure. Hydrogen
combustion can further heat the containment
atmosphere and/or pressurise the containment, thus
challenging containment integrity. Hydrogencountermeasures, especially the use of passive
autocatalytic recombiners, is a worldwide used SAM
concept. Experiments are needed to show theefficiency of PARS under various conditions.

PANDA(Switzerland)

Large-scale-multi compartment
capability for 3D mixing and studying theeffect of safety systems.

THAI(Germany)

Two-vessel facility build to study
thermal-hydraulics, hydrogen, aerosol
and iodine related containment issues.
Many PAR-related experiments havebeen performed.
Instrumentation can measure 3D
thermal-hydraulics data.

MISTRA(France)

Large-scale tests in multi-compartment
configuration for mixing and distributionof H2 and studying the effect of safetysystems.

TOSQAN(France)

Small scale, uses helium as a simulant of
H2.
HYMIT (China) This facility at SJTU is a medium-scalehydrogen mitigation test facility
designated for investigations ofhydrogen recombination andcombustion. It can operate withhydrogen concentrations between 0 and
30 vol.%, and steam concentrations
between 0 and 50 vol %. The main partof the facility is a steel cylinder tank
4 metres high and 2 metres in diameter.
MCTHBT
The MCTHBF is a hydrogen mitigation(China)test facility at the NPIC with a steel
cylinder tank 5 metres high and2.8 metres in diameter. It is designated totest the hydrogen recombination andcombustion behaviour. The topics ofstudies includes scaling from small tolarge volumes by comparing results ofother previous experiments andvalidation of lumped-parameter and CFD
codes.
CIGMA (Japan) A facility with 60 m3 containment forthermal-hydraulic experiments related tohydrogen risk and over-temperature
containment damage under a broad
range of thermal-hydraulic conditions(pressure up to 1.5 MPa, temperature up
to 973 K).

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3. TECHNICAL ISSUES AND ASSOCIATED FACILITIES

Table 3.5. Severe accident and containment-related issues – past and current (cont’d)IssuesDescription18) Containment
cooling

Several cooling measures are considered in different
containment designs as well as in severe accident
management strategies to prevent over-temperature
containment damage. Those may rely on water spray
injections, fan coolers, stagnant water pool covering
the containment top outer surface, free convection
air flow along the outer surface and so on.
Effectiveness of a cooling measure is dependent on
integral effects by the cooling on phenomena
including gas specimen distributions, condensation,flows discharged from the pressure vessel, heattransfer deterioration due to the presence of noncondensable gas, etc.
Interactions with hydrogen risk issues should also be
considered because condensation might increasehydrogen concentration. Therefore, knowledge ofthe integrated effects caused by cooling is relevant
to validate the accident management measures toprevent over-temperature damage.

Facilities

Notes
In principle, the same facilities as listed
under Issue 15.

E) Spent fuel pool
19) Core melting

Understanding the conditions which could lead tosevere accidents in the SFP is important. Loss-ofcoolant accidents are typically prevented by designof the SFP. Loss of cooling and evaporation of thewater in the SFP can cause heat-up of the fuelassemblies dependent on their decay power level.
Specific phenomena to be studied are oxidation,
melting and fission product releases under air
containing atmosphere. As the time available toprevent core melting under loss of coolingconditions are typically long, accident management
concepts prefer the use of preventive measures toreinject water from outside. Mitigative measures areless developed.

SFP (Sandia
Fuel Project)

Used a 17x17 full-length PWR bundle to
do experiments on ignition of Zirconium
alloy in a prototypical PWR fuel assemblyin an SFP during a complete drain-down;
experiments and programme completed.

Note: PHWR: pressurised heavy water reactor; APHWR: advanced pressurised heavy water reactor; BWR: boiling water reactor; LWR: light water
reactor; PWR: pressurised water reactor.

CFD was previously listed as an issue for the severe accident technical area (NEA, 2011a),although the real issue is that facilities in which severe accident research is conducted musthave the appropriate instrumentation to evaluate 3D distribution of temperature, hydrogen andsteam, and the possibility to measure condensation and gas flows. Thermal-hydraulicbehaviour in containment is generally 3D, for which analyses are difficult using a conventional
system code such as TRACE and MELCOR. With recent developments in computing capability,
the application of the CFD methodology to safety evaluation is possible. However, further efforts
are needed for efficient computation and enhancements of analysis capabilities for two-phaseflow, hydrogen combustion and so on. For the validation of the CFD code, experimental datawith sufficient instrumentation are required. PANDA, MISTRA, THAI and the LSCF can be used
to obtain CFD grade data.

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3. TECHNICAL ISSUES AND ASSOCIATED FACILITIES

3.3. Integrity of equipment and structures
3.3.1. Introduction
Many of the current problems with operating reactors are related to material degradation issues.
While plants were initially designed for a 40-or-more-year lifetime, a number of unanticipatedmaterial problems have occurred. As plants continue to operate and seek to extend their lifetimes,
and in some cases raise their power levels, issues related to component and structural integrity
still need to be investigated and solved. Accordingly, ensuring the condition of equipment andstructures is monitored and known becomes increasingly important. Ageing mechanisms include
stress corrosion cracking, corrosion, erosion, wear, fatigue, thermal and irradiation embrittlement,
creep/stress relaxation, and irradiation-induced void swelling and deformation. Thesemechanisms can affect metallic and concrete equipment and structures as well as electrical andinstrumentation and control (I&C) cables. Identifying, monitoring and controlling the ageing
mechanisms are important for the continued safe plant operation.

3.3.2. Scope
The issues addressed in this section are related to identifying the phenomena that canpotentially cause problems, improving techniques for detecting and repairing problems, andusing effective ageing management programmes to anticipate and prevent problems before
they become safety issues. The safety issues that could benefit from additional research arethose that are associated with the ageing of existing plants and those that are associated withinitiatives to improve plant performance or develop new plant designs, as listed in Table 3.6.
The plant ageing mechanisms of wear, creep/stress relaxation, and void swelling have been
added to Table 3.6 since the original SFEAR report to provide a more complete list of ageing
mechanisms that can affect commercial nuclear power plants.

3.3.3. Description
The timely detection and mitigation of ageing degradation in safety-significant plant systems,
structures and components (SSC) are important to ensure SSC integrity and functional operability
throughout plant service life. Maintaining SSCs is achieved through effective ageing management
programmes. General guidance for such programmes related to nuclear power plant operation,
inspection, testing, examination, maintenance and surveillance is provided by the International
Atomic Energy Agency (IAEA) Nuclear Safety Standard (NUSS) Code. Specifically, the IAEA has
developed the International Generic Aging Lessons Learned (iGALL) programme to support ageing
management of safety-significant SSCs and support long-term nuclear power plant operation. The
objective of iGALL is to provide a technical basis and practical guidance on managing the ageingof mechanical, electrical and I&C components, and civil structures of nuclear power plants
important to safety. The programme is relevant for plants in operation, for plants considering
long-term operation, as well as for new plants with both conventional and new designs.
Research programmes will continue to be needed to augment and improve this guidance byassessing how the reactor environment and operating conditions degrade the strength andintegrity of equipment and structures over their operational lifetime. More specifically, in thearea of integrity of equipment and structures, research programmes should address:

70•

The initial and potentially degraded state of materials that comprise equipment andstructures. Degradation is affected by material composition, manufacturing processes
and operational parameters.

•

The loads imposed on the equipment and structures during operation (i.e. normal andtransient operation, incidents, accidents and events), which are combined with the initialfabrication stresses.

•

The historical and planned future environmental conditions associated with theequipment or structure.

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3. TECHNICAL ISSUES AND ASSOCIATED FACILITIES•The overall susceptibility of SSCs to ageing degradation, which is determined by the
combined material, loading and environment attributes.

•

The presence of defects which result from manufacturing practice and environmental
attack.

•The safety margins available in the design.

•

The sensitivity and effectiveness of pre-service and in-service examination and testing
methods.

The safety significance and state of knowledge related to either the plant ageing mechanisms
or the application of new materials in existing reactors or advanced designs are ranked in Table 3.6.
These rankings were provided in the original SFEAR report and the current rankings are largely
consistent with the previous ones. Noted differences are the increases in safety significance on
crack initiation and propagation (medium to high), in-service inspection (medium to high) andflow-induced vibration (low to medium). The significance of cracking has been elevated to reflect
the greater propensity for cracking as the plant ages as evidenced by the continued discovery ofcracking since the original SFEAR report. The importance of in-service inspection has been
elevated to recognise its importance in finding flaws and other types of degradation before thedegradation affects component functionality. The importance of flow-induced vibration waselevated from low to medium based on the fretting concerns with steam generator tubes and also
the emergence of this issue in EPR surge lines. Further, the state of knowledge of erosion/corrosion,
thermal and irradiation embrittlement, and materials for high-temperature advanced reactors
has increased since the original SFEAR report to reflect work in these areas in the 2000s and 2010s.
Examples of facilities needed to address these issues are also provided in Table 3.6. The
research needed to address irradiation-related issues typically requires test reactors to irradiatematerials under conditions representative of those found in commercial nuclear power plants.
Then the materials are transferred to hot cells with autoclaves and environmental loops or, insome cases, to other testing facilities. Other broad-based (e.g. nuclear and non-nuclear) issues
may require large capacity systems to evaluate components at near-full scale, environmental
chambers to accurately simulate plant conditions or specialised equipment to characterise thematerial properties and damage evolution. While all of these facilities are important, the focusof this report is on the test reactors and other unique large-scale experimental facilities neededto address issues specific to commercial nuclear power plants. Therefore, only those uniquefacilities are indicated in Table 3.6.
Additionally, due to the large number of hot cells and autoclaves and the fact that mostmember countries have those facilities, specific recommendations on their preservation is
outside the scope of this evaluation. However, it is recommended that each member country
continue to monitor the status of these facilities and identify any concerns regarding loss ofcritical infrastructure to the CSNI and its member countries. It should also be noted that theclosure risk of major facilities needed to address the issues in Table 3.6 that are not highly safetysignificant is not of sufficient concern to be assessed in this report.
Since the last version of this report was published, several events have changed theresearch activities in this area and the facilities that are available to address them. First, theclosure of two powerful research reactors, Halden and the NRU, have necessitated OECD actions
and national task forces to fill the neutron gap, the expertise in designing in-reactor evaluationsof fuel and materials, and facilities to examine irradiated fuels and materials. Second, theFukushima accident, and numerous national initiatives to safely continue the operation ofcommercial nuclear reactors beyond their original design lifetimes, have led to increased
activities in evaluating fitness for service, and research on the effects of ageing on structural
materials and components. Third, numerous new reactor types and fuel types (including ATF)
are under development, and which require new materials to realise performance requirements.

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Table 3.6. Current integrity of equipment and structures issues
A) Plant ageing issues

Facilities

Notes

1) Erosion/corrosion

As plants age, environmental conditions can causesome materials to corrode or erode more than was
originally anticipated in the design. The corrosion canbe either internal to the component or external due toleakage elsewhere within the plant. Corrosion can leadto cracks, leaks or even large failures. Mitigation of theconditions that lead to excessive degradation needs tobe implemented to support continued operation
and/or plant life extension.

N/A

Most erosion/corrosion issues arefor balance of plant systems anddo not require unique, large-scale
nuclear-specific facilities.
Corrosion issues within the
primary system (e.g. feeder tubes)
require nuclear-specific facilities.

2) Thermal andirradiation
embrittlement

Embrittlement of steels, particularly in the reactor
pressure vessel, reactor internals and other primarypressure boundary components, due to either
prolonged exposure to relatively high operating
temperatures and/or to fast neutrons, can reduce theirability to withstand thermal and mechanical stresses.
The properties of embrittled materials and the ability to
predict the amount of their embrittlement need
continued confirmation by experimental data. Also, theeffectiveness of corrective actions needs to bevalidated.

ATR
HFIR
BR-2
LVR-15
HFR
BOR-60

Facilities listed are among themore prevalent test reactors withcapabilities for irradiating
materials. More information onthese and other similar reactors
can be found in Table 3.3.
No unique large-scale facilities
needed to address thermal
embrittlement issues.

3) Crack initiationand propagation

Crack initiation and propagation (i.e. cracking) in steel,
PHWR pressure tubes, and concrete equipment andstructures has caused problems at operating reactors.
Cracking is caused by a combination of stresses,
environmental conditions (including irradiation) andmaterial properties and may result from poor design,poor material selection, or an inaccurate knowledge ofactual stresses and environmental conditions. Cracking
may occur under constant or alternating loads(i.e. fatigue). The causes and corrective action forcracking need additional experimental confirmation.

ATR
HFIR
BR-2
LVR-15
HFR
BOR-60PNNL

Material test reactors identified
specifically here and moregenerally in Table 3.3 are neededto address irradiation-related
issues.
The PNNL has developed
capabilities for simultaneous
corrosion cracking initiation andpropagation testing.

4) In-service
inspection

Inspection techniques to look for cracks, erosion or theeffects of other ageing degradation mechanisms isimportant to detect and correct ageing effects before
they lead to a safety concern. Development andvalidation of inspection techniques is essential.

EPRI-N.C.
(United States)
Pressure Tube
Surveillance
Facilities(Canada)Performance demonstration
initiative facility and extensive
mock-ups for technical andpersonnel qualification.
Analysis of hydrogen ingress into
pressure tubes by analysis of“scrapes” or removed pressure
tubes. Hot cells withaccompanying analyticallaboratories required.

5) Cable insulation
degradation

Cable insulation (both power and I&C cables) can crack
and become brittle over time. Environmental
conditions such as irradiation, high temperature andmoisture affect the rate at which this happens. Thisdegradation can lead to shorts, fires or unexpectedbehaviour such that the cable cannot perform its
intended function. Detection and mitigation
techniques need to be verified.
As plants age, concrete properties change and/or
cracks develop. Ageing can be exacerbated bymoisture, irradiation, high temperatures and otherenvironmental conditions. The safety implications ofconcrete ageing need to be understood to support thecontinued safe long-term operation of existing plants.

N/A

Numerous CSNI countries havethese types of facilities and somehave been referenced in NEA(2018a).

ODOBA (IRSN,France)NIST(United States)
IETcc and Enresa(Spain)University ofToronto(Canada)1 700 m² platform, 60 large-scale
experimental blocks.
Large-scale facilities for testing
full thickness components.
Examination of concrete fromJose Cabrera, a 160 MWe PWR
decommissioned in Spain.
Facilities used in the OECD
programme ASCET.

6) Long-term
behaviour ofconcrete structures

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3. TECHNICAL ISSUES AND ASSOCIATED FACILITIES

Table 3.6. Current integrity of equipment and structures issues (cont’d)A) Plant ageing issues

Facilities

Notes

VERCORS(France)
Various facilities
studying related
phenomena

A one-third scale mock-up of areactor containment building in
Paris. More than 500 sensors and
2 km of fiber optic cables are
positioned in concrete slabs, bothon rebar and pre-stressing cables.
From the first concreting to theend of the programme,
temperature, strain, water content
of the concrete will be tested
daily. Concrete samples will also
be prepared and material
behaviours (resistance andmodulii, drying, shrinkage, creep
and permeability) tested.
Facilities addressed in
Section 3.1.4 and in conjunction
with Item 6, above.

7) Containment
integrity

The conditions under which containments fail, and thetiming and modes of failure, are important tounderstand in order to assess safety margins,consequences (i.e. fission product release) and risk.
Therefore, the structural analysis methods require
experimental confirmation due to the complex natureof containment designs and penetrations. Much of theexperimental validation data are dated and may not be
sufficient to validate analytical codes attempting toaddress current safety issues.

8) Flow-induced
vibrations*

As current plants continue to pursue power increases,
MSUB facility
flow distributions, particularly in-vessel, and their(Canada)contribution to mechanical loads and vibration ofequipment needs to be understood, monitored andevaluated to ensure continued functionality. Predicting
such flow distributions and assessing vibratory loadingfrom both mechanical and thermal-hydraulic sources
needs experimental data to validate analytical tools.

For steam generator U-bend andtube bank evaluation.
Replacement costs are high.

9) Void swelling anddeformation

Void swelling is the gradual increase in an equipment’s
or structure’s volume caused by nucleation and growthof clusters of vacancies produced by irradiation. Voidswelling may cause dimensional changes that exceed
the tolerances of a component and may also induce
stresses due to differential swelling. Void swelling
susceptibility increases at higher temperatures andlevels of irradiation and requires understanding andmitigation for some components that are expected tofunction beyond 40 years.

ATR
HFIR
BR-2
LVR-15
HFR
BOR-60Material test reactors identified
specifically here and moregenerally in Table 3.3 are neededto address irradiation-related
issues.

10) Creep/stressrelaxation

Creep (or alternatively stress relaxation) can occur inequipment or structures exposed to elevatedtemperatures and irradiation. Components such as
bolts or pre-compression structures that need tomaintain a preload for functionality are potentiallysusceptible to this ageing mechanism. Creep and stressrelaxation may also lead to cracking due to othermechanisms discussed in this table. The complex
nature and interrelationship to other degradation
mechanism makes the understanding of this ageing
mechanism important for certain components. PHWR
designs with horizontal pressure tubes are subject toboth axial and diametral creep, which has an impact oncore flow.

ATR
HFIR
BR-2
LVR-15
HFR
BOR-60Pressure tube
creep research in
Canada, see
Section 3.1.1.

Material test reactors identified
specifically here and moregenerally in Table 3-3 are neededto address irradiation effects.
Effects of creep on core flow anddistribution in full-scale CANDU
bundles.
No unique, large-scale facilities
are needed to address thermal
creep effects.

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Table 3.6. Current integrity of equipment and structures issues (cont’d)B) Performance improvement/new design issues

Facilities

Notes11) New materials –existing plants andfuture conventional
plantsBoth current and future plants will consider using newmaterials that are intended to be less susceptible torelevant ageing degradation mechanisms to mitigate
materials problems (e.g. cracking, corrosion) in both
replacement and new components. There will be little
service experience associated with the material in theintended application. Therefore, both short- and longterm application-specific performance of thesematerials needs to be understood and evaluated, andthe materials will need to be appropriately qualified forthe intended applications.

Facilities
previously
identified arealso used toevaluate specificdegradation
issues.

The evaluation and qualificationof new materials is also an
important issue for nuclear fuels.
See Section 3.1.2.

12) New materials –ALWR and non-lightwater reactor(ANLWR) designs:
e.g. APHWR, MSR,
HTGR, LSR

A variety of advanced reactor designs that are being
developed and considered throughout the world plan
to operate at significantly higher temperatures thanmost existing commercial nuclear power plants. Thesedesigns will likely utilise both existing and newmaterials to operate under these conditions. However,
the intended operating conditions are still uncertain
and significant gaps are associated with the short- andlong-term performance of candidate materials in theseapplications. Creep, creep-fracture and creep-fatigue
behaviour are prominent gaps. Many ANLWRs operatein salt or highly corrosive environments such that
corrosion control and stress corrosion cracking areimportant degradation considerations. See Chapter 2
for more information on the related design
considerations for proposed ANLWRs.

New irradiationand testing
facilities ordemonstration
reactors will beneeded toappropriately
simulate the hightemperature,irradiation andenvironmental
conditions inorder todemonstrate andqualify material
performance.

Addressing ageing issues otherthan those associated with
irradiation effects may be partiallyachievable using existing
facilities. Facilities to address
ageing due to the aggressive
environments being consideredin some proposed reactor designs
may also be costly.

* During commissioning of Olkiluoto 3 (EPR), hot functional tests at full temperature and pressure were performed, during which strong vibrations
were observed in pressuriser surge line. Analyses that have been done since have not yet revealed with certainty the root cause of the vibrations.
The current assumption is that the excitations in the primary circuit match the same frequencies as the characteristic frequencies of the surge line.
The excitations can be acoustic, flow-induced or mechanical and combinations thereof. According to preliminary analyses, the vibrations would
not threaten the integrity of the surge line even during the design lifetime of 60 years. Furthermore, the break of surge line is included in the designbasis of the unit. However, steps have been taken to damp these vibrations to values considered in the design to increase safety margins. This
example highlights the importance of fluid structure interaction and needs for capabilities to model these interactions.
Note: PHWR: pressurised heavy water reactor; PWR: pressurised water reactor; ALWR: advanced light water reactor; APHWR: pressurised heavy water
reactor; MSR: molten salt reactor; HTGR: high-temperature gas reactor.

3.4. Issues and facilities not unique to the nuclear industry
3.4.1. Human and organisational factors
Introduction
The importance of human and organisational factors in nuclear reactor safety has beenrecognised for decades. Due to a series of accidents in the 1980s, it was increasingly acceptedthat human behaviour could not meaningfully be understood in isolation from the context inwhich it takes place, and that in complex systems, activities had to be organised in a way thatmatched human characteristics. Research in this area involves non-engineering disciplines
such as social, psychological and organisational topics, and aims at generating insights about
the interactions between humans, technology and work organisation, to help promote safety.

Scope
Human and organisational factors apply to both currently operating and future plants, andseveral areas may benefit from further research: staffing, human system interface, external
influences on safety, operating experience, management of unanticipated events, managementof decommissioning, as well as teamwork, conduct of operations and crew roles.

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3. TECHNICAL ISSUES AND ASSOCIATED FACILITIESDescriptionCurrently operating plants

For currently operating plants, operating experience can identify root causes and contributing
factors of events, while human reliability analysis identifies human error-related events to help
enhance safety performance. From such analyses, solutions and remedial actions can be
identified.



Plant refurbishment

Several operating plants are currently being upgraded to include digital solutions. Early
consideration of human and organisational factors in the design process is important in order
to contribute to safe human performance in upgraded plants. Consideration of human andorganisational factors is also crucial with regard to control room and human system interface
verification and integrated system validation.



New reactor design

For future plants, knowledge is needed regarding opportunities and potential new human-failure
mechanisms when digital and modernised control rooms are introduced. Empirical research iscentral for acquiring knowledge that can be used to guide nuclear safety in future plants.



Organisational factors

In addition, external organisational factors may influence human performance, and such factors
should be considered regarding their impact on reactor safety. Organisations, social systems andways of operating within organisations are evolving; an example is the liberalisation of theelectricity markets in most countries. Globalisation of corporate activities and increased
interactions with different cultures also introduce new organisational structures and behaviour
patterns.
Table 3.7 shows the human and organisational factors safety issues that could benefit fromadditional research in current and future plants.

Table 3.7. Current human and organisational factors issues and facilities
IssuesDescription

Facility name1) Staffing

New designs are incorporating passive safety features andemploying more automation. They are also being
Halden Man-Machine
designed with longer response times. Accordingly, theLaboratory (HAMMLAB)
role of the operator and the number of operational staff(Norway)
are changing. How to decide on correct staffing levels
remains an issue. In addition, the analysis of new tasks
Human Systemand the qualification of the staff are to be considered. In
Simulation Laboratory
both new and existing plants, the effect of staffing cuts
(HSSL) (United States)
should be investigated. Many or most plants have cut
staff as a result of market liberalisation and deregulation.

Ability to
Halden MTO
simulate control programme,room
includingenvironment
HAMMLABand conduct
continues
with plant
beyondoperators.
2020.

2) Humanmachineinterface

As more plants upgrade or introduce advanced
instrumentation, issues related to how humans interface
with the system must be addressed. This includes issues
such as:– the role of the human vs. automation;– navigation through software controlled displays;– inputting commands.
Testing of new human-machine interfaces will be useful.
This should include verification and integrated systemtests.

Halden Man-Machine
Laboratory (HAMMLAB)(Norway)
VTT Technical ResearchCentre of Finland
Human System
Simulation Laboratory(HSSL) (United States)
CANDU Mock-Up
Darlington Energy
Centre (Canada)

Ability to
Halden MTO
simulate control programme,room
includingenvironment
HAMMLABand conduct
continues
with plant
beyondoperators.
2020.

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Capability

Notes

75

3. TECHNICAL ISSUES AND ASSOCIATED FACILITIES

Table 3.7. Current human and organisational factors issues and facilities (cont’d)IssuesDescription

Facility nameCapability

Notes3) Organisational
influences

Organisational factors, market liberalisation andderegulation can influence the performance of sociotechnical systems and have negative effects on safety.
Factors such as communication, organisational
learning, safety culture, etc. influence employees’knowledge and behaviour and have been fundamental
factors in actual accidents. Although facilities are not
relevant for this issue, understanding, monitoring andaddressing these factors can improve safety.
Accordingly, developing and maintaining expertise is a
key concern. Objective measures of organisational
performance would also be useful.

-

-

-

4) Human
performances
model

Developing reliable models of human performance will
greatly enhance the accuracy of risk assessments andthe ability to evaluate human-related issues(e.g. procedures, training). Expertise, rather thanfacilities, is the critical need.

Halden Man-Machine –Laboratory(HAMMLAB) (Norway)
Human System
Simulation
Laboratory (HSSL)(United States)

5) Review ofoperatingexperience

Many events at operating reactors have as their initiator
or as an important element in the event human and/ororganisational factors. A review of operating experience –to identify and correct those human and organisational
contributors is key to maintaining/improving safety.

Management of unplanned/unanticipated events areneeded throughout the whole accident sequence (PSA
levels 1, 2 and 3). Components required are:– decision making under uncertainty;
6) Management of– training for the unforeseen, to help establish mental
unplanned/unantic
models/schema to support projections, diagnosis andipated events
detecting anomalies;– guidance, support tools.
Events such as fire or flooding can lead to specificchallenges such as main control room abandonment.
7) Management of
dismantling/
decommissioning
operations–

Halden MTO
programme,
including
HAMMLAB
continues
beyond2020.

–

HAMMLAB (Norway)
Human System
Simulation
Laboratory (HSSL)(United States)

It would be of value to dedicate an old installation for
HBWR
testing practices, methods, tools, etc. applicable todismantling, in an international project, in the same
(IAEA training
way as the experimental control room of the HAMMLAB centre facility)
is used in Halden for testing various HMI.

New technology
such as VR, AR for
training fordecommissioning

Assuring and improving the safety culture in complex
projects requires a systemic approach for leading safety
culture change, based on contemporary views of safety
8) Safety
management and science and management research. System dynamics
VTT Technicalmodelling offered in an interactive simulation gamesafety culture,
Research Centre of
could be used to support decision makers in gaining a
including in
Finlandholistic
perspectiveonpatterns
ofdynamic
inter-organisationa
l project networks interrelations and potential safety effects in projects.
How to benchmark standards in safety culture –considering standards are not clearly defined.
9) The role ofautomation and
How do we decide when to have shared control
personnel: New
between humans and automation and when to assign
concepts offinal authority to humans?
teamwork in
advanced systems

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3. TECHNICAL ISSUES AND ASSOCIATED FACILITIES

Table 3.7. Current human and organisational factors issues and facilities (cont’d)IssuesDescription10) Evolving
concepts for theoperation of nuclear
power plants

What are appropriate models of teamwork for multiagent systems?

11) Managementof operational
resilience

To identify the key characteristics of operational
resilience and study how these characteristics are
manifested in operator behaviour and how they canbe operationalised and measured.

New types of nuclear power plant control andoperation system prototype development,
i.e. operation concept research, MCR design, etc.
12) Nuclear power
Advanced HSI design, verification and application.
plant human factor
Advanced operator supporting system research andengineering (HFE)
development, i.e. computerised procedure system,
design andonline monitoring system, etc.
verification
Human performance research and analysis.
Other HFE research, analysis and application.
HFE-related safety review work.
What are the strengths and limitations of new HFE13) Human factors
methods and tools, and what criteria should beengineering methods
applied to evaluating their acceptability for use in
and tools
human factors R&D.
14) Complexity issues How to assess and minimise the complexity of a newin advanced systems design.
How crews operate and interact in new andupgraded control rooms.
Input to conduct of operations:15) Teamwork, crew – “out-of-the-loop” challenges with regard to plantroles and conduct of automation;
operations– crew roles and teamwork;– crew composition in new plant designs such as
small modular reactors;– independent peer check.

Facility nameCapability

Notes

VTT Technical ResearchCentre of Finland

Ability tosimulate control
State Power Investment
Long-time
room and plantCompany (SPIC)
operation,
local
Nuclear Power Plant
as aenvironment,
Human Factors
company
and conduct
Engineering Laboratory
researchwith nuclear(China)laboratory.
power plant
operators.

HAMMLAB
Human System
Simulation Laboratory(HSSL)CANDU Mock-Up
Darlington Energy
CentreA topic closely tied to human and organisational factors is maintaining expertise on reactor
design and operation, including emergency operations. Human actions are needed to assess
reactor safety and design and to control nuclear reactor operation including transient (or offnormal, including accident) situations. Urgent action may be required in case of a system crisis.
This may require rapid access to detailed information through interactive media. Plant simulatorsand plant response software for emergency operations centres play a crucial role in this response.
An example of one of several facilities is NUTEMA, whose software has been designed toaccess comprehensive and systematic storage of knowledge at the basis of the design and safety(including licensing) of nuclear reactors, coupled with access to any piece of information(D’Auria et al., 2011). The reference information is organised into topics which include details
such as blue-prints (now computer aided design drawings), 3D views, CFR (US Code of Federal
Regulations) statements, experimental data, chemical compositions, thicknesses, notes frominstallation and maintenance of any system, results of numerical codes application, manuals of
those codes, features of turbulence, chemical products of fire, etc.
The NUTEMA hardware consists of several dozen screens arranged in a hierarchic
configuration suitable to be handled by one or a team of two to three specialists. The standard
views for all the screens, at any time, provide the outlook of a given topic. Each screen is the tipof a “virtual pyramid” ending in more in-depth sections: accessible by the user, one may imagine
several virtual screens as axial sections of the pyramid covering the available knowledge andconnecting with other screens. Basically, any small “cube-of-information” part can be dug out
following one or more coherent paths.
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3. TECHNICAL ISSUES AND ASSOCIATED FACILITIESThe presence of interactive screens in the NUTEMA facility make possible:•

the training of engineers and technicians, including nuclear power plant operators andanalysts, as well as regulators;•

the working mode as a remote crisis centre should an event occur in an assigned reactor.

3.4.2. Plant control and monitoring
Introduction
All nuclear power reactors require plant control, and monitoring and protection systems(commonly referred to as I&C systems), and there is growing use of digital instrumentation andcontrol I&C systems.
The issues associated with plant control and monitoring centre around ensuring that systems
continue to perform reliably as they age, or are subjected to harsh conditions following an accident,
and that replacement systems meet reliability goals. For the former, the challenges are to ensurethat degradation mechanisms are understood and mitigated, and systems are appropriately
qualified for post-accident operation. For the latter, in many cases it is neither possible nor
desirable to replace existing systems with equipment of a similar vintage and capability. In thatregard, the increasing use of digital I&C presents both opportunities and challenges. The primary
opportunity is to replace systems with new equipment with enhanced functionalities. The
challenges are to ensure that the new systems perform with equal or better reliability.
Demonstration of reliability requires consideration of hardware and software performance.
A particular concern is that while enhanced functionality has benefits – for example, the use ofsmart systems that have some assessment capability to improve operator response – it also has
the drawback of increased complexity that makes reliability difficult to ensure.

Scope
The I&C areas within a nuclear plant can be divided into the following three categories (in orderof decreasing safety significance and increasing functionality and complexity):•safety (or protection) systems, primarily responsible for mitigating against theconsequences of failure of other plant systems;•

control systems, primarily responsible for maintaining the operating state of the plant;•

monitoring systems, primarily responsible for collecting, logging and presenting current
or past data on the status of plant systems.

Table 3.8 lists the plant control, monitoring and protection system safety issues that could
benefit from additional research. There has been little change to the issues since the last time thisdocument was updated; however, regulations regarding I&C safety have advanced significantly.

Description
Complex I&C systems are used to control complex nuclear power plants, and there is an issue
with the need to model, or to simulate, I&C performance as part of safety and licensing analyses.
Early nuclear power plants had relatively simple I&C and, as far as applicable, each related
component was tested to show no impact or safe impact upon reactor transient (or off-normal
and including accident) evolution (or performance).

78•

More and more I&C systems have been added: the complexity prevented comprehensive
and systematic testing of interaction with transient reactor performance (here one could
recall the recent case of the Boeing 737-800).

•

Digital I&C transfers the safety issues related to analogue I&C and shifts the problem toa different region of phase-space (e.g. fast reaction of individual components) concernedby stability and bifurcation analysis, should any transient event occur.

•

Proprietary information of I&C prevents the possibility of independent assessments fromscientists who are not staff of vendor or designer organisations.
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3. TECHNICAL ISSUES AND ASSOCIATED FACILITIESThe I&C systems following any postulated initiating event, that is, events not requiring
immediate SCRAM and the ATWS, have the potential to bring the operational status of thereactor to a condition that is unpredictable by safety analysts: this may impact the subsequentevolution of the accidents. I&C shall also be considered within the overall context of the defence
in depth where the single failure principle is applied. Therefore, both the “postulated initiating
event driving away” and the “single failure applications” impose the need to model I&C in safety
analyses. As a side issue, avoiding the simulation of I&C prevents the possibility of performing
independent assessments, or the fulfillment of another principle in nuclear reactor safety(D’Auria et al., 2012).
The challenges are to ensure that the new systems, primarily digital ones, perform withequal or better reliability. Demonstration of reliability requires consideration of hardware andsoftware performance. A particular concern is that while enhanced functionality has benefits –for example, the use of smart systems that have some assessment capability to improve
operator response – it also has the drawback of increased complexity that makes reliability
difficult to ensure. The challenges are the same for every reactor type, but it should be notedthat remote and off-grid reactors (SMR) face additional challenges because of their need forremote monitoring and the implication of remote operating and monitoring with respect tosafeguards, security and cybersecurity.
Common position papers have been developed by the Digital Instrumentation and Controls
Working Group (DICWG) of the Multinational Design Evaluation Programme to obtain a consistentunderstanding of issues regarding digital I&C. For example, in 2016, it was agreed that a common
position should be adopted regarding hazard identification and control for digital I&C systems
given the increase of use of digital I&C in new reactor designs, its safety implications and the needto develop a common understanding from the perspectives of regulatory authorities. This action
followed the DICWG’s examination of the regulatory requirements of the participating members
and of relevant industry standards and IAEA documents.
Several regulators worldwide have used the Multinational Design Evaluation Programme’s
common position papers as a framework against which to revise current, or inform new,regulation regarding digital I&C. Table 3.8 identifies issues for nuclear power plants and thefacilities available to address them.

Table 3.8. Current plant control and monitoring issues
IssuesDescription

1) Software quality
and reliability

With the increasing use of digital instrumentation and control (I&C) andprotection systems in currently operating plants, the extensive plans forcomplete control room retrofits using digital systems and the plans for their use
in future plants, how to ensure the quality and reliability of the software used toperform safety functions is a growing concern. Software verification andvalidation methods, as well as qualitative and qualitative software testing
methods, need to be assessed and their attributes and effectiveness establishedto aid in the review and regulation of software-based systems important tosafety.
The environment in which they operate (e.g. temperature, humidity, radiation,smoke, electro-magnetic/radiofrequency interference, etc.) can affect theperformance (reliability, failure rate and failure mode) of digital I&C systems.
Several standards developed by the IEEE and the IEC are used to guidequalification testing. For future plants with different environmental conditions,
new methods and tests will be needed to establish failure thresholds andmodes of failure.
To understand the performance of digital systems, the integration ofsoftware, hardware and humans is needed. Although checks of the various
system components individually are also required, they are not sufficient toconfirm overall system performance and reliability. Facilities where suchtesting can be done in prototypical fashion are needed. Additionally,methods are needed to support the integration of digital system reliability
models into current generation probabilistic safety assessments.

2) Environmental
qualification

3) Digital system
reliability

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University of Virginia – Center for
Safety Critical Systems(United States)

Sandia National Laboratory –environmental qualification lab(United States)

University of Virginia – Center for
Safety Critical Systems(United States)793. TECHNICAL ISSUES AND ASSOCIATED FACILITIES

Table 3.8. Current plant control and monitoring issues (cont’d)IssuesDescription

4) Wirelesscommunication

The use of wireless communication for monitoring and control in nuclear power
plants is expanding rapidly. The technology that supports the current
generation of wireless applications was not designed for the challenging
environments in nuclear plants that have the potential to disrupt signals.
Research is needed to understand the possible effects before such
communication is used for safety functions. Testing is needed to confirm design
and performance. Additionally, the security aspects of wireless communicationneeds to be explored.
The use of advanced online monitoring and systems diagnostics in nuclear
power plant instrumentation and control systems has added a higher level ofcomplexity to the current generation I&C and protection systems, in addition to
the complexity already added by the use of digital systems. Although thesesystems have the potential to reduce operator workload and increase system
reliability, these systems’ new and complex failure modes need to beinvestigated. Facilities where these systems can be tested and reviewed in a
prototypical fashion are needed.

5) Online
monitoring andadvanced
instruments

CNL National Innovation Centre
for Cybersecurity (Canada)

Sandia National Laboratory –environmental qualification lab(United States)
Ohio State University – INL
Academic Center of Excellence in
Nuclear Instrumentation andControl and Safety Analysis(United States)

3.4.3. Cybersecurity
Introduction
New nuclear facilities and the modernisation of existing ones have led to a prominent role for
digital systems to control and guard the safe operation of nuclear reactors. As a consequence,cyberattacks on these systems can have serious consequences on nuclear safety, on economics,
due to down time, and on public perceptions of the safety and security of the nuclear sector. The
evolutionary path of digital security mirrors that of digital safety, which, over a 30-year period,transitioned from new standards to regulation. Nuclear cybersecurity is just entering theregulation phase and is creating a demand for accredited and certified products and services. In
the early 1980s, when technology moved from analogue to digital components, the transition
created a multi-billion dollar industry in the “safety” of software and led to the development ofsoftware safety standards and “safety” devices that are now mandatory in any industrial
environment where safety functions are performed. Similarly, digital security concerns andevolving standards and regulations are driving the demand for safe and secure solutions by bothbusiness and industrial consumers. By 2020, the cybersecurity industry is predicted to reach
USD 170 billion.
Unlike IT systems, industrial control systems are often bound by strict regulatory
requirements and rigorous change controls that introduce complexities, risks and costs that aresubstantially higher for any modernisation or modification and therefore requires specialists in
I&C systems with cybersecurity expertise.

Scope
The risk of cyberattacks has increased significantly over the last decade. Several factors play arole in this increase:80•

Threat actors: Digital attacks have evolved from hobbyist activities by hackers to
profitable criminal activities and even state-sponsored research and development ofdigital weaponry for both defensive and offensive purposes. In order of competence, themain threat actors are nation states, cybercriminals and terrorist groups, and hacktivists.
Unfortunately, all threat actors are relevant for the nuclear sector.

•

Connectivity: The Internet of Things is growing rapidly, leading to all kinds of devices
being accessible via networks and thus becoming potential gateways for network
intrusion. Many of these devices have service lifetimes of ten or more years, but limited
capabilities to protect themselves against digital threats that will emerge during theirservice lifetime. I&C systems as used in nuclear reactors are no exception.

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3. TECHNICAL ISSUES AND ASSOCIATED FACILITIES•

Dark web: A black market of hacking software, vulnerabilities unknown to vendors (zerodays) and even hackers-for-hire is growing rapidly.

Alas, the cybersecurity risk is real and needs to be managed on several levels:•

Governance and policy framework: Nuclear operators must develop a governancemodel and a set of organisational measures, policies and procedures to guarantee thatcybersecurity risks are managed and cybersecurity measures are implemented,
maintained and regularly reviewed. As not all rules can be technically enforced,
controlling the human factor by clear procedures (and awareness) is paramount.
–•

Network security: Networks of nuclear operators must be segmented into network zones
with decoupling, access control and filtering mechanisms at zone borders making surethat sensitive equipment such as industrial control systems reside behind firewalls withrespect to the normal business network to provide a trusted network area with aminimised threat landscape. Network activity crossing zone borders must be monitored.
–•

Relevant standards: ISO 27K series, IAEA NSS-17.

Relevant standards: IEC 62443, IAEA NSS-17, NIST SP 800-82.

Asset risk management: The impact of a successful cyberattack depends strongly on therole that a particular device plays in a nuclear installation. Therefore, the potential impactof losing the confidentiality, integrity and availability of a device should be evaluated in its
context. Moreover, the impact of losing integrity and/or availability can increase with thetime the device does not play its intended role. Hence, the impact of not having a device
in its normal state in function of time together with the time required to bring a rogue
device back to a normal state should be evaluated as well. For critical digital I&C systems
for which loss of availability or integrity could lead to an unacceptable radiologicalconsequence, a secondary analogue safeguard should be considered.
–

Relevant standards:–•

Asset security: Security feeds provided by vendors must be followed in order to be awareof the potential impact of vulnerabilities that can be exploited by threat actors.
Application of security patches should be integrated in the normal maintenanceplanning taking into account the risk induced by the vulnerability. Physical access isalways a threat, but easier to control than logical access. For critical systems which are
known to be vulnerable, logical access should be reduced to physical access by making
sure that logical access can only be achieved from locations where physical access isstrictly controlled and monitored.
–•

NIST: SP 800-39 (Corporate Risk Management), FIPS 199/SP 800-60 (Categorization),
FIPS 200/SP 800-53 (Security Controls), 800-34/61/82/128/… (Implementation), SP
800-53A (Assessment/Monitoring Controls), SP 800-37 (Authorize/Monitor), SP
800-137 (Monitor Controls).

Relevant standards: ISO/IEC 15408 Common Criteria.

Incident response: Nuclear operators must have a cybersecurity incident response plan
describing the actions to be taken in order to contain and mitigate the impact of acybersecurity incident. In addition, a communication plan should also be developed toreport the state of affairs to relevant stakeholders and, if required, communicate withthe press.

As prescribed by all standards, a risk-based, graded approach should be applied: more risk
implies more security measures to minimise the probability of a threat manifesting itself and/or
the impact of the threat when it occurs. Protection of information systems boils down toprotecting the three main pillars of information security: confidentiality, integrity and availability– the so-called CIA. The priority of these pillars strongly depends on the environment: in abusiness network, confidentiality typically tops the list, while in an industrial network, availability
is absolute key. Hence, security measures should always be tailored to the specific needs and risksof the equipment and processes present in the environment.

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3. TECHNICAL ISSUES AND ASSOCIATED FACILITIESDescription
There is one encompassing issue in this area, that of cyberthreats to the security of major digitalassets. A main shortcoming is cybersecurity certification on the asset level. The ISO/IEC 15408
Common Criteria Standard is a major step in this direction, but has a number of shortcomings
in providing the assurance required for nuclear operators. It does provide assurance that the
specifications provided by vendors are rigorously implemented and evaluated in a standardised,
repeatable manner, but leaves too many openings for vendors to restrict the scope of theevaluation. In particular, a vendor can select specific security attributes to be evaluated andspecify (potentially unrealistic) assumptions with regard to the operating environment of theirproduct and the threat landscape to which the product is exposed. Hence, such a certification
only provides assurance for the selected attributes and only if the device effectively operates inan environment adhering to the evaluated configuration, i.e. the specific circumstances
provided by the vendor for evaluation. In addition, the effort and time necessary to prepare allof the necessary documents before, during and after a Common Criteria evaluation makes it avery costly process, which takes so much time that the evaluation criteria and sometimes eventhe product itself are obsolete before the process is complete.
A number of test beds exist where the resilience of digital assets against cyberthreats canbe tested in a controlled environment. The main challenge for such test labs is to configure thetesting environment to emulate the real operating environment as closely as possible. Table 3.9
gives a non-exhaustive list of such labs.

Table 3.9. Cybersecurity research facilities
Facility

Description

Reference

EPIC: Experimental
Platform for Internet
Contingencies (Joint
Research Centre,
European Commission,
Ispra, Italy)

Christos Siaterlis (2013)
Test bed for networked critical infrastructure capable of emulating the real
operating environment. It uses Emulab to emulate the topology and typicalcomponents of the network part of the operating environment such asservers, switches, routers, etc. The physical part of the operating environment
is simulated by a generic PC with multi-tasking operating systems runningreal-time simulations of the physical components. The simulation software
itself is built using Matlab Simulink and Matlab Real Time Workshop. At thetime of writing, this test bed was only open for organisations collaborating
directly with the Joint Research Centre, but they formally stated that thisaccess policy will change in the future.

Cyber Security Platform(CEA-Leti Grenoble andList Paris-Saclay,
France)Platform staffed by 100+ experts in security of integrated circuits, embedded www.leti-cea.com/ceasystems and mobile devices. Uses advanced tools to identify vulnerabilities
tech/leti/english/Pages/Appliand develop innovative ways to protect systems against cyberattacks, such as ed-Research/Facilities/cybersecure communication platforms for sensor networks and cryptographic
security-platform.aspx
implementations of hardware and software. Its Information Technology
www.leti-cea.fr/ceaSecurity Evaluation Facility (ITSEF) [3] is a security evaluation laboratory
tech/leti/english/Pages/Indus
certified by the ANSSI (French Cyber Security Agency) to conduct CSPN of
trialhardware products (first level of ANSSI) and also accredited to do Common
Innovation/Innovate%20with
Criteria evaluations up to EAL7 for secure components with embedded%20Leti/ITSEF.aspx
software, hardware devices with security boxes, and manufacturing andproduction sites. In addition, it is also recognised as an evaluation laboratory
for other certification bodies like EMVCo, VISA MASTERCARD, NXP MIFARE
and BAROC. It is one of the laboratories that offers security tests of electronic
components and equipment for industrial developers and also performs
security audits of design and production sites for secure products.

Cyber Security Training Cyberlab that according to THALES is capable of reproducing information
Thales (2017)
Centre (Thales, Tubize, networks of an organisation to test its resistance against cyberattacks. It offers
Belgium)
validation of the security level of information system architectures for its clients.
Its test bed was delivered by DIATEAM, a French company that is a pioneer invirtualisation and simulation platforms for cyberdefense (both civil andmilitary).

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3. TECHNICAL ISSUES AND ASSOCIATED FACILITIES

Table 3.9. Cybersecurity research facilities (cont’d)
Facility

Description

Reference

HNS Platform(DIATEAM, Brest,
France)Hybrid virtualisation and simulation environment allowing to combine real and www.hns-platform.com
simulated devices in a controlled environment for cybersecurity tests andbenchmarking. It is capable of simulating IT and industrial control systems
environments for several purposes related to cybersecurity:– learning and training: simulated attacks in Red Team (attackers) and Blue
Team (defenders) setup with realistic network traffic generators;– analysis and prototyping of architectures in design or pre-production phase.

Cyber Security Lab(Sofia Tech Park, Sofia,
Bulgaria)

Cyber security research group that, among other information security R&D
activities, tests and audits the security of components, systems and evenorganisations. It is capable of simulating cyberattacks on complex
infrastructures and systems.

https://sofiatech.bg/en/activit
ies/laboratories/laboratoriyapo-kibersigurnost/

SCADA Lab (Scada
Laboratory, INTECO
Data Centre, León,
Spain and SCADA test
bed, HQ Telvent
Energy, Seville, Spain)

Test environment for assessing the security of an industrial control system
environment split up into two sites: the laboratory area and the test bed area.
The laboratory area hosts the hardware and software to launch a security
assessment against the test bed that contains the real-time industrial control
systems and physical systems. The testing methodology is based on a goodpractice guide for assessing the cybersecurity of industrial controls systems of
the Centre for the Protection of National Infrastructure of the UK government[8] and financed by the European Commission. In [7], the authors state that
operators can request a security assessment and that their framework is
capable of providing a “first view” of the security level of critical infrastructuresin a detailed, fast and simple way.

Sánchez Aragó, Redondo
Martínez and Salán Clares(2014)CPNI (2011),
Stouffer Security Research,
2014

NIST Cybersecurity
Testbed for Industrial
Control Systems(Gaithersburg,
Maryland, UnitedStates)CPNI (2011)
The main goal of this test bed is to measure the cyber resilience of industrial
control systems setup and instrumented with cybersecurity protections
R.C.T.Z.K. Stouffer
according to best practices, in particular the recommendations in IEC 62443
and NIST SP 800-82. It is capable of emulating real-world industrial systems anda variety of industrial scenarios with slow and fast dynamics. Its main purpose is
the validation of existing security guidelines and standards, but it also serves asa platform for academic, governmental and commercial research to experimentwith security technologies and high assurance designs.

CNL National
Innovation Centre for
Cybersecurity (New
Brunswick, Canada)

Focusing on industrial control systems, this facility contains a distributed
control systems testing platform equipped with industrial controllers tosimulate real-time process control functions which include a representative
suite of field instrumentation such as smart devices and industrial
communication network protocols. Technologies to be investigated include:boundary protection devices (e.g. firewalls), network intrusion protection andanomaly detection, endpoint protection, security information and event
management tools, application sandboxing, vulnerability scanners, multiengine virus scanners, network devices (e.g. smart switches, network taps),
virtual servers, penetration testing tools.
Security operations centres equipped with situational awareness tools anddashboards that provide real-time monitoring, analysis, alerting and reporting
capabilities.

CNL (2018)

Many guidelines and standards exist with regard to managing cybersecurity threats ofindustrial control systems that are also applicable to the I&C systems of nuclear facilities. IAEA
NSS-17 “Computer security at nuclear facilities” provides an interpretation of the two mostprominent standards for cybersecurity relevant for nuclear operators specifically for the nuclear
sector: ISO 27001 standard for information security management systems and IEC 62443 for
network segmentation. Even when nuclear operators follow these recommendations to the letter,they are still for a large part dependent on the cyber resilience of the digital equipment they
employ in the complete chain of a nuclear facility: firewalls, routers, switches, servers, humaninterface machines and software, industrial control systems, and so on. At the moment, ISO/IEC
15408 Common Criteria is the only standardised certification for cybersecurity that is generally
accepted, but it leaves too many openings for vendors to restrict the scope of the evaluation to

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83

3. TECHNICAL ISSUES AND ASSOCIATED FACILITIES

provide the assurance that nuclear operators require. In addition, the Common Criteria
certification process is costly and so time-consuming that certification of a particular device
often comes once the devices are already operating in the field. Cybersecurity test environments
can fill this gap by simulating the setup and components of real-world industrial infrastructures
as accurately as possible, allowing to test the behaviour and resilience of these infrastructures
against known cyberattacks without the manifestation of the associated real-world risks.

3.4.4. External events
Introduction
External events have the potential to impact plant safety by simultaneously affecting many plantsystems, structures and components. To ensure that plants are designed for such events, data to
confirm the design and safety evaluation methods are essential, depending on experimental
facilities and simulations.

Scope
There is a wide range of external events, including natural phenomena like earthquakes,
flooding and extreme weather phenomena, as well as man-made events. These phenomena are
addressed by an equally wide spectrum of experimental facilities.

DescriptionSeismic events

Issues related to the seismic behaviour of components and structures are applicable to all
reactor types, both currently operating and future plants. The magnitude of seismic events forwhich plants must be designed varies across member countries and with plant age, since
seismic concerns are site-specific and methods are evolving. It is expected that future plants(which are likely to be standard designs to be marketed worldwide) will be designed to higher
seismic standards to enable them to be sited in many member countries. Also, future designs
may employ new features to improve plant seismic safety (e.g. below ground structures, seismic
isolation devices), which will need experimental confirmation. Table 3.10 lists the safety issues
associated with the seismic behaviour of components and structures that could benefit fromadditional research.

Table 3.10. Current seismic effects issues

84

Issues and relevant reactors

Description

1) Confirmation of seismic
design

New plant designs are incorporating new safety features (e.g. passive emergency corecooling system, passive containment cooling) that need to be designed to withstandseismic events. Also, some designs may incorporate seismic isolation features to limit thetransmission of ground motion to plant structures and equipment. In both cases,
experimental confirmation of the design’s ability to withstand seismic events and data tovalidate analytical tools will be necessary.

2) Below ground siting

Some future designs may locate all or some critical systems, structures and components
below ground to protect them from external events. The response of below groundstructures to seismic events needs experimental data to confirm analysis methods.

3) Continued safe operation

As seismic events continue to occur and plants continue to age, data to confirm
continued safe operation may be necessary. These data could be in the form of simulating
the earthquake and the aged plant structure.

4) Seismic isolation devices

Some future designs may incorporate seismic isolation devices into the design to reduce
the seismic ground motion transferral to vital plant structures. Experimental confirmationof the performance of these devices should be obtained.

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3. TECHNICAL ISSUES AND ASSOCIATED FACILITIESShaking tables are used for investigating seismic effects on buildings. Most of them do not
specifically deal with nuclear installations, with the exception of the SHAKESPEARE shaking tablein Sint-Genesius-Rode near Brussels, Belgium, which was explicitly set up to address issues of
future reactor designs (maximum load: 500 kg). This facility, however, is currently not at risk.

Other external events
Investigations of flooding and tsunami scenarios rely on high computer capacity used formodelling. A prominent example is the Tsunami Assessment Modeling System at the Joint
Research Centre in Ispra (Italy). In addition, there are experimental facilities for storm surges andtsunamis, which allow addressing phenomena like runup, overland flow and wave-structureinteraction, like Oregon State University's multidirectional wave basin.
The impact of projectiles against the walls of nuclear facilities may be caused by strong
wind (tornadoes) or by man-made activities like an airplane crash. Damages resulting fromtornado-borne debris were studied in the 1970s by Sandia Nuclear Laboratories (United States)
using reinforced concrete panels. The Impact Testing Facility at VTT (Finland) is an example ofan installation dedicated to investigate experimentally the possible outcomes of airplane crash
scenarios, as in the ongoing international Impact project. Such issues are specifically related tocontainment integrity (see also Section 3.1.5).
In general, facilities investigating the above-mentioned events do not primarily depend on
research of the nuclear sector. Initiatives launched by the nuclear sector do not constituteeffective means to maintain such facilities. In view of its mandate, they were not the righttargets for the SESAR group to focus on, and in line with an efficient approach, they were notfurther considered as candidates for actions to be taken by the CSNI.

Further information about facilities related to external events

Earthquakes

A long list of shaking tables worldwide including their technical details can be found at:https://ipfs.io/ipfs/QmXoypizjW3WknFiJnKLwHCnL72vedxjQkDDP1mXWo6uco/wiki/Earthquake_shaking_table.html. It includes tables bigger than 2 m x 2 m or with a capacity of more than
4 tonnes.
Experimental facilities for earthquake engineering simulation worldwide are addressed
NEA (2004), which lists a smaller number of shaking tables, but also more than 30 reaction walls
worldwide.
Experimental facilities in Europe (shaking tables and reaction walls) are listed at:https://sera-ta.eucentre.it.



Tsunamis and other flooding phenomena

The JRC Tsunami Assessment Modelling System at Ispra (Italy) is described in EC (2007) URI
https://publications.jrc.ec.europa.eu/repository/handle/JRC41997.
Information on the multidirectional wave basin as part of the O.H. Hinsdale Wave ResearchLaboratory of the Oregon State University can be found at: wave.oregonstate.edu.

Wind-driven projectiles

The test programme on tornado-borne debris by Sandia Nuclear Laboratories (United States)
can be found in Stephenson and Sliter (1977).



Airplane crashes

The VTT Impact Testing Facility (Finland) is described in Heckötter and Vepsä (2015) and Vepsäet al. (2017).

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3. TECHNICAL ISSUES AND ASSOCIATED FACILITIES

3.4.5. Fire assessment
Introduction
Fires present a very demanding generic problem to plant safety, which has been demonstrated bysome serious incidents in the past and by several plant-specific safety analyses. The longexperience of fire protection has resulted in well-known codes and standards and good practices
in design, construction and operation. Fire research has traditionally supported these goals by
producing experimental results on active and passive fire prevention and mitigation. Theoreticalmodelling of fire is very demanding because of its multiple effects and because some key
parameters are not well known. Modelling experience of fires has, however, gradually proceeded
hand-in-hand with experimental work. In particular, zone models and multi-compartment zone
models have shown promising results due to their ease of use and because they are not very timeor memory consuming for simulations of compartment fires. In recent years, progress incomputing power caused CFD models to become everyday tools for engineering applications. Zone
models have been complemented by CFD codes, since CFD models are much more versatile, haveno compartment size or configuration limitations, and allow prediction of all physical variables offires if the fire source is known. Remaining problems include determination of fire sources suchas the complex solid fuels (geometry, fire properties of solid combustibles, homogeneous/heterogeneous materials, porous/non-porous medium, effect of oxygen depletion for ventilated
and confined fire scenarios, and development of reliable models for distributed fire loads like
cables. Brute force methods like solving Navier-Stokes equations numerically are not within theforeseeable future. Promising theoretical and experimental results in micro-gravity and otheraerospace-related industries have indicated that new analytical modelling for the flame spread ispossible. The major efforts should be directed to work on this topic as a near-term goal to
implement them into CFD codes. Particularly, numerical fire simulation has prompted thedevelopment of deterministic and stochastic fire modelling. Comparisons of code predictions with
relevant experiments, benchmarking and other similar international comparisons of the codes
have made the technology fairly reliable in general, and also helped to select the most useful
numerical codes. The development of computational tools and accumulating experience isgradually enabling fire probabilistic safety assessment (PSA) on the same realistic level as in otherbranches of PSA.
However, additional experimental data are needed to assess the codes and new safetyissues have arisen which may need experimental data to resolve.

Scope
The scope of this area includes fire safety issues related to plant design, fire analysis andquantitative fire risk assessment. Table 3.11 shows the fire assessment safety issues that could
benefit from additional research.

Description
Because of the generic nature of fire, nuclear power plant-specific experimental facilities are
not necessary. Most of the experimental work has been conducted using inexpensive smallscale equipment.
Facilitated by the increased calculation power, the computational fire modelling is
progressing fast. Much work in refining codes, models, computing algorithms and model
validation is still needed until the methods are considered reliable enough for safety analysis.
Additional experimental data are needed to fix crucial parameters of the modern fire models. The
major problem is calculating the fire source, in connection with active air-solid interface ofdistributed fire loads like cables. Actions should be taken to test various proposed models of theemerging technology at all relevant scales, and to implement the promising models in CFD codes.
Quantitative assessment of a fire scenario needs to be calculated using Monte Carlotechniques. Most of the needed deterministic fire models already exist, as well as some calculation
platforms. Efforts of creating needed input databanks should be started soon.

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3. TECHNICAL ISSUES AND ASSOCIATED FACILITIES

While the computing and simulation models tools are, to some degree, able to use alreadyavailable knowledge and tools, in practice, several parallel development lines are needed:
databases from the most safety-relevant fire scenarios; fire properties (ignition time, flame spread
velocity, etc.) for the relevant materials; models of automatic fire protection; and quantitative
assessment of manual fire protection.

Table 3.11 Current fire assessment issues
Issues and relevant
Description
reactors

Facilities

1) Fire growth andpropagation

Accurate modelling of fire growth and propagation is key to determining
the time and extent of equipment affected. While the best CFD simulation
models are able to predict consequences for given fires rather satisfactorily,
efforts should be made to simulate most safety-relevant typical scenarios
as benchmarks, and make the results available in databases for plantspecific work. For example, it’s important to study complex fuel such as
electrical cabinets, electrical cables and gloveboxes.
Efforts should be made to utilise the emerging technology of flame spread
modelling on solids. Actions to test various proposed models at all relevant
scales, and implement the promising models in the best CFD codes areneeded. Special efforts are needed to select the most suitable testing
methods from existing or new concepts, which are necessary to determine
flame spread parameters for practical commercial products. For example,
there are no available methods for cables and glovebox. Benchmarking
efforts are needed to transfer the technology from laboratories to industrial
practice. For complex fuel, it is important to combine both analytical andfull-scale R&D to study pyrolysis of each constituent and the behaviour ofthe equipment with all of its constituents in the full-scale conditions of a
fire.

DIVA (France) GALAXIE platform: 5 rooms
between 120 m3 and 160 m3, connected by aventilation system network to investigate fordifferent type of fires (oil, electrical cabinets,
electrical cables, glovebox), heat and smoke
propagation from room to room. Able tomeasure soot deposits or radioactive
surrogates in outlet duct of ventilation
network (to 2021).
PLUTON (France), GALAXIE platform: 1 roomof 400 m3 allowing confined and ventilated
fire tests using different types of fuel.
Ventilation up to 50 000 m3/h for fire power
up to 5 MW. Soot and radioactive surrogate
measurement (to 2023).
US National Institute of Standards andTechnology (NIST): Issues covered include fire
growth and propagation (indefinite).

2) Hot shorts

Fires in cable trays not only cause the loss of the cable, but can also
cause inadvertent signals in control cables affecting equipment. The
likelihood and consequences of such “hot shorts” are not wellunderstood or modelled in safety analysis. Experimental data areneeded. Some data are already available for simple basic scenarios.
Theoretical modelling is needed for assessing effects on systems
performance.

Omega Point (United States)
DIVA, PLUTON

3) Smokepropagation

The spread of smoke during a fire can have strong impacts. Smoke canaffect the operability of certain equipment and inhibit firefighting efforts
by limiting access and visibility. This is not unique to the nuclear
industry; however, the specificity of the confined and ventilated
environment encountered in nuclear power plants requires
experimental data to increase knowledge, improve associated models
and validation codes.

4) Equipment
vulnerability

Understanding when and how equipment fails under fire conditions isessential to evaluate the fire consequences and the consecutive risk
assessment on the nuclear facility. This includes failures from heat,smoke, suppression system activation, shorts, etc. Experimental data will
likely be needed to address this issue.
Some data and basic calculation models are available on heat and smoke
effects for some electrical and electronic equipment. Establishing a
databank with benchmarking examples would be a good way to educate
utilities to use that information. Establishing these databanks is
mandatory for Monte Carlo analyses.

DANAIDES (France) GALAXIE platform test
bench for testing electrical equipment
operating under severe temperature and sootconditions similar to real compartment fire(gas temperature from ambient to 300°C andfrom 0 to 5g/m3 for soot concentration) (to2023).
CISCCO (France) GALAXIE: Experimental device
for testing cable fire in various configurations(nature of cables, cable arrangement on thetray, distance between trays, etc.) (to 2025).
SIMBAG (France) GALAXIE test bench fortesting all the combustion regimes, inparticular air and thermal conditions
dependent on various parameters evolving intime (geometrical evolutions of the glovebox
and the ventilation), exchanges of flow of heat,
relocation of the combustible materials andtheir pyrolysis).

5) High-energy
arcing faults

Fires caused by arcing from high-energy lines need to be modelled andincluded in risk assessments.

N/A

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3. TECHNICAL ISSUES AND ASSOCIATED FACILITIESSánchez Aragó, A., E. Redondo Martínez and S. Salán Clares (2014), “SCADA laboratory and testbed as a service for critical infrastructure protection”, Proceedings of the 2nd International
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4. CONCLUSIONS AND RECOMMENDATIONS

4. Conclusions and recommendations

4.1. Introduction
This chapter provides the conclusions and recommendations of the Expert Group on Support
Facilities for Existing and Advanced Reactors (SFEAR2) regarding key facilities unique to nuclear
safety research in danger of being lost in the short term (in the next one to three years) and thosethat should be monitored by the Nuclear Energy Agency (NEA) Committee on the Safety of
Nuclear Installations (CSNI) in the longer term (more than three years) to ensure a minimum
facility infrastructure to support operating water cooled reactors, including advanced fuelconcept developments in member countries. Unique facilities for new reactor designs are
identified in a cursory manner in Chapter 2. This chapter also provides some general conclusions
and recommendations (independent of short or long term) for the CSNI’s consideration.
The overall strategy employed in developing recommendations was the same as followedin the past; namely, to identify minimum research infrastructure needs and the facilities thatshould be maintained to ensure that this infrastructure is available. As in previous efforts, thegroup took an approach that focused on those facilities with unique capabilities and that
represented a substantial investment of resources such that, in the current climate of reduced
funding for safety research, they would likely not be replaced if lost.
Specifically, the following factors were considered in determining whether or not torecommend CSNI action to preserve a facility or to recommend longer term monitoring of status:•

the importance of the facility to resolving an identified safety issue, within a given
technical area;•

the versatility of the facility;•

the importance of the facility to maintain a minimum infrastructure of safety researchcapability (i.e. uniqueness and replacement cost).

Only facilities with medium or high replacement costs and identified as unique wereconsidered as candidates for CSNI action. The conclusions and recommendations summarised
below are organised by the five technical areas discussed in Section 3.1.

4.2. Short-term conclusions and recommendations
To assess short-term concerns, the facilities in Chapter 3 were examined, and those in dangerof being shut down in the next one to three years were identified. These facilities are shown in
Table 4.1 by technical area, along with the issues and reactors they support. It is instructive tocompare the facilities currently on the list with those identified in the last SESAR report as being
either in short-term danger or warranting monitoring in the long term. These are also displayedin Table 4.1. Facilities that were identified in 2007 as being at risk in the short term appear initalics. Those that are now closed are highlighted in grey. A third category is facilities that werepreviously identified as requiring monitoring in the long term, and are now not identified to be
at risk by member countries, either because more industrial funding has been obtained or
because new facilities have obviated the need. These are discussed in earlier chapters. Actions
being taken to preserve the capability of these facilities are also identified. Finally, the GALAXIE
facility in France appears on the list of facilities at short-term risk, for the first time.

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4. CONCLUSIONS AND RECOMMENDATIONS

Table 4.1. Facilities in danger in the short term (2019-21) and facilities
identified in 2007 as being at risk
Facility

Applicable reactors

Issues addressed

Notes

PWR, BWR, ALWR

Passive safety system
performance.

Identified as being at short-term risk in 2001
but still operating as the result of NEA projects.
A new proposal will be planned before the endof the current HYMERES project.

Thermal-hydraulics
PANDA(Switzerland)Two-phase natural circulation.
Thermal stratification.
Thermal cycling.
Accidents initiated during
shutdown.
PHEBUS (France)

PWR, BWR, VVER, ALWR

Response to loss-of-coolant
accidents.

Closed.

PKL (Germany)

PWR

Boron dilution.

Previously identified as a critical facility tomonitor in the long term. Discussions for a next
phase of the NEA PKL project have started withthe aim of continuing soon after the end ofcurrent phase (September 2020).

Passive safety system
performance.
Steam generator tube ruptures(SGTR).
Two-phase – natural circulation.
Thermal stratification.
Accidents initiated during
shutdown.
LSTF/ROSA (Japan)

PWR

Boron dilution.
Passive safety system
performance.
SGTR.
Two-phase – natural circulation.

Previously identified as a critical facility tomonitor in the long term. Future joint
projects will be discussed with the otherorganisations operating an integral test
facility.

Thermal stratification.
Accidents initiated during
shutdown.
Fuel
Halden (Norway)

All

Previously identified as a critical facility tomonitor in the long term.
Closed.

PHEBUS (France)

All

Previously identified as a critical facility tomonitor in the long term.
Closed.

NRU (Canada)

PHWR

Previously identified as a critical facility tomonitor in the long term.
Closed.

Severe accidents and containment
CTF (Canada)All

Combustible gas control.

Closed.

LSVCTF (Canada)All

Combustible gas control.

Closed.

ARTIST (Switzerland)

PWR, PHWR, APHWR,
ALWR

FP chemistry and release.

Closed.

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4. CONCLUSIONS AND RECOMMENDATIONS

Table 4.1. Facilities in danger in the short term (2019-21) and facilities
identified in 2007 as being at risk (cont’d)
Facility

Applicable reactors

Issues addressed

PHEBUS (France)

BWR, PWR, VVER, ALWR, Pre-core melt conditions.
PHWR, APHWR
In-vessel melt progression.
Effect of air on core melt
progression.

Notes
Previously identified as at risk in the shortterm.
Closed.

Effect of HB and MOX fuel.
Fission product chemistry andrelease.
Coolability of overheated cores.
THAI (Becker
Technologies, Germany)All

PANDA (Switzerland)

BWR, ALWR

Combustible gas control.
Containment integrity.
Combustible gas control.
Containment integrity.

QUENCH (Germany)

PWR, BWR, VVER, ALWR

Pre-core melt conditions.
In-vessel melt progression.
Coolability of overheated cores.

Previously identified as a critical facility tomonitor in the long term. A new NEA project,
THEMIS, has been launched.
Previously identified as a short-term risk. Still
operating as the result of NEA projects.
A proposal for a new joint project will beplanned before the end of the current
HYMERES project (2021).
Facility previously identified as a short-termrisk. A new NEA project (QUENCH ATF) has
been launched.

COMET-FZK (Germany)

PWR, BWR, VVER, ALWR

Core concrete interaction.

Closed.

VERDON (CEA, France)

All

Fission product release andtransport at high temperature.

Previously identified as a facility requiring
monitoring in the long term. A new NEA
project (ESTER) has been launched.

CHROMIA (IRSN, France)

All

Fission product release inPreviously identified as requiring monitoring
primary circuit and containment, in the long term. A new NEA project (ESTER)EPICUR, CHIP and start arehas been launched.
components.

Halden (Norway)

All

Closed.

JMTR (Japan)

All

Closed.

NRU (Canada)

PHWR

Closed.

Halden (Norway)

All

Closed.

Integrity of equipment

OthersGALAXIE Platform (IRSN,France)All

Fire and soot propagation
through ventilation, from cabletrays, in gloveboxes. A numberof large, medium and separate
effects scale facilities co-located.
Unique in ability to examine
cable, glovebox fires andventilation fires of particularconcern to the nuclear industry.

Not previously identified as being at risk. A
new NEA project (PRISME follow-up) is being
contemplated.

Notes: PWR: pressurised water reactor; BWR: boiling water reactor; VVER: water-water energetic reactor; ALWR: advanced light water reactor; PHWR:
pressurised heavy water reactor; APHWR: advanced pressurised heavy water reactor. Gray shading includes those identified in 2007 as being at risk
and that are now closed. Italics identify those previously identified as being at short-term risk, or requiring long-term monitoring.

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4. CONCLUSIONS AND RECOMMENDATIONSIn the thermal-hydraulics area, three facilities are in short-term danger. These include PKL
and the Large Scale Test Facility (LSTF, Japan), which support pressurised water reactor (PWR)
thermal-hydraulic work. PANDA, which supports both light water reactor (LWR) and boiling water
reactor (BWR), as well as containment thermal-hydraulics. Both PANDA (PSI Switzerland) and PKL(Areva, Germany) proponents are anticipated to propose new NEA projects prior to the current
projects coming to an end (2021 for PANDA and 2020 for PKL respectively). The situation for the
LSTF (Japan) is more unclear, pending a decision on whether Japan will continue to support it.
RD-14M, a full-height pressurised heavy water reactor (PHWR) thermal-hydraulics facility was
decommissioned in 2019, in consultation with both the Canadian industry and the regulator who
identified no outstanding issues that warranted maintaining the infrastructure. A similar scaled
facility (but lower power) has been constructed in India, but the status of its operation is unknown.
Numerous facilities have been closed in the containment and severe accident area since
the last evaluation was published. These include PHEBUS, discussed in the fuels and materials
section, combustion test facilities in Canada, the ARTIST aerosol retention facility (Switzerland)
and the COMET facility in FZK Germany. The last report identified the QUENCH facility as beingat risk and concluded that the long-term prognosis would depend on identifying a future
experimental programme that could provide useful information beyond what had already been
done in QUENCH. The facility is scheduled to operate until 2022, and a new NEA joint project on
accident-tolerant fuel is being contemplated, but this facility will remain in short-term risk ofclosure unless the project is implemented.
The aforementioned PANDA facility, the VERDON facility and the THAI facility have been
identified as being at risk in the near term. They had all been previously identified as being at
risk, or critical for monitoring in the long term in the last report. There are NEA joint projects
planned for each. The CHROMIA platform at the IRSN is also at risk. It has been supported
extensively by NEA projects over the past 12 years (STEM) and a new joint project proposal,
ESTER, is being prepared which also utilises the VERDON facility.
In the fuel and materials areas, several large, versatile reactors have been closed. However,regardless of having been previously identified as being short-term risks (PHEBUS) or facilities tomonitor in the long term, substantial national support is required to sustain/refurbish ageing or
specialised reactors. The National Research Universal Reactor in Canada and the Halden Reactor
in IFE Norway both closed in 2018, and this represents a significant loss worldwide, both formaterials and fuel testing. Additionally, the Japan Material Testing Reactor (JMTR) in Japan, a test
reactor used primarily for materials testing, has been shut down as well. Reactors worldwide,
including CABRI, BR-2, LVR-15, MIR, TREAT, HFR, ATR and others have been identified as being
suitable to replace some of the capabilities lost by the closure of these reactors.
A review of sub-critical and zero-power reactors reported on in Section 3.1.3 has found thatseveral of the reactors identified in the last report for monitoring in the long term have beenclosed. The Nuclear Science Committee has recommended that the remainder of the mostversatile (VENUS, ZED-2, LR-0 and KUCA) be monitored closely. The STACEY facility (Japan) was
mentioned as a candidate to replace some capacity, but it is currently shut down. The CROCUS
facility in Switzerland also may have some capacity. Neither of these have been identified as
being at risk, however.
Finally, although the scope of the previous 2007 activity excluded facilities that were notunique to the nuclear industry, the GALAXIE platform at the IRSN was identified in the current
evaluation as being at risk in the near term. The collection of co-located separate effects,
intermediate and large-scale facilities in this platform, as well as experience gained from threephases of PRISME projects make it ideal for addressing fire safety issues unique to the nuclear
industry (e.g. glovebox fires, propagation through ventilation systems, fire initiated in cable trays).
An NEA joint project is planned to be proposed once the current PRISME project ends in 2021.
Despite the large number of facilities that have closed since 2007, there are several whichwere previously on the list as being at short-term risk, but which are no longer on the list. At thesame time, new facilities have been built that can be used to address the issues formerly requiring
the closed facilities.
As an example of facilities which are no longer at risk, in the previous report, both the APEX(PWR, Oregon State University) and PUMA (BWR, Purdue University) thermal-hydraulic facilities

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4. CONCLUSIONS AND RECOMMENDATIONS

were identified as being at risk in the short term, but are both currently operating, and were not
identified as being at risk for the current assessment. This is presumably the result of strong
industry support over the last several years. Also identified at risk were the VULCANO facility
at KTH, which was not identified in the current evaluation as being at risk because of its
continued use by national and EU projects.
An example of facilities that have been closed, but whose research purpose can be at least
partially fulfilled by other facilities is the ARTIST integral facility in PSI, a scaled-down model ofthe FRAMATOME 33/19 SG. Although ARTIST has been closed, some of the more generic aerosol
retention questions can be addressed by the AEOLUS integral facility in KAERI, a scaled-down
model of the SG in OPR1000 & APR1400, which is currently performing aerosol retention studies
in the break vicinity.
Some of the capability of the hydrogen combustion facilities in Canada, although unique intheir scale and versatility, can be fulfilled by the THAI facility operated by Becker Technologies
in Germany (PAR behaviour, hydrogen stratification and mixing). This highlights the importanceof the THAI facility being maintained in the future. Nonetheless, issues such as deflagration todetonation if required by CSNI member countries, will require additional facilities to investigate.
It is proposed that some of the large facilities identified by the EC Network of Excellence for
Hydrogen Safety “HySafe” be assessed, should the need arise (HySafe, 2021).
Despite the large number of versatile reactors that have been shut down since 2007 when
the last report was written, collaborative irradiation programmes, refurbished reactors and newbuilds are, or will be, available to close some of the gaps. The Jules Horowitz Reactor in Franceis planned to start by the mid-2020s and the MYRRHA reactor in Belgium in the mid-2030s. In
addition, the pulse reactor TREAT in the United States has been successfully restarted and the
US Department of Energy has launched its Versatile Fast Neutron Source Project to provide fastneutron testing capability to aid US development of advanced nuclear reactor technology. The
Versatile Test Reactor, as it is also known, could be completed by 2026. Finally, the NEA has
recently launched the FIDES network, with associated irradiation projects (JEEPs), which willallow member countries to access various reactors and test programmes for materials and fuels.
This step towards collaborative irradiation projects is expected to somewhat mitigate theirradiation gap both in the short and long term.

4.3. Longer-term conclusions and recommendations
Many of the factors used in the last two reports to arrive at conclusions and recommendations
have resulted in effective measures for retaining key facilities at risk. These measures should
continue to be used in the future, with consideration of the factors below:•

Cost of facility operation and replacement (i.e. limit CSNI involvement to large facilities
needing multinational support).

•

Consistency with SFEAR-recommended list of facilities for long-term preservation(discussed below).

•

Ability to define a useful experimental programme (i.e. one that will provide information
useful to the resolution of one or more safety issues).

•

Long-term planning to ensure the most important facilities receive the highest priority forlong-term preservation (i.e. not first come, first served). This would include assessing thelong-term resource implications (i.e. consider impact of cost of a co-operative programmeon resources available for other projects) and the host country’s long-term plans for thefacility.

•

Industry participation.

•

Host country commitment.

It is recommended that the CSNI consider each factor when developing a strategy for facility
preservation and in assessing and initiating future co-operative research programmes.

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95

4. CONCLUSIONS AND RECOMMENDATIONSBuilding on previous SESAR reports and the safety issues contained in Section 3.1 of thisreport, a table of critical research facility infrastructure needs was developed, along with a list, byreactor type, of existing facilities that could fulfil those needs. These are shown in Table 4.2 forvarious reactor types. For reference, facilities previously identified but no longer available areshown in grey shading and italics; those that have been identified for the first time are in bold.
The facilities listed are those considered unique, hard to replace and identified as playing asignificant role in resolving issues in their technical area, as discussed in Chapter 3. Accordingly,
it is recommended that the CSNI continue to focus on these facilities in developing a strategy forlong-term infrastructure preservation. The CSNI should monitor the status of these facilities witha goal of taking action, as appropriate, to ensure that critical facilities are available for each reactor
type to meet the critical research infrastructure needs. In addition, for investigating safety issues
associated with the new reactors and technologies identified in Chapter 2, the CSNI should takean active role in encouraging and organising co-operative research efforts. This will also
contribute to infrastructure preservation. Similar to the short-term recommendations above, hostcountry interest will be an important factor in determining which facilities to preserve.
In the thermal-hydraulics area, over the long term (beyond 2022), the Rod Bundle Heattransfer facility will require monitoring. This facility is unique, costly to maintain and has a highreplacement cost. A unique situation exists in Korea, where R&D on pressurised heavy water
reactor (PHWR) topics is on the decline, while advanced reactor design is continuing. This has
prompted the inclusion of the Moderator Cooling Test facility as a thermal-hydraulic facility
requiring monitoring in the long term.
In the containment and severe accident area, the TROI facility, which can be used to study
molten corium interactions with water, has been previously utilised for NEA joint projects but
had not been used for a few years in 2021. It is one of only a few facilities available that can use
large amounts of real corium (20 kg), and should therefore be monitored in the long term as well,
despite the fact that it is not in imminent danger of being decommissioned. The MCCI facility
at Argonne National Laboratory is well advanced in planning new NEA joint projects, and if
these are launched successfully, they will sustain the facilities for another few years. However,there are no plans past this project.

Table 4.2. Critical facilities to be monitored in the long term
Technical area

BWR

PWR

Thermalhydraulics

PANDA1(Switzerland)

Fuels

Reactor physics

VVER

PHWR/APHWR

ALWR

LSTF/ROSA (Japan) PSB-VVER (Russia)
PKL (Germany)
PACTEL (Finland)
ATLAS (Korea)
RBHT(United States)

RD-14-M (Canada)
MCT (Korea)

LSTF/ROSA (Japan)
PKL (Germany)
PANDA1(Switzerland)
ATLAS

Halden (Norway)
NSRR (Japan)
PHEBUS (France)
TREAT(United States)BR-2 (Belgium)

Halden (Norway)
NSRR (Japan)
CABRI (France)
PHEBUS (France)
TREAT(United States)BR-2 (Belgium)

Halden (Norway)
MIR (Russia)
CABRI (France)
PHEBUS (France)
TREAT(United States)BR-2 (Belgium)

Halden (Norway)
NRU (Canada)
TREAT(United States)BR-2 (Belgium)

Halden (Norway)
NSRR (Japan)
CABRI (France)
PHEBUS (France)
TREAT(United States)BR-2 (Belgium)

PROTEUS
Venus (Belgium)

PROTEUS
Venus (Belgium)

PROTEUS
LR-0 (Czech
Republic)

PROTEUSZED-2 (Canada)
Venus (Belgium)
KUCA (Japan)

PROTEUS
Venus (Belgium)

Table 4.2. Critical facilities to be monitored in the long term (cont’d)96

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4. CONCLUSIONS AND RECOMMENDATIONSTechnical area

BWR

Severeaccidents andcontainment

Integral testing
PHEBUS (France)

PWR

VVER

PHWR/APHWR

ALWR

PHEBUS (France)

PHEBUS (France)

PHEBUS (France)

PHEBUS (France)

QUENCH1(Germany)
VERDON (France)
KROTOS (France)

QUENCH1(Germany)
VERDON (France)
KROTOS (France)

Fuel Channel
Safety Facility(Canada)MFMI (Canada)
VERDON (France)
KROTOS (France)

QUENCH1(Germany)
VERDON (France)
KROTOS (France)

MCCI(United States)
VULCANO (France)
THAI1 (Germany)
KROTOS (France)
TROI (Korea)

MCCI(United States)
VULCANO (France)
THAI1 (Germany)
KROTOS (France)
TROI (Korea)

MCCI(United States)
VULCANO (France)
THAI1 (Germany)
KROTOS (France)
TROI (Korea)

MCCI(United States)
VULCANO (France)
THAI1 (Germany)
KROTOS (France)
TROI (Korea)

PANDA1(Switzerland)
LSCF (Canada)
THAI1 (Germany)
MISTRA (France)

PANDA1(Switzerland)
LSCF (Canada)
THAI1 (Germany)
MISTRA (France)

PANDA1(Switzerland)
LSCF (Canada)
THAI1 (Germany)
MISTRA (France)

In-vessel phenomena
QUENCH1(Germany)
VERDON (France)
KROTOS (France)

Ex-vessel phenomena
MCCI(United States)
VULCANO (France)
THAI1 (Germany)
KROTOS (France)
TROI (Korea)

Containment mixing/combustion
PANDA1(Switzerland)
LSCF (Canada)
THAI1 (Germany)
MISTRA (France)

PANDA1(Switzerland)
LSCF (Canada)
THAI1 (Germany)
MISTRA (France)

Accident management
Uses data generated in resolution of other issues. No unique facility needs.
Integrity ofequipment andstructures

Halden (Norway)
JMTR (Japan)
LVR-15(Czech Republic)
ATR (United States)

Halden (Norway)
JMTR (Japan)
LVR-15(Czech Republic)
ATR (United States)

Halden (Norway)
JMTR (Japan)
LVR-15(Czech Republic)
ATR (United States)

Halden (Norway)
NRU (Canada)
LVR-15(Czech Republic)
ATR (United States)

Halden (Norway)
JMTR (Japan)
LVR-15(Czech Republic)
ATR (United States)

1. Assumes actions will be taken in the short term to preserve these facilities.
Note: BWR: boiling water reactor; PWR: pressurised water reactor; VVER: water-water energetic reactor; PHWR: pressurised heavy water
reactor; APHWR: advanced pressurised heavy water reactor, ALWR: advanced light water reactor.

4.4. General conclusions and recommendations
The following conclusions and recommendations pertain to both the short term and long term.
They result from the group’s observations and experience in carrying out the current activity anddesire to develop a practical set of recommendations with facility preservation being a
co-ordinated effort among the NEA standing committees. Specific general conclusions andrecommendations are listed below. It is worthwhile to note that many of the conclusions reachedfrom the previous SESAR/SFEAR activities are still valid, and are included in the list below.
•

Recommendation: CSNI members are encouraged to continue their excellent support for
facilities at risk, which has already resulted in several valuable projects for current
facilities at risk or to be monitored in the long term (e.g. PKL, PANDA, GALAXIE).

•

Recommendation: As recommended in the previous report, test reactor availability should
be given special scrutiny, due to the high cost of operation and replacement. The new
FIDES network and associated JEEPs is an essential step in maintaining global capability.

•

Recommendation: Regardless of FIDES and the success of JEEP, continued and ongoing
attention must be focused on smaller unique facilities at risk.

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97

4. CONCLUSIONS AND RECOMMENDATIONS•

Recommendation: The Nuclear Science Committee should maintain a close watch on
facilities used to support criticality and reactor physics codes.

The shutdown of Halden and subsequent project activities have highlighted the need forpreserving key experiments in international databases. Consequently, the group developed aseries of recommendations specifically targeting data preservation.
•

Recommendation: NEA joint safety research projects should clearly outline their planfor data preservation, and should stipulate that a copy of the primary data needs to besent to the NEA for storage.

•

Recommendation: CSNI working groups should be asked to identify key datasets in theirareas. Some of this may have been done with code validation matrices and datasets to
support the development and implementation of standards.

•

Recommendation: There should be a cross-functional (CSNI, NSC, etc.) NEA task group
established to consider what should be done to preserve the key experimental datasets.
This could include possible options for data libraries, how to screen datasets, what
information needs to accompany the primary data, etc.

•

Recommendation: CSNI working groups should select an appropriate option for preserving
each key dataset and develop an activity to put it in place (CAPS, joint project, etc.).

Reference
HySafe (2021), List of Experimental Facilities, www.hysafe.org/facilities (accessed 12 July 2021).

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Nuclear Safety Research Support

Facilities for Existing and Advanced
Reactors: 2021 Update

Experimental facilities in nuclear energy are key to addressing safety issues. The recent loss of some criticalinfrastructure, from facilities to industry expertise, has therefore become a concern for many countries. In
response, the Nuclear Energy Agency (NEA) has launched several efforts to address the matter as outlinedin this report. Current safety issues, research needs and research facilities associated with currently
operating water-cooled reactors in NEA countries are all addressed. Also included is an assessment of
the present needs to maintain experimental databases. The Senior Group of Experts on Nuclear Safety
Research, which produced this update of the 2007 report on the same issue, noted the success ofprevious reviews in helping maintain critical infrastructure and make a number of recommendations topreserve key research facilities and capabilities.

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NEA No. 7565