Nuclear Safety
NEA/CSNI/R(2007)6

ISBN 978-92-64-99005-0

Nuclear Safety Research
in OECD Countries

Support Facilities for Existingand Advanced Reactors (SFEAR)© OECD 2007
NEA No. 6158

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 30 democracies work together to address the economic, socialand environmental challenges of globalisation. The OECD is also at the forefront of efforts to understand and to help
governments respond to new developments and concerns, such as corporate governance, the information economy and thechallenges of an ageing population. The Organisation provides a setting where governments can compare policy experiences, seek
answers to common problems, identify good practice and work to co-ordinate domestic and international policies.
The OECD member countries are: Australia, Austria, Belgium, Canada, the Czech Republic, Denmark, Finland, France,
Germany, Greece, Hungary, Iceland, Ireland, Italy, Japan, Korea, Luxembourg, Mexico, the Netherlands, New Zealand, Norway,
Poland, Portugal, the Slovak Republic, Spain, Sweden, Switzerland, Turkey, the United Kingdom and the United States. The
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***This work is published on the responsibility of the Secretary-General of the OECD. The opinions expressed andarguments employed herein do not necessarily reflect the official views of the Organisation or of the governments of its member
countries.

NUCLEAR ENERGY AGENCYThe OECD Nuclear Energy Agency (NEA) was established on 1st February 1958 under the name of the OEEC European
Nuclear Energy Agency. It received its present designation on 20th April 1972, when Japan became its first non-European full
member. NEA membership today consists of 28 OECD member countries: Australia, Austria, Belgium, Canada, the Czech
Republic, Denmark, Finland, France, Germany, Greece, Hungary, Iceland, Ireland, Italy, Japan, Luxembourg, Mexico, theNetherlands, Norway, Portugal, Republic of Korea, the Slovak Republic, Spain, Sweden, Switzerland, Turkey, the United
Kingdom and the United States. The Commission of the European Communities also takes 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, thescientific, technological and legal bases required for a safe, environmentally friendly and economical use of nuclear
energy for peaceful purposes, as well as
to provide authoritative assessments and to forge common understandings on key issues, as input to government
decisions on nuclear energy policy and to broader OECD policy analyses in areas such as energy and sustainable
development.

Specific areas of competence of the NEA include safety and regulation of nuclear activities, radioactive waste
management, radiological protection, nuclear science, economic and technical analyses of the nuclear fuel cycle, nuclear law andliability, and public information. The NEA Data Bank provides nuclear data and computer program services for participating
countries.
In these and related tasks, the NEA works in close collaboration with the International Atomic Energy Agency in Vienna,
with which it has a Co-operation Agreement, as well as with other international organisations in the nuclear field.
© OECD 2007
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Cover credits: PHEBUS, CEA Cadarache, MISTRA, CEA and QUENCH, Forschungszentrum Karlsruhe GmbH.

FOREWORD

A Senior Group of Experts on Nuclear Safety Research (SESAR) was established by the NEA
Committee on the Safety of Nuclear Installations (CSNI) to assess the need for and strategy ofmaintaining key research facilities. This activity is a follow-on to a similar activity conducted by the
CSNI in the late 1990s which led to a number of actions by the CSNI to establish co-operative
research projects directed at developing information relevant to safety issues on operating water
reactors, while at the same time preserving key facilities and programmes. A report on this activity
was issued by the NEA in 2001 entitled Nuclear Safety Research in OECD Countries: Major
Facilities and Programmes at Risk (SESAR/FAP). In response to the recommendations expressed inthat report, the CSNI has undertaken initiatives, notably in the thermal-hydraulics and severe accident
areas. These initiatives mainly consisted of initiating and carrying out internationally funded OECD
projects on relevant safety issues, centred on the capabilities of key facilities identified in the SESARFAP report. These projects are ongoing and constitute a means for effectively maintaining basic
technical infrastructure through international co-operation.

3

TABLE OF CONTENTS

Foreword..........................................................................................................................................

3

Executive summary .........................................................................................................................

7

1.

Introduction .............................................................................................................................
1.1 Purpose.............................................................................................................................
1.2 Scope................................................................................................................................
1.3 Approach..........................................................................................................................
1.4 Co-ordination ...................................................................................................................
1.5 Organisation of report ......................................................................................................

15
16
16
17
18
18

2.

Overview of reactor designs and safety issues ........................................................................
2.1 Boiling water reactors ......................................................................................................
2.2 Pressurised water reactors ................................................................................................
2.3 VVER reactors .................................................................................................................
2.4 Pressurised heavy water reactors (PHWRs) and advanced PHWRs................................
2.5 Advanced LWRs (ALWRs) .............................................................................................
2.6 High-temperature gas-cooled reactors (HTGRs) .............................................................

21
21
24
26
28
30
31

3.

Technical issues and associated facilities................................................................................
3.1 Issues and facilities unique to the nuclear industry..........................................................
3.2 Issues and facilities not unique to the nuclear industry....................................................
3.3 HTGR unique safety issues and research needs...............................................................

35
36
78
87

4.

Conclusions and recommendations .........................................................................................
4.1 Introduction .....................................................................................................................
4.2 General conclusions and recommendations .....................................................................
4.3 Short-term conclusions and recommendations ................................................................
4.4 Longer-term conclusions and recommendations..............................................................

99
99
99
100
102

5

EXECUTIVE SUMMARY

At its December 2002 meeting, the CSNI approved the establishment of a senior group of experts
on nuclear safety research (SESAR) to assess the need for and strategy of maintaining key researchfacilities. This activity is a follow-on to a similar activity conducted by the CSNI in the late 1990s
which led to a number of actions by the CSNI to establish co-operative research projects directed atdeveloping information relevant to safety issues on operating LWRs and PHWRs, while at the sametime preserving key facilities and programmes. A report on this activity was issued by the NEA in
2001 entitled: Nuclear Safety Research in OECD Countries: Major Facilities and Programmes at Risk
(SESAR/FAP). In response to the recommendations expressed in that report, the CSNI has undertaken
initiatives, notably in the thermal-hydraulics and severe accident areas. These initiatives mainly
consisted of initiating and carrying out internationally funded OECD co-operative projects on relevant
safety issues, centred on the capabilities of key facilities identified in the SESAR/FAP report. Foursuch projects (involving five facilities) were initiated and are ongoing, constituting a means for
effectively maintaining basic technical infrastructure through international co-operation.
Since publication of the SESAR/FAP report, research facilities have continued to be shut downworldwide. In fact, of the facilities listed in the SESAR/FAP report in the areas of thermal-hydraulics,
fuel, reactor physics, severe accidents and integrity of equipment and structures (i.e. those areas mostunique to the nuclear power industry), approximately 35% have been shut down in the past five years.
Accordingly, loss of critical research infrastructure (i.e. facilities, capabilities and expertise) remains a
concern and is a major factor in conducting the current study. However, it should be recognised that
the SESAR/FAP effort led to CSNI actions that preserved five key facilities during the 2000-2006
time period. These are discussed in Chapter 4.
The activity described in this report builds upon and updates the SESAR/FAP work, but also
expands its scope to cover advanced LWRs (ALWRs), VVERs, advanced PHWRs (APHWRs) andhigh-temperature gas-cooled reactors (HTGRs). Accordingly, the title of this activity is SESAR
Support Facilities for Existing and Advanced Reactors (SESAR/SFEAR). The need to maintaindatabases of experimental data is also recognised as an important issue, but is not treated in this report,
since preservation of data is being addressed separately by the NEA.
The focus of this report is on the safety issues, research needs and supporting research facilities
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 (PHWRs) and Russian-designed VVERs. For these reactors, the main purpose of this report isto:xSummarise the currently identified safety issues, whose resolution depends upon additional
research work.

x

Provide the current status of those research facilities unique to the nuclear industry that
support resolution of the safety issues.

7x

Where such facilities represent a substantial investment of resources and are in danger ofpremature closure, recommend actions the CSNI could take in the short term to help
maintain them.

x

Provide recommendations on long-term nuclear safety research facility infrastructure needs
and preservation.

In addition, where research facilities do not exist, but may be useful to address currently
identified safety issues, these areas are identified. The report also provides information on safetyissues and research needs not unique to the nuclear industry and on safety issues and research needs
associated with HTGRs. This information is presented for completeness and for use by designers,
operators and researchers in planning and conducting future work.
The issues addressed in this report are those associated with nuclear reactor safety (excluding
spent fuel storage) and are organised into the following technical areas:a)Those unique to the nuclear industry:x Thermal-hydraulics.
x Fuel.
x Reactor physics.
x Severe accidents.
x Integrity of equipment and structures.

b)

Those not unique to the nuclear industry:x Human and organisational factors.
x Plant control and monitoring.
x Seismic effects.
x Fire assessment.

c)Those unique to HTGRs.

The report is organised into four chapters. Chapter 1 is the introduction describing the scope,
purpose and approach used in assessing the safety issues and facilities. Chapter 2 provides a shortoverview of each of the reactor types covered in this report and their associated safety issues.
Chapter 3 contains sections on each of the technical areas. Each section contains a description of thesafety issues and research facilities associated with that technical area. For areas unique to the nuclear
industry (item (a) above), information is provided in tabular form showing:xxx

The safety relevance of each issue.
The state of knowledge on each issue.
The relevance of each facility to address the safety issue.

This information is then used to calculate a relative numerical ranking for each facility in eachtechnical area (as described in the beginning of Chapter 3) which is one factor used in developing
recommendations for CSNI consideration.
Chapter 4 then provides conclusions and recommendations for CSNI consideration, based upon
the information in Chapter 3. The conclusions and recommendations only pertain to those technicalareas unique to the nuclear industry [item (a) above].
In general, in developing recommendations for CSNI consideration, the group focused on thosefacilities that have unique capabilities, would be very expensive to replace and have high relevance tothe resolution of current safety issues (as indicated by their relative numerical ranking), as well as thepotential to be highly relevant in support of the resolution of ALWR and APHWR safety issues.
8Accordingly, such facilities represent an infrastructure of substantial resource investment and, if lost, it
is unlikely that such facilities would be replaced due to the reality of cut backs in nuclear safety
research funding over the past few years. Due to the cost of operating such facilities, co-operative
efforts would most likely be needed to maintain them in the longer term.
The conclusions and recommendations in Chapter 4 are organised into general, short-term andlong-term items as follows:(a) The following are the general conclusions and recommendations (i.e. they pertain to both theshort term and long term).
x

CSNI efforts aimed at facility preservation should focus on large facilities, whose loss
would mean the loss of unique capability as well as the loss of substantial investmentthat in the current climate of tight resources would not likely be replaced. Such
preservation also includes maintaining the expertise, knowledge, capabilities andpersonnel essential to infrastructure preservation. In this regard, it should be noted that
due to previous CSNI efforts, several large facilities (i.e. PANDA, PKL, MACE,
ROSA) have been kept active over the past five years, thus helping the current SFEAR
effort. However, many large, expensive and unique facilities are projected to close overthe next 1-5 years. Examples include thermal-hydraulic and severe accident facilities. In
addition, many of the test reactors are old and will reach their end of life without
substantial refurbishment. The loss of such facilities would severely detract from thenuclear safety research infrastructure. Additional discussion on a strategy for long-term
facility preservation is discussed in item c) below.

x

The NEA Nuclear Science Committee (NSC) should take the lead to monitor the statusof and make recommendations for actions to preserve key facilities in the reactor
physics area. The facilities and information in this report in the reactor physics area
represent the SESAR group’s views on the safety issues and facilities important tonuclear safety research and are for NSC use in carrying out this responsibility.

x

To help stimulate industry interest in facility and infrastructure preservation, it isrecommended that both the CSNI and the CNRA take steps to encourage industry cooperation by emphasizing: 1) the responsibility of industry to develop sufficient data to
support their applications, 2) the benefits of co-operative research, and 3) the value ofpreserving critical research infrastructure.

x

Hot cells and autoclaves are essential to nuclear safety research. However, due to thelarge number of hot cells and autoclaves, it is impractical for the CSNI to monitor theirstatus. Accordingly, each country should monitor the status of these facilities and bringto the CSNI’s attention any concerns regarding loss of critical infrastructure.

x

Certain safety issues have no large-scale facilities identified for the conduct of relevant
research. The appropriate CSNI working groups should evaluate whether or not largescale facilities are needed to support resolution of these issues. The issues that fall inthis category are:––––

ECCS strainer clogging (thermal-hydraulic issue #6).
3-D core flow distribution (thermal-hydraulic issue #12).
Long-term behaviour of concrete structures (structural integrity issue #7).
Flow-induced vibrations (structural integrity issue #9).

9

(b) The following recommendations are directed toward those actions that the CSNI could takein the short term (2006-2007) to prevent the loss of key facilities in imminent danger ofclosure.
x

In the thermal-hydraulics area, both existing large integral BWR thermal-hydraulic test
facilities (PANDA and PUMA) are in danger of being closed in the next 1-2 years.
These facilities are unique and expensive, and at least one should be maintained to beavailable for supporting research related to current or future BWR safety issues.
Accordingly, preservation of one integral BWR thermal-hydraulic test facility (either
PANDA or PUMA) is considered essential for preserving a BWR thermal-hydraulicresearch infrastructure. SESAR is of the view that PANDA is the preferred facility forpreservation due to its scale, replacement cost and versatility (i.e. it is useful in thesevere accident as well as the thermal-hydraulics area). Accordingly, CSNI action isrecommended in the short term to support a co-operative research programme in
PANDA. It should be noted that CSNI actions resulting from the SESAR/FAP report
played a major role in the preservation of PANDA over the past five years.

x

In the severe accident area, most facilities supporting the resolution of the following
safety issues for BWRs, PWRs, VVERs and ALWRs are in danger in the short term:
–
–
–

x

Pre-core melt conditions.
Combustible gas control.
Coolability of over-heated cores.

Based upon a review of the facilities in short-term danger, their importance to theresolution of the above safety issues and long-term infrastructure preservation, thegroup concluded that the following should be preserved due to their replacement cost,high relative ranking and versatility:–––

PHEBUS.
QUENCH.
MISTRA.

Each of these is discussed further below.
x

PHEBUS is a unique facility representing a substantial financial investment, capabilities
and expertise. Due to the high cost of its operation and the long time frame necessary toplan and conduct experiments, it is not considered practical to propose that the CSNI
organise a co-operative research programme in PHEBUS. Accordingly, it is recommended that PHEBUS be treated as a special case, with the French authorities taking thelead to propose and organise a future research programme using PHEBUS. In this regard
it should be noted that a PHEBUS Expert Group has been organised to assess future
experimental programmes in PHEBUS in the areas of LOCA (fuel response to LOCAs)
and of severe accidents (fuel degradation and fission product release and transport). Bothseparate affects and integral tests are included in the assessment. The recommendations
from this group should provide valuable input for justifying and planning future
programmes in PHEBUS.

x

The QUENCH facility has been used extensively in the past to investigate pre-core melt
conditions in LWRs. Although it is a unique facility in near-term danger, the group has
concluded that any effort to preserve it for the long term should be dependent upon
identifying a future experimental programme that can provide useful information

10

beyond what has already been done in QUENCH. Accordingly, it is recommended that
QUENCH be treated as a special case, with the German authorities taking the lead topropose a future research programme in QUENCH capable of generating useful new
information. In this regard, it is also recommended that the appropriate CSNI Working
Group (WGAMA) be requested to consider future uses for QUENCH and provide a
recommendation to the CSNI that can be factored into deliberations on the future of
QUENCH.
x

The MISTRA facility has the capability for conducting experiments on combustible gas
mixing and transport in multi-compartmental configurations with detailed instrumentationand helium as a simulant for hydrogen gas (H2). As such, it can measure 3-D effects
useful for assessing 3-D analytical tools. MISTRA complements the THAI facility whichuses H2 and can conduct experiments on H2 combustion and aerosol distribution. THAI is
not in near-term danger and is recommended for long-term preservation (see below).
Accordingly, it is recommended that the CSNI take action to reserve the MISTRA facility(so as to maintain the complementary infrastructure and expertise) by organising andconducting a co-operative research programme in MISTRA.

x

In the other technical areas (fuels, and integrity of equipment and structures), no
short-term CSNI actions are recommended.

x

It should be recognised that implementation of the above recommendations are
dependent upon interest and commitment of the “host countries” to provide sufficient
resources to attract participation of other interested parties and the ability to propose
experimental programmes relevant to resolution of the issues and of interest to member
countries.

(c) In the longer term (beyond 2007), it is recommended that the CSNI adopt a strategy for thepreservation of a research facility infrastructure, based upon preserving unique, versatile andhard-to-replace facilities. The number and nature of these facilities should be based upon
supporting currently operating LWRs and PHWRs and the licensing of future ALWRs and
APHWRs. The strategy should include consideration of short- and long-term priorities, costof preservation (e.g. would the cost of preservation detract substantially from otherprogrammes/facilities) and contingency plans in case of facility loss.
In this regard, many of the factors used in the report to arrive at conclusions andrecommendations could be useful in developing a long-term strategy for assessing and initiating future
co-operative research projects.
These factors include:
x
x
x
x
x

Facility operating and replacement cost,
The ability to define a useful experimental programme,Long-term resource implication and priorities,
Industry participation,
Host country long-term plans and commitment.

Additional discussion on each of these factors is provided in Chapter 4.
In addition, critical research capabilities and expertise are defined qualitatively in the
OECD/NEA Collective Statement Concerning Nuclear Safety Research (2004). Using this statement
and the safety issues contained in Section 3.1 of this publication, a table of critical research facility
infrastructure needs was developed along with a list, by reactor type, of existing facilities that could11fulfil those needs. These are shown herein as Tables ES-1 and ES-2, respectively. The facilities listed
in Table ES-2 are those considered unique, hard-to-replace and identified as having high relative
importance in their technical area, as discussed in Chapter 3 of this report. Accordingly, based upon
Tables ES-1 and ES-2, it is recommended that the CSNI focus on these facilities in developing astrategy for long-term infrastructure preservation. The CSNI should monitor the status of thesefacilities in the longer term with a 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, forinvestigating safety issues associated with new reactors and technologies, the CSNI should take anactive role in encouraging and organising co-operative research efforts. This will also contribute to
infrastructure preservation. Similar to the short-term recommendations above, host country interest
will be an important factor in determining which facilities to preserve.
Table ES-1. Critical research facility infrastructure
Technical expertise needed
Thermal-hydraulics:
modelling and analysis.

Facility capability needs

Important factors for facilities

Large-scale integral test facilities foreach reactor type.

Scale, temperature, pressure and instrumentation capability the key factors.
Also, the completeness of the facility
with respect to factors such as: auxiliary
systems, number of loops and instrumentation capability are important.

Fuels:
Test reactor for steady state andperformance and phenomena. reactivity insertion testing.
Hot cells for PIE and simulated
LOCA testing.

Ability to achieve representativevalues of energy deposition in transient
tests. Ability to achieve linear heatrating, burn-up and adequate in-core
instrumentation for steady state testing.
Ability to do experiments with MOX
and high burn-up fuel. Hot cells forfull length and pin segment PIE.

Critical facilities for measuringphysics parameters and performing
benchmark experiments.

Ability to do experiments with MOX
and high burn-up fuel.

Severe accidents: phenomena, Facilities for testing:accident progression,x Integral SA phenomena.
modelling and analysis.
x In-vessel phenomena.
x Ex-vessel phenomena.
x Containment atmospheremixing/combustion.
x Accident management strategies.

Use of prototypic materials and large
scale are important.

Reactor physics:
modelling, cross-sections,parameters and analysis.

Integrity of equipment andstructures: materials
behaviour, structural design.

Test reactors for irradiating material Ability to achieve fluence and othersamples under controlled conditions. prototypic conditions (e.g. temperature,
simulate impurities, stress, etc.). Hot
Hot cells for examining large andcells and autoclaves for ex-reactor testingsmall irradiated material samples.
of irradiated materials.
Autoclaves for materials testing.

12

Table ES-2. Critical facilities to be monitored in the long term
Technical area

BWR
1

PWR

VVER

Thermalhydraulics

PANDA

LSTF/ROSA
PKL
ATLAS

PSB-VVERPACTEL

Fuels2

HaldenNSRR
PHEBUS

HaldenNSRRCABRIPHEBUS
Proteus
Venus

HaldenMIRCABRIPHEBUS
Proteus
LR-O

Reactor physics* Proteus
Venus

Severe accidents Integral testing
PHEBUS1
PHEBUS1
In-vessel phenomena– QUENCH1– QUENCH1– VERDON– VERDON– KROTOS– KROTOS

Integrity ofequipment andstructures3
*

PHWR/
APHWR
RD-14-M

HaldenNRU

Proteus
ZED-2
Venus

ALWR
LSTF/ROSAPKLPANDA1ATLAS
HaldenNSRRCABRIPHEBUS
ProteusVenus

PHEBUS1

PHEBUS1

PHEBUS1– QUENCH1– VERDON– KROTOS– Fuel Channel
Safety Facility– VERDON– KROTOS– QUENCH1– VERDON– KROTOS

Ex-vessel phenomena– MCCI– MCCI– MCCI– MCCI– MCCI– VULCANO – VULCANO – VULCANO – VULCANO– VULCANO– THAI2– THAI2– THAI2– THAI2– THAI2
44
44– KROTOS– KROTOS– KROTOS– KROTOS– KROTOS4
Containment mixing/combustion– PANDA1– PANDA1– PANDA1– PANDA1– PANDA1– LSCF– LSCF– LSCF– LSCF– LSCF– THAI2– THAI2– THAI2– THAI2– THAI2
1
1
1
1– MISTRA– MISTRA– MISTRA– MISTRA– MISTRA1
Accident management
Uses data generated in the resolution of other severe accident and thermal-hydraulic issues.
No unique facility needs.
HaldenHaldenHaldenHaldenHalden
JMTR
NRU
JMTR
JMTR
JMTR
LVR-15
LVR-15
LVR-15
LVR-15
LVR-15ATR
ATR
ATR(Included for completeness; NSC to monitor status).

Notes:
1. Assumes actions will be taken in the short term to preserve these facilities.
2. Assumes ongoing effort to initiate a co-operative research programme will be successful.
3. Due to the large number of hot cells and autoclaves, each country should monitor the status and identify
concerns.
4. Experimental programme under discussion in the CSNI/SERENA programme.

13Chapter 1
INTRODUCTION

At its December 2002 meeting, the CSNI approved the establishment of a senior group of experts
on nuclear safety research (SESAR) to assess the need for and strategy of maintaining key researchfacilities. This activity is a follow-on to a similar activity conducted by CSNI in the late 1990s andwhich led to a number of actions by CSNI to establish co-operative research projects directed atdeveloping information relevant to safety issues on operating LWRs and PHWRs, while at the sametime preserving key facilities and programmes. A report on this activity was issued by NEA in 2000
titled Senior Group of Experts for Nuclear Safety Research: Facilities and Programmes(SESAR/FAP).
The activity described in this report builds upon and updates the SESAR/FAP work, but also
expands its scope to cover advanced LWRs (ALWRs), VVERs and high-temperature gas-cooled
reactors (HTGRs). Accordingly, the title of this activity is SESAR: Support Facilities for Existing andAdvanced Reactors (SESAR/SFEAR). The need to maintain data bases of experimental data is also
recognised as an important issue, but is not treated in this report, since preservation of data is being
addressed separately by the NEA. The SESAR/SFEAR work can be considered a follow-on to the2000 NEA SESAR/FAP report, which was highly successful in stimulating co-operative research
projects as well the preservation of important facilities.
Specifically, SESAR/FAP led to a number of recommendations, many of which resulted in CSNI
organising and initiating co-operative research projects to preserve key facilities that were in near-term
danger and to improve the state of knowledge in selected areas. The actions taken by CSNI as a resultof the SESAR/FAP recommendations have also impacted the current SFEAR study by keeping active
a number of facilities that would otherwise have been lost or in need of short-term CSNI action. A
summary of the SESAR/FAP recommendations related to facilities, the resulting CSNI action and theimpact on the SFEAR report is contained in Table 1-1.
However, despite the efforts of CSNI, the overall situation with respect to loss of researchfacilities and infrastructure has continued to deteriorate since issuance of the SESAR/FAP report in
2002. In the six years since the SESAR/FAP report was issued, approximately 35% of the researchfacilities listed therein have been closed. While it is appropriate that some facilities and programmes
be closed, such a closure rate points to the need to monitor the situation closely so as to be able to takeaction if key facilities are in danger. Accordingly, the SFEAR effort represents a look at the current
situation, including the previous SESAR/FAP recommendations and resulting actions, and makes
recommendations related to short-term and long-term actions which should be taken to preserve thekey safety research facility infrastructure in member countries and Russia. The scope and approach
used in the SFEAR programme are discussed below.

151.1 Purpose
For operating reactors, the purpose of the SESAR/SFEAR activity is to (1) summarise thecurrently identified safety issues, whose resolution depends upon additional research work, 2) provide
the current status of those research facilities unique to the nuclear industry that support resolution of
the safety issues, 3) where such facilities represent a substantial investment of resources and are indanger of premature closure, recommend actions CSNI could take to help maintain them and,4) provide recommendations on long-term nuclear safety research facility infrastructure needs and astrategy for its preservation. In addition, where research facilities do not exist, but would be useful toaddress currently identified safety issues, facility needs are identified. For future reactors, the purposeof the SESAR/SFEAR activity is to identify safety issues, safety research and facility needs for use by
designers, operators and researchers.
This activity is considered vital to achieving one of the main goals for CSNI, as described in its
operating plan, which is to help maintain the critical infrastructure and expertise to ensure thecontinued safety of nuclear power. Maintaining such infrastructure and expertise is necessary tounderstand and address safety issues and ensure the safe operation of existing and future NPPs. The
existence of interesting international research projects is also a factor in being able to attract and train
new people into the nuclear safety field. This is an incentive for countries to serve as the host country
for co-operative research projects involving their facilities.
Finally, in 2004 the NEA issued a publication entitled Collective Statement Concerning Nuclear
Safety Research. This publication emphasised the role research plays in support of the efficient andeffective regulation of nuclear power plants. In this publication, research capabilities and expertise
which should be maintained as part of the overall regulatory infrastructure were identified.
Accordingly, the SESAR/SFEAR work is also directed toward helping to maintain this infrastructure.
1.2 Scope
The scope of this activity is limited to safety issues and facilities associated with nuclear reactor
design, construction and operation (spent fuel storage is not included in the scope of this activity) inOECD member countries and Russia and covers the following technical areas:
x
x
x
x
x
x
x
x
x
x

Thermal-hydraulics.
Fuel.
Reactor physics.
Severe accidents.
Integrity of equipment and structures.
Human and organisational factors.
Plant control and monitoring.
Seismic effects.
Fire assessment.
High temperature gas reactor (HTGRs) safety issues.

Of the technical areas, the first five address phenomena, safety issues and facilities unique to the
nuclear industry. The next four technical areas address phenomena, safety issues and facilities that are
relevant to the nuclear industry but are also relevant to other industries and, thus, research andfacilities may be supported by others. Finally, the last technical area (HTGR safety issues) is related tophenomena and safety issues for a technology that has yet to see widespread use and, thus, muchresearch remains to be done by the nuclear community to support its deployment. It should also be
noted that the technical areas contained in this report are the same as those contained in the SESAR:FAP report, except for HTGR issues, which have been added; and risk assessment, which has been16

deleted since there are no facilities associated with this area. 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:x

Currently operating PWRs, BWRs, VVERs and PHWRs in member countries.

x

Future designs that could be deployed in the next 5-10 years. These future designs include
advanced LWRs, PHWRs, and HTGRs.

Due to their long-term schedule and preliminary nature of their designs, it was decided not tocover designs where deployment is beyond 10 years (e.g. Generation IV) in this report.
It should also be recognised that research in some of the technical areas listed above does not
require large facilities, but rather large-scale coordinated programmes.
1.3 Approach
In each of the technical areas above, the SESAR/SFEAR participants developed two sets ofrecommendations: (1) those for the technical areas unique to the nuclear industry that are applicable tocurrently operating and future LWRs and PHWRs and (2) safety research and facility needs for future
HTGRs. For those technical areas not unique to the nuclear industry, no recommendations areprovided. Rather, the safety issue and facility information is provided for information only. For item
(1) above, operating and future plants, the members identified the safety issues still needing research
work and the major facilities currently performing or available to perform research directly relevant tothese issues. For each of the five technical areas unique to the nuclear industry, the following
information related to each safety issue and facility was generated:x

Safety issues:–
Safety relevance.
–
State of knowledge.
–
Applicability (to which reactor types does the safety issue apply).
–
Importance of relevant facilities to resolve the issue.

xFacilities:–
Applicability (to which reactor types does it apply).
–
Annual operational cost.
–
Replacement cost.
–
Capability and versatility of facility (including quality of instrumentation).
–
Planned duration of facility operation.

This information is displayed in the form of two tables for each technical area, as illustrated at theend of this chapter (Tables 1-2 and 1-3). Also shown are the definitions for the terms used in thetables.
To ensure that the SFEAR activity focused on only those major facilities large enough to warrant
CSNI consideration (i.e., those that would need a co-operative effort to finance), the following
guidelines were established regarding facility size considered in this study:x

To be considered in the SESAR/SFEAR activity, the facility would need to have at least anannual operating cost of >P$1 000 000 per year or at least a replacement value of>$ 2 000 000.

17

In addition, for a viable co-operative research project, there must also be a host country
commitment for a programme and resources consistent with CSNI guidelines.
Based upon the safety issue and facility information, the group then developed a relative ranking foreach of the facilities identified as relevant to one or more of the safety issues in those technical areas
unique to the nuclear industry. The relative ranking process is described in the introduction to Chapter 3.
Based upon this relative ranking, the importance, versatility and critical nature of the facility the groupdeveloped recommendations for CSNI consideration. These are described in Chapter 4.
For item (2) above, HTGRs, the group identified the safety issues, and then developed high levelrecommendations regarding the safety research needs and facility needs that should be considered toaddress the issues. The group’s recommendations are not for CSNI action, but rather for HTGR plantdesigners and regulators to consider in planning future programmes.
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 in their organisations.
However, in assembling the information contained in the report the participants had the benefit of
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 and otherexternal organisations (CNRA, IAEA and others) were offered an opportunity to comment on a draftof the report.
1.5 Organisation of report
The report has been organised to provide first, a short overview (Chapter 2) of the reactor designs
and their safety issues addressed within the scope of this report, followed by a chapter (Chapter 3)
addressing each of the technical areas listed above. Chapter 3 is organised into three sections pertaining
to those technical areas unique to the nuclear industry, those not unique and HTGR unique issues. The
introduction to Chapter 3 describes in more detail the organisation and purpose of each of these three
sections. Chapter 4 then provides the group’s conclusions and recommendations regarding criticalfacilities unique to nuclear safety research in danger of being lost that deserve and need international
support and possible actions for CSNI consideration.
The conclusion and recommendations are based upon several factors. These include:x

The importance of the facility to resolving the safety issues, based upon the relative ranking
of a facility within a given technical area using the ranking approach described in the
introduction to Chapter 3.

x

The versatility of the facility, including quality of instrumentation.

x

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

The conclusions and recommendations are divided into near term (CSNI action needed in thenext 1-2 years) and long term (CSNI should monitor the status).
However, any action by CSNI would be contingent 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., CSNI does not intend to serve as host country for facility preservation).

18

Table 1-1. Impact of SESAR/FAP facility recommendations on SESAR/SFEARSESAR/FAP recommendation

Resulting CSNI action

Impact on SFEAR

x PANDA maintained through2005. Currently in near-term
danger and addressed in the
SFEAR study.
x PKL active and not in near-term
danger.
Facility status monitored. Initiated ROSA is active and not in2) Monitor and maintain key
thermal-hydraulic facilities in the programme utilising the ROSA
near-term danger. Other T/Hfacility when it was in danger oflong term. T/H facilities should
facilities continue to be monitored.
be maintained in North America, being shut down.
Europe and Asia.
3) Maintain the RASPLAV andx MASCA is active and not inx Initiated the MASCA
MACE facilities in the severenear-term danger.
programme as a follow-on toaccident area (these facilities
RASPLAV to maintainx MCCI is active and therefore thewere in near-term danger offacilities.
MACE facility is not inclosure.
near-term danger.
x Initiated the MCCI programme
utilising the MACE facility.
Initiated the SERENA programme SERENA programme has
4) Develop centre of excellenceon fuel-coolant interaction (FCI) (group of experts to discuss status recommended an experimental
programme be conducted inin consideration of potential loss of FCI and future experimental
KROTOS and thus may impact
needs). FARO shutdown.
of the FARO and KROTOS
preservation of KROTOS facility.
KROTOS kept in standby.
facilities.
CSNI expert group to review
SERENA recommendation.
5) Develop centre of excellence
Proposal for COE currently under No additional CSNI action needed.
(COE) on iodine chemistry andevaluation.
fission product behaviour.
1) Maintain the PANDA, PKL and
SPES facilities in the thermalhydraulic area (these facilities
were in near-term danger ofclosure).

Initiated the SETH programme
utilising the PANDA and PKL
facilities (no host country support
for SPES).

Table 1-2. Issues versus facilities
Issue

Facility
Applicability Safety relevance State of knowledge
Importance of facility toof issueof issueon issue
Name
Versability
resolution of the issue?

Table 1-3. Facilities
Facility Applicability Cost/Yr. * Replacement* Issues
Planned duration
Capability
name (type of reactor) operationcost
coveredof operation

Notes:– High, Medium, Low:
– Operational Cost = Low is < $1.0 million/yr.; Medium is $1-2 million/yr.; High is > $2 million/yr.
– Replacement Cost = Low is < $2.0 million; Medium is $2-10 million; High is > $10 million

19

RelativerankingDefinitions for terms used in tables
A)

Issues versus facilities table

Applicability of issue
To what types of reactor designs does the issue apply?
Safety relevance of issue
HIGH

is directly related to establishing the safety parameters affecting the successful
performance of critical safety functions (e.g., reactor shutdown, decay heat removal,
containment of fission products, etc.), ensuring key assumptions in the safety analysis arevalid and/or determining the scope and significance of potential new issues.

MEDIUMis related to reducing uncertainties regarding known safety issues to support realistic
analysis and issue resolution.

LOW

is not related to critical safety functions or issues.

State of knowledge on issue
HIGH

no additional information beyond that obtained from existing research programmes isrequired to resolve the issue.

MEDIUM

additional information beyond that from existing research programmes is useful to
reduce uncertainties and support realistic issue resolution, but the issue can be resolved
conservatively without such additional information.

LOW

additional information is necessary to understand the phenomena and/or scope of theissue to allow its resolution.

Importance of facility to resolution of the issue
HIGH

Can directly contribute to issue resolution.

MEDIUMCan help reduce uncertainties.

LOW

Can get the information in other ways.

Versatility of facility
Would the facility likely be used in the future (i.e., can it be used for other issues, for future plants,etc.) with reasonable modifications?
B)

Facilities table

Capability
What types of experiments at what scale can be done?
Current planned duration of operation
What is the current status of the facility? When may the facility likely be shut down without CSNI
action?
Relative ranking
Based on process described in Chapter 3.
20

Chapter 2
OVERVIEW OF REACTOR DESIGNS AND SAFETY ISSUESThis section contains short overviews of each of the major reactor types included in the scope of
this report. These overviews are for the purpose of familiarising the reader with the basic characteristicsof the designs as well as the major safety features and issues. The reactor designs included in this section
are:
x
x
x
x
x
x

Boiling water reactors (BWRs).
Pressurised water reactors (PWRs).
VVER reactors.
Pressurised heavy water reactors (PHWRs) and advanced PHWRs.
Advanced light water reactors (ALWRs).
High-temperature gas-cooled reactors (HTGRs).

2.1 Boiling water reactors
Introduction
The direct cycle boiling water reactor (BWR) nuclear system is a steam generation and steam
utilisation system consisting of a nuclear core located inside a reactor pressure vessel and aconventional turbine-generator and feedwater supply system. Associated with the nuclear core areauxiliary systems to accommodate the operational and safety requirements and necessary controls andinstrumentation. Water is circulated through the reactor core, producing saturated steam which isseparated from the recirculating water, dried in the top of the vessel, and directed to the steam turbinegenerator. The turbine employs a conventional regenerative cycle with condenser deaeration andcondensate demineralisation.
The steam produced by the nuclear core is radioactive. The radioactivity is primarily due to -16, avery short-lived isotope (7 seconds half-life) so that the radioactivity of the steam from the reactor
vessel is mostly present during power generation. However, other radioactive material (e.g., from fuelcladding 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 upon the age of the design andproduct type. However, they all have certain common characteristics which include:x

The almost universal use of recirculation inside the reactor vessel (using either jet pumps orcentrifugal pumps) to increase water flow through the core (and thus control boiling andpower level).

x

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 thepressure suppression pool) thus allowing a smaller containment building.

21

x

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 the coreoutlet. Such a control rod location also allows removal of the upper reactor vessel head forrefueling without disturbing the control rod drives.

x

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

x

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

Described below are the main design features and safety issues associated with BWRs.
BWR design features
The nuclear core consists of fuel assemblies and control rods contained within the reactor
pressure vessel and cooled by a recirculating water system. The recirculating water system consists ofthe feedwater flow and flow internal to the reactor vessel (recirculation flow) which is the result ofpumps 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 the reactor
vessel. BWR power level is maintained or adjusted by positioning control rods up and down within thecore. The BWR core power level is further adjustable by changing the recirculation flow rate throughthe core without changing control rod position. This BWR feature helps achieve load-following
capability for the BWR.
The BWR employs bottom-entry control rods and bottom-mounted control rod drives whichallow 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 (approx. 70 bars or 1 000 psi) at the corresponding
saturation temperature. They employ moisture separators and steam dryers to enhance the quality ofthe steam entering the turbine.
The BWR reactor is housed in a reactor building structure. The reactor building structure 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, the reactor water
recirculation system, the main steam lines, main steam line safety/relief valves and discharge piping,
control rod drives and piping, piping and valves associated with reactor vessel, nuclear system
instrumentation, and heating and ventilation. Components located outside the drywell, but inside thecontainment 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 systempiping, refuelling bridge, polar crane, nuclear systems instrumentation heating and ventilating, and thepressure suppression pool.
The containment is a steel leakage barrier which prevents significant fission product release to theouter shield building in the event of an accident. The containment, including all penetrations andwelded attachments, acts as an independent structural component within the reactor building for the
maximum temperature and pressure conditions that can occur as the result of a LOCA, andaccommodates reactor blowdown through the safety/relief valve discharge piping to the suppression
pool.

22

The suppression pool is toroidal pool of demineralised water located in the wetwell drywell andwithin the containment boundary. The suppression pool provides (a) a means to condense any steam
released in the drywell area during a LOCA; (b) a heat sink for the reactor core isolation coolingsystem during hot standby operation until the decay heat can be removed by the residual heat removal(RHR) heat exchangers; (c) a heat sink for venting the nuclear system safety/relief valves; (d) a sourceof water for emergency core cooling, and (e) a source of water to the containment spray system.
Surrounding the containment is the shield building. The shield building is a cylindrical shell of
reinforced concrete. It completely encloses the reference free-standing containment. The primaryfunction 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 and external threats, suchas missiles.
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 adesign basis accident. The annulus is normally kept at a negative pressure relative to atmosphericpressure so any leakage through the shield building or containment is into this space. Under accident
conditions, the ventilation exhaust from this space is automatically diverted 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 of power
and reactivity insertion events. These consist of:
x
x
x
x
x

A fast acting SCRAM system.
A standby liquid control system for emergency boron injection into the core in the event of 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.

BWR safety issues
Design, operation and research associated with BWRs since the mid-1950s has generated
information used to address many safety issues. These issues include:
x
x
x

x
x
x
x

Anticipated transient without scram.
Recirculation pipe cracking.
Severe accident concerns:– MK-I containment shell melt through.
– MK-I containment vent.
– Combustible gas control.
– In-vessel melt retention.
– Ex-vessel core debris coolability.
– Source term.
– Accident management.
Station blackout.
Core spray distribution.
Suppression pool dynamics.
Stability.

23However, these and other issues have undergone re-evaluation due to the development of new
designs, operating experience feedback (lessons learned) or industry initiatives to extend plant
lifetime, raise power levels, increase fuel burn-up levels or increase operating cycle length. Theseissues include:
x
x
x
x
x
x
x
x
x
x

Plant ageing.
Power uprates.
High burn-up fuel.
Materials cracking/corrosion.
Installation of digital I&C.
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 andothers are shared with PWRs.
Although the basic safety issues are the same, research to address the above issues may bedifferent due to the differences in design and operation between BWRs and PWRs. Accordingly, inassessing facility and programme needs, these differences need to be considered.
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 newissues needing research, such as:xx

Redefinition of large break LOCA.
Break location and orientation.

2.2 Pressurised water reactors
Introduction
The pressurised water reactor (PWR) consists of a high pressure reactor vessel (about 140 bar or
2 200 psi) with anywhere from 2 to 4 coolant loops. Each coolant loop has a reactor coolant pump anda 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 psi.
PWR core and plant designs can vary depending upon the product type (2, 3 or 4 loop plant).
However, they all have certain common characteristics, which include:
x

x
x
x

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.
24Described below are the main design features and safety issues associated with PWRs.
PWR 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 the assemblies.
In the typical initial core loading, three fuel enrichments are generally used. Fuel assemblies with thehighest enrichments are placed in the core periphery, or outer region and the two groups of lower
enrichment fuel assemblies are arranged in a selected pattern in the central region. In subsequent
refuelings, 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 core in such amanner 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) to generate steam for theturbine generator. The primary coolant loop is closed when the water is pumped back to the reactor
vessel. The entire reactor coolant system is composed of leak tight components to ensure that allradioactivity is confined to the system.
The reactor system containment building is usually a reinforced concrete or steel shell pressure
vessel. The contained volume and design pressure of the vessel are sufficient to withstand and containthe contents of the reactor coolant system in the unlikely event of a loss-of-coolant accident 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, and the 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 thecontainment 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; by control
rod cluster motion, which is required to follow load transients and for startup and shutdown; and by asoluble 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 product poisons which determine the core
reactivity tend to slow the nuclear chain reaction
Rod cluster control (RCC) 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, whichis raised and lowered by a drive mechanism mounted on the reactor vessel head. Downward trip of the
RCC 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:
x
x
x
x
x

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.
Back-up power supplies.
25

x
x

In-containment spray systems to condense steam and scrub fission products from thecontainment atmosphere (although not all PWRs have this feature).
Containment cooling systems.

PWR safety issues
Design, operation and research associated with PWRs have generated information to address
many safety issues, which include:
x
x
x
x
x
x

Anticipated transient without scram.
Loss of coolant accidents.
Loss of feedwater transients.
Steam line breaks.
Station blackout.
Severe accident concerns:– Direct containment heating due to RPV bottom head failure.
– Accident management.
– In-vessel melt retention.
– Ex-vessel core debris coolability.
– In and ex-vessel FCI.
– Core debris criticality under reflood conditions.
– Combustible gas control.
– Source term.

However, these and other PWR safety issues have undergone reevaluation for example, in viewof industry initiatives to raise power levels, increase fuel burn-up levels, increase operating cycle
lengths or extend plant lifetime. These issues include:
x
x
x
x
x
x
x
x
x

Reactivity initiated events (e.g. boron dilution).
Plant ageing.
Power uprates.
High burn-up fuel.
Materials cracking corrosion.
Installation of digital I and 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, similar to BWRs, emphasis on riskinformed regulation in some countries and more detailed assessments of accident scenerios andaccident management actions has led to new issues needing research, such as:xx

Redefinition of large break LOCA.
Break location and orientation.

2.3 VVER reactors
Introduction
Nuclear power plants with VVER reactors of Soviet origin are presently operated in sevencountries. Four of these are NEA/OECD countries (Czech republic, Finland, Hungary and Slovakia)
26

where the NPP with VVER-440/213 and VVER-1000/320 plants are in operation. VVER reactors areclassified as a specific type of PWR reactor. The review of their design and safety features shows thatthe main concept of these reactors is similar to PWR units designed at the same time in the othercountries.
VVER design features
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 of equipment
failures due to 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 core cooling 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 of these positive and negative features
were inherited by the next VVER generations, but the design organisations started to improve safety inline with the Western safety standards. Back-fitting of the existing reactors has been intensive during
the recent decade and the process is not much different from that which is going on in plants built toearlier safety standards all over the world.
VVER 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 NPPs with VVER reactors was reviewed in the framework of the Extrabudgetary
Programme (EBP) of the IAEA on the safety of VVER and RBMK NPPs during the period
1990-1998. 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.
As concerns safety ranking of the more modern reactor type VVER-440/213, no safety issues of
the highest category were identified. The safety issues with high safety concern, included 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 inservice inspection of reactor coolant system, ECCS clogging under LOCA, layout of the emergency
feedwater system, fire protection and possible multiple failures of safety related systems in highenergy pipe breaks at certain locations. These safety issues have since been resolved at all the plants
via back-fitting programmes.
The VVER-1000 concept may be considered much closer to the other PWRs. Its power is higher
and safety margins smaller than in the VVER-440 concepts, but in its safety philosophy the defense indepth has been taken into account from the beginning. In the IAEA safety review for the standard
series VVER-1000/320, no safety issues of the highest category were identified. Safety issues 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.
27

The majority of safety issues at all the VVER plants have been identified as deviations fromcurrent standards and practices, which have evolved since these NPPs were designed. In all countries
the extensive back-fitting and upgrading programmes have been performed which have resolved agreat majority of remaining safety issues.
The particular problem of VVERs is that their safety analysis was not validated via experimental
facilities. For this reason the experimental validation of computer codes and system behaviour is of agreat importance. As an example, the successful experimental validation of the bubble-condenser
system of VVER-440/213 was completed and experimental validation of thermal-hydraulic computercodes for VVER-1000/320 at the PSB-VVER facility is underway. During recent years the safety
analysis methodology has also been improved considerably via international co-operation inbenchmarking and validation exercises.
In the future it may be expected that the lifetime of the current VVERs will be extended. This
will lead to having to address many of the same issues as PWRs, such as:xxx

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 these will not be specific to VVERS.
2.4 Pressurised heavy water reactors (PHWRs) and advanced PHWRs
Introduction
PHWRs are unique for containing the nuclear fuel and coolant in an array of horizontal fuelchannels, rather than a pressure vessel. Beyond the headers that distribute coolant to individual feeders
for the fuel channels, the remainder of the 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 in a cylindrical calandria vessel. This moderator water is kept below 100oC,
and serves as a back-up 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
biological shield that provides an additional heat sink in a severe core damage accident.
PHWR design features
The fuel for a PHWR consists of a 0.5 m long bundle of fuel elements (28, 37 or 43, dependingon 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 made from Zircaloy.
PHWRs can be refueled on-power through the use of fuelling machines that connect to the ends
of a fuel channel. Once connected, the short fuel bundles can be repositioned in the channel tooptimise 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 no requirement 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.

28

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 fast-acting shutdown
systems. The first is spring-assisted shut-off rods that drive down between the fuel channels in themoderator. 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 tohigh pressure, nor jamming due to fuel damage, in the event of an accident.
Similar to all water-cooled reactors, PHWRs use an emergency core cooling system (ECCS) to
provide back up cooling in a LOCA. There are typically three modes of operation: high pressure
injection, intermediate/low pressure injection, long-term recovery and recirculation. Emergency corecooling is accompanied by venting steam from the secondary side to “crash cool” the steam generatorsand reduce the primary side pressure below the ECCS injection pressure.
There are currently three major types of containment, single unit containment, multiple unit
containment (incorporating a common vacuum building), and a double containment system used inIndian PHWRs.
The single-unit containment consists of a cylindrical, pre-stressed, post-tensioned concretebuilding with a concrete dome. The building has an epoxy lining to reduce leakage. Short-termpressure rises are mitigated with a dousing system, while local air coolers are used to provide
long-term pressure control and heat removal. Hydrogen igniters prevent build-up of hydrogen to
explosive levels.
In a multi-unit vacuum containment, 4 or 8 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 from the 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, an emergency filtered air
discharge system is used to control pressure, while filtering out fission products.
Current Indian PHWRs use a double concrete containment. The inner containment is a cylinderand dome of prestressed concrete, with an epoxy lining for leak tightness. The outer containment is acylinder and dome of reinforced concrete. The intervening space is maintained at a negative pressure
with a purging arrangement. A suppression pool between drywell and wetwell volumes in containmentis 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 isa filtered system for controlled gas discharge in the longer term.
PHWR safety issues
For operating PHWR reactors, the main residual safety issues revolve around improved
understanding of phenomena, and reduced uncertainties in safety code predictions. Information has
been generated to address the following generic safety issues:
x
x
x
x
x
x

Limited core damage accident (damage is contained within 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 subcooling requirements.
29

x
x

ECCS sump strainer clogging.
Flow distribution in headers during LOCAs.

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

Plant ageing.
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.

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.
Advanced PHWR (APHWR) designs are also being developed to improve the safety andeconomics of the PHWR. Key features being explored in APHWR designs include:xxx

The use of light water coolant in the primary coolant system with heavy water remaining as
the moderator in the calandria.
A more compact fuel channel arrangement to optimise neutron moderation and help
eliminate any positive coolant void coefficient.
The use of slightly enriched u instead of natural uranium.

Such features affect reactor physics parameters and, thus, the response of the plant under certain
accident situations. Accordingly, it is important to understand these affects in assessing plant safety.
NEA has published the results of a workshop held in February 2002 Advanced Nuclear Reactor
Safety Issues and Research Needs. It also contains information which may be of use to researchers,
designers and regulators in planning and conducting future work on APHWRs.
2.5 Advanced LWRs (ALWRs)
Introduction
ALWRs designs constitute improvements of current generation pressurised water reactors(PWRs) and boiling water reactors (BWRs). For the purpose of the SFEAR study, the ALWR designs
considered are those being developed for deployment in the next 5-10 years. One or more of thefollowing ALWR design features are likely to be employed on ALWR designs:
x
x
x
x
x

Longer design life (up to 60 years.).
Advanced materials more resistant to corrosion and cracking.
Advanced fuel designs.
Longer operating cycles.
The use of passive safety systems for emergency core and containment cooling and for decay
heat removal.

30

x
x

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., prestressed 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.
ALWR 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:xxxx

Advanced fuel transient performance.
Advanced materials performance.
Passive safety feature performance.
Use of digital technology and the impact on human performance.

NEA has published the results of a workshop held in February 2002 Advanced Nuclear Reactor
Safety Issues and Research Needs. It also contains information which may be of use to researchers,
designers and regulators in planning and conducting future work on ALWRs.
2.6 High-temperature gas-cooled reactors (HTGRs)
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 HTGR concept, dates in the United States from the latter
1950s when the design of the fully ceramic core and the use of the helium gas for cooling were
pioneered by the General Atomics. This development effort resulted in the 40 MWe Peach Bottom 1
HTGR, and the 330 MWe Fort St. Vrain HTGR, which adopted the block-type core. Also in the late
1950s, the Germany began designing the pebble-bed type of HTGR. Two HTGRs were constructed in
the Germany, the experimental 15 MWe AVR and the 300 MWe THTR300. All HTGRs mentioned
above have been decommissioned.
During the 1980s, the modular high temperature gas cooled reactor (HTGR) concept wasdeveloped, primarily in Germany and the United States of America. This concept utilised
characteristics of HTGR technology to arrive at a design wherein safety issues were addressed throughthe inherent response characteristics of the system. These initial modular HTGR designs wereprimarily directed toward electricity generation using a steam turbine.
The 1990s were witness to the initiation of plant designs that incorporated the reactor directly
coupled to a gas turbine power conversion system for the production of electricity. This design
replaced the steam cycle components with fewer gas turbine cycle components, and with an attendant
benefit of increasing net plant electrical efficiency from approximately 40% into the range of 45 to
50%. The elimination of the steam cycle reduces the potential for large water ingress events.
Investigation of the modular HTGRs as the high temperature heat source for industrial co-generationand non-electric applications (to realise products including hydrogen and synthesis fuels as well as the
production of electricity) is also taking place.
31

HTGR design features
HTGRs utilise a single phase gas coolant (usually helium) and graphite as a moderator. They may
range from small modular designs (100 MWe) to large central station plants. They can utilise steel
reactor pressure vessels (RPVs) or prestressed concrete reactor vessels (PCRVs) with a steel liner.
Their fuel can be in the shape of tennis ball size pebbles (to support on-line refueling) or be embedded
in stationary graphite blocks. However, in either case the basic component of the fuel is a small
particle of UO2, UCO or UC coated with multiple layers of graphite and silicon carbide, which areintended to retain the fission products. The current state-of-the-art in fuel coating is called the TRISO
process which involves three layers of coatings. HTGRs normally operate at core exit temperatures in
the range of 900qC to 1 100qC and claim that the coated fuel particles maintain their integrity up totemperatures of 1 600qC.
HTGR safety is centreed around ensuring that fuel quality is sufficient to maintain fuel integrity
at high temperatures. HTGRs also have a long response time for core heat-up due to the large heat
capacity of the graphite moderator. HTGRs may be designed with steam generators (usually in the
primary coolant system) or may utilise an in-line helium turbine (also usually in the primary system).
In addition to the above, future HTGRs are likely to contain one or more of the following features:
x
x
x
x
x
x

Passive reactor shutdown capability.
Passive decay heat removal.
Confinement building in lieu of a conventional low leakage containment building.
Modular design.
High temperature materials.
Automated control and safety systems with less reliance on operator action.

HTGR safety issues
Key safety issues associated with HTGRs include:x

Ensuring fuel quality over the life of the plant (this may involve more regulatory attention
being given to fuel fabrication controls, QA and sampling).

x

Performance qualification and in-service testing of passive safety features.

x

Potential for and the effect of air and/or water ingress into the primary system on:–
Reactivity.
–
Fuel integrity.
–
Graphite oxidation.

x

Qualification and acceptance of graphite structures (used in the RPV).

x

Fission product transport and release.

x

Staffing, in view of long response times, passive features and automatic systems.

xOn-line refueling of pebble-bed HTGRs.

x

High-temperature material performance (metallic materials): creep fatigue data; environmental
characteristics; and in-service inspection and surveillance plan and techniques.

x

Nuclear-grade graphite behaviour (including ceramic materials such as c/c composite):measurements of changes in physical properties induced by thermal, radiation and chemical
exposures; oxidation measurements in the event of an air-ingress accident; and in-service
inspection plans and techniques.

32xFuel performance: irradiation testing of fuel simulating steady state, reactivity insertion, andslow heat-up during transients, including fission product release data.

x

Analytical programmes and tools.

x

Thermo-fluid dynamics codes as well as severe accident analysis codes; data for code
validation and assessment; and development of probabilistic risk assessment models andapproaches, considering the new and different equipment that will be used.

xAnalytical models for events such as, air and water ingress, fission products release in an air
environment, fuel behaviour under reactivity insertion accidents.

x

Containment performance: evaluation of containment versus confinement option for all
accident scenarios, radiological source terms, and emergency planning.

NEA has published the results of a workshop held in February 2002 Advanced Nuclear Reactor
Safety Issues and Research Needs. It also contains information which may be of use to researchers,
designers and regulators in planning and conducting future work on HTGRs.

33

Chapter 3
TECHNICAL ISSUES AND ASSOCIATED FACILITIESThis chapter describes the safety research issues currently being investigated or identified as
needing investigation to support the continued operation of current plants and/or the development of
future plants. This chapter is organised into three sections as follows.
The section on issues and facilities unique to the nuclear industry addresses those technical areas
where experimental data specific to the nuclear industry may be essential for addressing the safetyissues. The technical areas addressed in this section are:
x
x
x
x
x

Thermal-hydraulics.
Fuel.
Reactor physics.
Severe accidents.
Integrity of equipment and structures.

For each technical area, the currently identified safety research issues associated with that area
are listed, along with the reactor types to which the issue applies. The reactor types addressed in thissub-section include BWRs, PWRs, VVERs, PHWRs, ALWRs and APHWRs. HTGR issues
corresponding to the above technical areas are addressed in Section 3.3.
In each technical area in Section 3.1, there are three tables. The first table (Table 1) is a list of thesafety issues that currently may require some degree of additional research to improve the state ofknowledge and support issue resolution. The second table (Table 2) lists the safety issues from Table 1
versus the facilities currently available that are conducting or have the potential to conduct research
relevant to each issue. In Table 2, the group’s views on the safety importance and state of knowledge
associated with each issue is shown, using the format and definitions for High, Medium and Low
shown in Chapter 1. Also shown on Table 2 is information related to the ability of each facility to
produce information relevant to the issue, also using the High, Medium and Low definitions listed at
the end of Chapter 1.
Finally, Table 2 also provides a brief description of the facility’s capability related to conducting
research on the issue. Table 3 provides a list of each of the facilities in Table 2 with information on
their cost of operation, replacement cost, planned duration of operation and additional description.
Also shown on Table 3 is a numerical score that represents the relative importance of the facility to beable to do research related to the issues in described Table 1. This numerical score was developed byassigning point values to each of the factors listed in Table 2, as shown below:
High
MediumLow

Safety relevanceState of knowledge

Facility importance

1.0
0.6
0.3

0.3
0.6
1.0

1.0
0.6
0.3

35

For each issue, a numerical score is calculated for each facility listed as applicable to that issue by
multiplying above the numerical scores that correspond to the high, medium or low designations listed inTable 2. The scores for each facility are summed over all the issues in a given technical area to whichthat facility applies. It is these summations that are shown in Table 3. These numerical rankings thenprovide a relative indication of the importance of each facility in the given technical area (the higher thescore, the more important, on a relative basis, is the facility to that technical area). It is important to note,however, that the numerical rankings are for the purpose of relative comparisons of facilities within agiven technical area. Numerical rankings should not be compared between technical areas.
The technical areas addressed in the section on issues and facilities not unique to the nuclear
industry are:xxxx

Human and organisational factors.
Plant control and monitoring.
Seismic behaviour of structures.
Fire assessment.

For each of these technical areas the currently identified safety research issues associated withthat area are listed, along with the reactor types to which the issue applies. The reactor types addressedin this section include BWRs, PWRs, VVERs, ALWRs, APHWRs and HTGRs.
However, data needs to resolve these issues can be met using facilities of low cost, not in danger
or not needing international support, not unique to the nuclear industry or where no facilities areneeded. Accordingly, this section does not make recommendations for CSNI considerations, but only
lists facilities for information, as appropriate.
The section on HTGR unique issues addresses the safety research issues unique to HTGRs. Theseissues are in the following technical areas:
x
x
x
x
x

Thermal-hydraulics.
Fuel.
Reactor physics.
Severe accidents.
Integrity of equipment and structures.

For each safety research issue listed in this section an assessment is made regarding the researchneeded to address each issue and high level recommendations are made regarding facility needs. No
recommendations for CSNI consideration are made in this section. Rather, the recommendations
regarding HTGRs are for use by HTGR designers and regulators in long-term planning.
Section 3.2 also contains safety research issues applicable to HTGRs, where those issues apply toother reactor types as well. Accordingly, to get a complete list of safety research issues applicable to
HTGRs, Section 3.2 must also be reviewed along with Section 3.3.
3.1 Issues and facilities unique to the nuclear industry
Thermal-hydraulics
Introduction
Thermal hydraulics became one of the main nuclear safety disciplines when postulated accidents
like the Loss-of-Coolant Accident (LOCA) and other thermal-hydraulic transients were identified as
the dominant safety concern for LWRs. As full-scale experimentation was not feasible in most36

situations, significant computational developments had to be undertaken to be able to properly
simulate such transients, as needed for the safety case of these reactors. Numerous national andinternational experimental programmes provided the data necessary for understanding the phenomena
and simulating them.
The CSNI has always considered with great attention the issue of thermal-hydraulic code
validation as well as the experimental database needed for such validation. An overview of the large
number of separate-effect test programmes that have been carried out in the past is given in /2/. The
results from these programmes provide a sound basis for model validation of traditional system codes,
whereas they are insufficient for multi-dimensional codes.
In recent years, the CSNI has taken initiatives 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. These projects are still ongoing and include SETH, PSB-VVVER and ROSA.
The SETH Project focuses on the capabilities of the PKL and PANDA facilities, which wererecommended for international consideration.1 The PKL experiments addressed the issue of potential
boron dilution accidents in PWR reactors. They are being continued under a new PKL project. The
PANDA experiments are to provide data on containment three-dimensional gas flow and distributionthat are important for code prediction capability improvements, accident management and design of
mitigating measures. In relation to VVER reactors the international PSB-VVER Project was started,with the objective to provide unique experimental data needed for the validation of thermal-hydrauliccodes used for the safety assessment of VVER-1000 reactors. Following a JAERI proposal, the CSNI
also recommended to make all necessary steps in order to establish an international experimental
project to be conducted in the Japanese ROSA facility.
Scope
As the scenarios of primary concern shifted from the Large Break LOCA to small breaks andother incidents (e.g. boron dilution), the thermal-hydraulic research effort shifted accordingly to cover
the more complex phenomena associated with this category of accidents. Improved computational
tools were also developed to properly handle these. Although reactivity-related accidents andtransients were of course considered from the beginning of the deployment of LWRs, increasingemphasis has been put on the accidents having a neutronic origin or 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 tosimulate properly some complex situations. Although many of the existing facilities are notsufficiently instrumented to be used to validate finely detailed analysis tools (e.g. CFD codes), they areincluded in this section for completeness.
The new concerns that regulators faced after the Chernobyl accident, more generally related tothe understanding and simulation of situations and phenomena in reactors designed in the former
Eastern block countries provided additional needs for research and development. The emergence ofadvanced LWRs having passive safety systems opened another new area of relatively new phenomena
and situations that had to be addressed. Power upgrading of existing reactors may also require
refinement and further validation of existing analytical tools and more extensive experimental databases. The thermal-hydraulic safety issues that could benefit from additional research are listed inTable 3.1.1-1.
1.

NEA (2001), Nuclear Safety Research in OECD Ccountries. Major Facilities and Programmes at Risk,
OECD, Paris.

37

Description
In spite of these continuing developments, often conducted internationally, a number of issues
still require some attention. New issues will certainly also arise in relation to the design and safety
analysis of future reactor systems. For example, as current plants continue 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 and near-term interest in the thermalhydraulics area.
x

LWRs. The CSNI carried out a review of current LWR system behaviour conducted throughlarge integral test programmes.2 This publication includes a systematic selection of openly
available data for code validation. Thermal-hydraulic data is routinely needed to assess
analytical tools, particularly when new issues or designs are to be analysed. The current
emphasis on risk-informed regulation in some countries, additional interest in more detailed
analysis of certain types of accidents and new plant designs provide incentive formaintaining experimental capability to validate analytical tools.

x

VVERs. The CSNI also carried out a review of the level of validation of thermal-hydrauliccodes applied to the analysis of VVER reactors.3 The aim was to supplement the review doneon integral and separate-effect test facilities, including special features of VVER reactor
systems with respect to large and small break LOCA.

x

PHWRs. For PHWR reactors, experimental programmes have been carried out in specialised
facilities (e.g. large-scale header facility, cold water injection facility) using full-size
components such as headers, fuel channels and end fittings. The facility RD-14M has beenused for a comprehensive programme on emergency core cooling system effectiveness,
natural circulation and shutdown cooling using a full-height, multi-channel test simulation ofa CANDU reactor cooling system.
Table 3.1.1-1. Current thermal-hydraulic safety issues and safety related phenomena

Issues and relevant reactors
1) Boron dilution:
PWR, VVER, ALWR

2.
3.

Description
Events where boiling and condensation can occur in the primarycooling system have the potential to form areas where non-borated
water can collect. If primary coolant pumps are started, or the restartof natural circulation occurs in combination with other unfavourable
circumstances, it was postulated that there wais the potential for aslug of non-borated water to enter the core and cause a criticality
concern. This issue involves understanding both the thermalhydraulic response as well as the reactor kinetics, with emphasis on
the 3-D aspects of boron transport and mixing in the RPV. Although
resolved in some countries (e.g., U.S.), this issue remains open inothers.

NEA (1996), CSNI Integral Test Facility for the Assessment of Thermal-hydraulic Codes for LWR LOCA
and Transients, NEA/CSNI/R(96)17, OECD, Paris.
NEA (1993), CSNI Separate Effect Test Matrix for Thermal-hydraulic Code Validation,
NEA/CSNI/R(93)14, OECD, Paris.

38

Table 3.1.1-1. Current thermal-hydraulic safety issues and safety related phenomena (Cont’d)Issues and relevant reactors

Description

2) Passive safety system
performance:
ALWR, APHWR

Some future designs propose the use of passive systems or features
for emergency core cooling, decay heat removal and/or containment
cooling. Although some experimental work has been done on
passive systems in the past, the applicability of this work to future
designs will be dependent upon the design and proposed operating
conditions. Accordingly, the capability to experimentally investigate
passive safety system performance should be maintained and used tovalidate analytical tools and confirm system performance, including
reliability and performance under a range of conditions that could
result from aging, environmental conditions, etc.

3) Non-pipe breaks:
BWR, PWR, VVER
ALWR, PHWR andAPHWR

Breaks in plant components can affect plant response to the event,including actuation of engineered safety systems. Additional data isneeded to confirm analysis of such conditions, considering locationand orientation of the break.

4) Steam generator tube
rupture:
PWR, VVER, ALWR,
PHWR, APHWR

Breaks in SG tubes have the potential to remove coolant inventory
from the reactor coolant system and bypass the containment, thus
affecting the coolant inventory available for recirculation and theability to provide sufficient coolant to makeup the loss. The responseof the plant to various numbers and locations of SG tube ruptures
requires experimental data to confirm analysis methods.

5) Stability and power
oscillations:
BWR, ALWR

Plant owners are raising power levels at some plants in order toimprove economics. The impact of higher power levels on BWR
stability and BWR power oscillations, particularly under ATWS
conditions, needs to be understood. Also, for future BWR designs,
this issue may need further research to confirm predicted behaviour.

6) ECCS strainer clogging:
PWR, BWR, VVER,
PHWR

Performance of the ECCS in the recirculation mode is dependent
upon the ability of the containment sump strainers to remove debris,without plugging to the point of cutting off recirculation flow. This
includes consideration of solid debris as well as chemical effects
which can cause gelatinous material.

7) Pressure tube reactor
system T-H:
PHWR, APHWR

In a PHWR, coolant is distributed to the fuel channels via individual
feeder pipes from a distribution header. The performance of the
PHWR primary coolant and ECCS system needs to be assessed for awide variety of possible LOCA scenarios.

39

Table 3.1.1-1. Current thermal-hydraulic safety issues and safety related phenomena (Cont’d)Issues and relevant reactors

Description

8) Two phase natural
Two phase natural circulations may be expected in PWRs during
circulation:
LOCA accident conditions or during an ATWS. The flow rate may
PWR, BWR, VVER,
be oscillatory under some conditions and depends on the circuit
PHWR, ALWR, APHWR geometry. In BWRs with natural circulation, the oscillations in twophase natural circulation may lead to power oscillations. Under low
power conditions in BWRs, non-uniform conditions may occur inthe reactor vessel. Circulation characteristics at low power havelarge uncertainty. In the VVER-400 the hot leg loop seals may
actuate oscillations. The phenomena may be studied by using plant
specific test facilities. Natural circulation can be effective underpressurised or depressurised conditions and can be used for decay
heat removal or power operation. Uncertainties associated withreliability, predictability, stability and effectiveness of two-phasenatural circulation heat removal need to be better understood,modelled and verified, including the effects of non-condensable gas,
to support improved analysis.
9) Thermal stratification:
BWR, ALWR

Thermal stratification in large components (e.g., pipes, tanks,
vessels) can cause structural stresses and/or affect the performanceof T/H systems (e.g., cause density gradients that could inhibit
passive system performance). Thermal stratification is moreprevalent and more a concern in systems that employ natural
circulation water injection and/or heat removal.

10) Thermal cycling:
PWR, BWR, VVER,
PHWR, ALWR and
APHWR

In piping systems, where two or more flow paths meet, there may be
the potential for fluids of different temperatures to cause localised
cycling of the temperature of surrounding components (e.g., fittings,etc.) due to the turbulent mixing of the streams of different
temperature fluids. Such thermal cycling of materials can cause
alternating thermal stresses and ultimately thermal fatigue failure of
the material. Identification of such locations, non-destructive
examination of crack status, and understanding the mixing
phenomena is important to developing corrective measures and toensuring system integrity.

11) Moderator thermalhydraulics:
PHWR, APHWR

The moderator in a PHWR serves as a back-up heat sink for
accidents involving loss of primary and emergency cooling. The
three-dimensional flow regime in a horizontal tube bank needs to beassessed.

12) 3-D core flow
distributions:
BWR, PWR, VVER
and ALWR

The trend towards higher power densities means that in open cores,
DNB in the maximum loaded fuel bundle and single rod may be
prevented only due to the cross flow connections inside the core. For
the simulation multidimensional analysis tools are needed. For
model development, multidimensional core void and temperatureexperiments are needed.

40

Table 3.1.1-1. Current thermal-hydraulic safety issues and safety related phenomena (Cont’d)Issues and relevant reactors

Description13) Flow distribution, mixing
and stratification in cold
legs and downcomer:
PWR and ALWR

Data from separate effects mixing experiments (e.g., Creare) andintegral system tests (e.g. UPTF, LSTF, APEX, and LOFT) showpronounced thermal stratification of ECC flow in the cold leg, butthat the downcomer is well mixed. However, more detailed data
would help define the mixing processes and provide information tobenchmark CFD codes. It should be noted that systems codes do not
represent fluid-fluid mixing and stratification phenomena.

14) Accidents initiatedduring shutdown:
BWR, PWR, VVER,
ALWR, PHWR and
APHWRMost work to date in the thermal-hydraulic area has been associatedwith analysing accidents initiated from full power conditions.
However, during shutdown conditions, plant configurations,temperature, pressure, system operability and coolant inventories
may be substantially different than they are at full power.
Accordingly, accidents that occur will likely have a much different
progression, timing and potential consequences than those at fullpower. For example, loss of residual heat removal (RHR) under
shutdown conditions, i.e. during mid-loop operation, has occurred
several times world-wide and still plays an important role in risk
studies for PWRs. When the RCS is open at the time of RHR failure,
the decay heat cannot be transferred to the steam generators andother measures become necessary in order to prevent or compensate
for the loss of inventory that finally would lead to core damage. To
ensure analytical capabilities are adequate, experimental
confirmation of system behaviour will likely be required.

41

Table 3.1.1-2. Issues versus facilities (Thermal-hydraulics)

Issue

Applicabilityof issue

Safety relevanceof issueState of knowledge
on issue1) Boron dilution

PWR, VVER, ALWR

MediumHigh(except for 3-D effects which are
medium)2) Passive safety system
performance

ALWR, APHWR

High

Medium3) Non-pipe breaks

BWR, PWR, VVER,
PHWR, ALWR

MediumLow4) S.G. tube rupture

PWR, VVER, ALWR,
PHWR

High

High

5) Stability and power
PWR, BWR, VVER
oscillations
6) ECCS strainer clogging PWR, BWR, VVER

High

MediumHigh

Medium7) Pressure tube reactor T/H PHWR, APHWR

High

Medium42

Table 3.1.1-2. Issues versus facilities (Thermal-hydraulics)
Facility
Name

Importance of facility toresolution of the issue?

Versatility

LSTF/ROSAPSB-VVER
PKL

High
High
High

PACTEL

MediumAPEX
ATLAS
LSTF/ROSA

MediumMediumHigh

PSB-VVER
PKL

High
MediumRD-14M
PACTEL
APEX

High
High
MediumPUMA

Medium

PANDA
ATLAS

High
High

LSTF/ROSAPSB-VVER
PKL

High
High
High

PACTEL
ATLAS

MediumHigh

RD-14M
LSTF/ROSAPSB-VVERSKODA-VVS
PKL
RD-14M
PACTEL
APEX
ATLAS
PUMA
THYNC
No large-scale
facilities identified
RD-14M

High
High
High
MediumHigh
High
MediumHigh
High
MediumMediumLarge-scale PWR systems simulation.
Large-scale VVER systems simulation.
Large-scale PWR systems simulation, with on-line boron
concentration measurement capability.
Can simulate non-borated slug formation in boiler condenser
mode or during S.G. tube ruptures.
¼ scale passive PWR systems.
Medium scale PWR Systems simulation for both DVI and CLI.
Large-scale active and passive PWR systems simulation. Fullheight, full pressure.
Large-scale active and passive VVER systems simulation.
Large-scale PWR system simulation. Can simulate gravity feedfrom accumulators.
PHWR full scale primary circuit.
Active and passive systems simulation.
¼ scale, passive PWR systems simulation from intermediatepress.
¼ scale, passive BWR systems simulation from intermediatepressure.
Large-scale passive systems.
Active and passive PWR system simulation for various injection
modes.
PWR systems simulation, including non-pipe breaks.
VVER systems simulation, including non-pipe breaks.
Can simulate non-pipe breaks with asymmetric loop behaviour
among PKL’s 4 loops.
Active and passive systems simulation.
Active and passive PWR system simulation for various injection
modes.
Can simulate header breaks.
PWR systems simulation.
VVER systems simulation.
PWR simulation.
Simulation with all relevant primary and secondary systems.
PHWR primary circuit.
Can simulate S.G. tube rupture in horizontal S.G.
¼ scale facility.
PWR systems simulation.
BWR systems with neutronic simulation.
Can simulate coupled neutronics.

High

PHWR full-scale primary circuit.

43

Table 3.1.1-2. Issues versus facilities (Thermal-hydraulics) (Cont’d)

Issue

Applicabilityof issue

Safety relevanceof issueState of knowledge
on issue

8)

Two-phase national
circulation

PWR, BWR, VVER,
PHWR, ALWR, APHWR

High

Medium

9)

Thermal stratification

BWR, ALWR

Low

Medium10) Thermal cycling

PWR, BWR, VVER,
PHWR, ALWR, APHWR

Low

Medium11) Moderator T/H
12) 3-D core flow
distributions
13) Downcomer flow
distribution

PHWR, APHWR
PWR, BWR, VVER,
ALWR
PWR, ALWR

MediumMediumHigh
MediumLow

Medium14) Accidents initiatedduring shutdown

BWR, PWR, VVER,
ALWR, PHWR, APHWR

High

Medium44

Table 3.1.1-2. Issues versus facilities (Thermal-hydraulics) (Cont’d)
Facility
Name

Importance of facility toresolution of the issue?

LSTF/ROSAPSB-VVER
PANDA

MediumMediumMediumRD-14M
PKL

MediumHigh

PACTEL
PUMA
APEX
ATLAS
SKODA-VVS
LSTF/ROSAPSB-VVER
PUMA
APEX
PANDA

MediumMediumMediumMediumMediumMediumMediumMediumMediumHigh

PACTEL
ATLAS
PKL
PANDA

MediumMediumMediumMediumPACTELMTF
No large-scale
facilities identified
LSTF/ROSA
PKL
ATLAS
MIDAS
LSTF/ROSA
PKL

MediumHigh

MediumMediumMediumMedium
High
High

ATLAS
PSB-VVER
PANDA
PUMA
APEX
RD-14M
PACTEL

Low
High
High
MediumMediumHigh
High

Versatility
Active and passive PWR systems simulation.
Active and passive VVER systems simulation.
RPV of PANDA has a large-scale 3-D-RPV and a natural
circulation 2-phase loop.
PHWR primary circuit.
PWR systems simulation with 4 identical loops for assessing
asymmetric effects.
Active and passive system simulation with horizontal S.G.
¼ scale.
¼ scale.
Active and passive systems. Small scale.
PWR simulation.
Active and passive PWR systems simulation.
Active and passive VVER systems simulation.
¼ scale.
¼ scale.
Large vessels of PANDA have been used for thermal
stratification studies.
Can simulate thermal stratification in tanks and pools.
PWR system simulation.
Active and passive PWR systems simulation.
Large vessels of PANDA can be used for detailed investigationsof mixing.
Can simulate thermal cycling at joints.
Moderator T/H flow distr.

Large scale, full height, full pressure.
Large scale, full height.
Annular downcomer.
Annular downcomer.
Can simulate open PWR primary system at large scale.
Can simulate open PWR primary system at large scale,including mid-loop operation.
Small volume scale.
Large scale.
Large scale.
¼ scale.
¼ scale.
Large scale.
Large scale.

45

Table 3.1.1-3 Facilities in the thermal-hydraulics area
Facility nameApplicability(type of reactor)

Cost/year,
operation

Replacement
cost

IssuescoveredLSTF/ROSA (Japan)*

PWR, ALWR

High

High

1,2,3,4,8,9, 13,14

PSB-VVER (Russia)

VVER, ALWR

MediumHigh

1,2,3,4,8,9,14

THYNC (Japan)
PKL (Germany)

BWR
PWR

Low
Medium

High
High

5
1,2,3,4,8,9,13,14

APEX (USA)

PWR, ALWR

Low

High

1,2,4,8,9,14

PUMA (USA)

BWR, ALWR

Low

Medium

2,5,8,9,14

PANDA (Switzerland)

ALWR, BWR, PWR

MediumHigh

2,8,9,10,14

RD-14M (Canada)
MTF (Canada)
SKODA-VVS(Czech Republic)
ATLAS (Korea)

PHWR, APHWR
PHWR, APHWR
VVER, BWR, PWR

Medium
Low
Low

High
MediumHigh

2,3,4,7,8,14
11
4,8

PWR, ALWR

MediumHigh

1,2,3,4,8,9,13,14

MiDAS (Korea)

PWR, ALWR

Low

Medium13

PACTEL (Finland)

VVER

MediumMedium

1,2,3,4,8,9,1014

Notes:
Specify range: Low, Medium, High.
Operation cost
: Low is < 1.0 MUS$/y; Medium is 1.0-2 MUS$/y; High is >2 MUS$/y.
Replacement cost : Low is < 2 MUS$; Medium is 2-10 MUS$; High is >10 MUS$.
* Also discussed in Severe Accidents – Section 3.1.4.

46

Table 3.1.1-3 Facilities in the thermal-hydraulics area
Planned durationof operation

Capability

Full height full pressure-2 loop PWR. Active and passive
Through  2009
system simulation capability (1/48 volume and power scale).
Full height, full pressure VVER facility. 1/300 volume and Through 2010
power scale. Active and passive systems.
Coupled, simulated neutronics and T/H simulation.
Indefinite
Scaled PWR Facility. Full-height, 40 bar pressure.
Through 2006
1/145 volume scale, 4 identical loops and secondary systems.
¼ height, reduced pressure, 2-loop PWR, including passive Through 2006
ECCS.
¼ linear scale – low pressure-BWR, including passive ECCS To be put in standby in 2007
and DHR systems.
Full height, very large interconnected volumes, low pressure, Until the end of the SETH Projectin 2006
large steam generation capability, air and helium supply
systems.

Relativeranking
2.9
2.8
0.4
2.9
1.6
1.6
1.9

Full height, full pressure.
¼ scale calandria vessel and loop.
Large Water loop for T/H experiments.

Through 2010
Through 2008
New programme (indefinite)

3.1
0.2
0.5

½ height, 1/1444 vol. scale, full pressure and temperaturefacility for APR-1400. Includes annular downcomer, direct
vessel injection, and cold leg injection. Can simulate mostaccidents and transients.

Through 2015

2.4

1/5 linear scale, 1 Mpa-pressure and 300 C super heatedThrough 2010
steam test facility for APR1400 with DVI; separate effecttest facility, steady-state ECC bypass for DVI injection mode
during LBLOCA late reflood phase.

0.1

Full height VVER-440 scale facility (1:305 volumetricscale), high pressure, 1/5 scaled full power. Facility to bemodified to simulate western PWR designs.

2.4

Indefinite

Notes:
Specify range: Low, Medium, High.
Operation cost
: Low is <1.0 MUS$/y; Medium is 1.0-2 MUS$/y; High is >2 MUS$/y.
Replacement cost : Low is <2 MUS$; Medium is 2-10 MUS$; High is >10 MUS$.
* Also discussed in Severe Accidents - Section 3.1.4.

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 (NPP). Fuel that performs well means fewer plant shutdowns,
less radioactive contamination, less radiation exposure to operations and maintenance personnel andless potential for releases of radioactive material offsite. The role of fuel performance in NPP safety
can vary depending upon plant design and technology. Plants are generally designed to prevent fueldamage for events that are expected to occur one or more times during the life of the plant and formore rare events varying degrees of fuel damage may be allowed.
47

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 the plant. Thesefission 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 the design and qualification ofplant safety systems, site suitability and emergency preparedness.
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 primarily UO2 for LWRs, ALWRs and
PHWRs, although mixed oxide (PuO2 and UO2) fuel is of interest in some countries. Burnable poison(such as Gd2O3) may also be present in the fuel to compensate for excess reactivity at the beginning oflife. Cladding materials are made from various zirconium based alloys and it is the cladding
performance, under accident conditions, which is of primary interest in assessing fuel performance.
The safety issues of interest that could benefit from additional research are those associated with
understanding and establishing fuel damage limits necessary for licensing purposes, as well as
understanding fuel behaviour over a range of design and beyond design basis conditions. These safetyissues are listed in Table 3.1.2-1. Programmes, facilities and analytical tools necessary to understandand resolve the outstanding safety issues are the subject of the remainder of this chapter.
Description
Fuel designs vary by reactor type and technology. Described below for each reactor type issummary information regarding fuel.
x

LWRs (including VVERs). For LWRs, the uranium enrichment ranges from two to nearly five
percent, although in some countries LWR recycle mixed oxide fuel is being used in portionsof 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) arehoused in zirconium alloy tubes (called cladding) 12 to 14 feet long. These tubes are thenassembled into tube bundles of varying sizes which are then inserted into the reactor core ina vertical orientation. The fuel design (U enrichment, cladding material, service condition,etc.) determines how long the fuel can stay in the reactor and still be able to perform
satisfactorily under steady state and accident conditions. For currently operating LWRs, theaccidents that pose the greatest potential for fuel damage are large reactivity insertions, lossof coolant events and events that cause dry out or departure from nucleate boiling on thecladding. The behaviour of the fuel under these conditions will be affected by the fuel burnup level (which affects internal fuel pin pressure), fuel cladding condition (e.g. corrosion,oxidation and embrittlement) and location in the core. For economic reasons, increases inburn-up levels are being considered in many countries. To support higher burn-ups, newcladding 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 upon how fuel performance under accident conditions changes withchanges in burn-up, cladding material and service condition.

x

ALWRs. For ALWRs, it is expected that high burn-up levels will be desired for economic
and possibly security reasons (e.g., proliferation resistance). To support such fuel designs
burnable positions and additional advances in cladding materials will likely be needed.
48

Ensuring these new fuel designs achieve an acceptable level of safety will require testing andanalysis to confirm fuel performance, validate analysis tools and establish safety limits.
However, such testing may only require analyses and unirradiated clad testing, or may
require more extensive testing, depending upon the fuel design and the types of accidents
that may be postulated.
x

PHWRs. PHWR fuel has many of the same basic characteristics as LWR fuel, although it isalways UO2 composition. The major fuel design differences are the shorter fuel pin length(1.5 feet versus 12-14 feet), smaller pin bundles, low enrichment or the use of naturaluranium and horizontal orientation in pressure tube core geometry. Generally, PHWR fuel isdesigned for lower burn-ups than LWR fuel; however, the safety issues and performance
concerns are essentially the same as for LWR fuel. However, PHWRs operate using an online refuelling system and thus the fuel handling (and potential refuelling accidents) of
PHWR fuel is very different than that for LWRs.
Table 3.1.2-1. Current nuclear fuel issues

Issues and relevant reactors
1) Response to LOCAs:
PWR, BWR, VVER,
PHWR, ALWR, APHWR

2) Response to reactivity
insertion accidents:
PWR, BWR, VVER,
PHWR, ALWR, APHWR

3) Response to power
oscillation events:
BWR, ALWR

4) Fuel performance understeady state conditions:
PWR, BWR, VVER,
PHWR, ALWR, APHWR

5) New materials:(fuel property database)
PWR, BWR, VVER,
ALWR

Description
As fuel designs change to achieve higher burn-up (e.g. through the use of
new cladding materials), or to utilise MOX fuel, the response under LOCA
conditions needs to be investigated to support development of appropriate
criteria that ensure coolable geometry is maintained during and after designbasis LOCA events. Experimental data is needed, consistent with the designbasis, and perhaps beyond the design basis LOCA. Also, small break LOCAs
can lead to clad embrittlement. Some experimental data on small break
LOCAs are useful to confirm fuel clad condition after such events and such
experiments are underway.
As fuel designs change to achieve higher burn-up (e.g. through the use of
new cladding materials) or to utilise MOX fuel, the response to reactivity
insertion accidents needs to be investigated. Failure modes and criteria
associated with new cladding materials and MOX fuel are currently being
investigated through experimental programmes, consistent with design basis
and beyond design basis events.
As fuel designs change to achieve higher burn-up (e.g. through the use of
new cladding material) the response to power oscillation events (such as
could occur due to an ATWS or a loss of stability) needs to be determined.
Experimental data is needed to establish failure modes and limits; however,conducting experiments simulating these conditions is difficult.
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 affectthe ability of the fuel to withstand design basis accidents. Therefore, theperformance of new fuel under steady state conditions is important toensuring the safety of operating personnel and to understanding andpredicting fuel performance during design basis accidents (i.e. the conditionof the fuel and cladding resulting from steady state operation represents theinitial conditions for transients).
Improving plant performance in both existing and future plants will likely
include improving fuel performance with respect to higher burn-up andpower levels. These improvements will require new cladding materials, andthe use of burnable poisons. The performance of these new materials andpoisons will need verification to ensure their safety and to establish a fuelproperty database for safety analyses.

49

Table 3.1.2-2. Issues versus facilities (Nuclear fuel)Applicabilityof issue

Safety relevanceof issueState of knowledge
on issue

1) Response to LOCAs

PWR, BWR, VVER,
PHWR, ALWR, APHWR

High

Low

2) Response to RIAs

PWR, BWR, VVER,
PHWR, ALWR, APHWR

MediumMedium3) Response to power oscillations

BWR, ALWR

High

Low

4) S.S. fuel performance

PWR, BWR, VVER,
ALWR, PHWR, APHWR

MediumMedium5) New materials(fuel properties data base)

PWR, BWR, VVER, ALWR

MediumMediumIssue

50

Table 3.1.2-2. Issues versus facilities (Nuclear fuel)FacilityName
ANL-Hot Cells
Halden + hot cells
NRU and CRL-hot cells
MIR + hot cells
RIAR-hot cells
Phebus
Cabri
Cabri
NSRR
BIGR
HaldenNSRR
Cabri
HaldenNRUATR
JMTR
HANARO
BR-2
HFR
OsirisHaldenATRJMTR
HANARO
BR-2
HFR
Osiris
NRU

Importance of facility toresolution of the issue?
High
MediumMediumHigh
Medium
High
High
High
High
High
High
High
High
High
High
High
High
High
High
High
High
High
High
High
High
High
High
High
High

51

Versatility
Can be used for other irradiated material tests.
Integral (single rod) test.
Can be modified to simulate fuel heatup.
In-reactor LOCA tests.
Simulated LOCA tests.
Capability for in reactor testing.
Capability for in reactor testing.
Can run a range of RIA tests.
Can run short pulse RIA tests.
Can run broad pulse tests.
Capability for in reactor testing.
Capability for in reactor testing.
Capability for in reactor testing.
Full range of instrumentation capability.
Can test full size PHWR fuel bundles.
Can be used for LWR and PHWR fuel testing.

Full range of instrumentation capability.
Can be used for LWR and PHWR fuel testing.

Can test full size PHWR fuel bundles.

Table 3.1.2-3. Facilities in the fuel area
Facility name
ANL hot cells* (USA)Phebùs and CEA Hot Cells(LECASTAR, PELECI) (France)

Applicability(type of reactor)
PWR,BWR, VVER, PHWR,
ALWR
PWR, BWR, VVER, ALWR

Operation
cost/year
High

Replacement
cost
High

High

High

Cabri + CEA hot cells(LECASTA, PELECI) (France)
NSRR + hot cells (Japan)

PWR, VVER, ALWR

High

High

PWR, BWR, ALWR

High

High

Halden Reactor + hot cells (Norway)

PWR, BWR, VVER, PHWR,
ALWR
LWR, PHWR, APHWR

High

High

High

High

PWR, BWR, ALWR
PWR, BWR, ALWR
PWR, BWR, VVER, PHWR,
ALWR
PWR, BWR, VVER, ALWR
PWR, BWR, VVER, ALWR
PWR, VVER, ALWR
PWR, BWR, VVER, ALWR

High
High
High

High
High
High

High
High
High
High

High
High
High
High

PWR, VVER, ALWR
PWR, VVER, ALWR

High
High

High
High

NRU and Chalk River Lab Hot Cells(Canada)ATR (USA)JMTR (Japan)
HANARO (Korea)
BR-2 + Hot Cells (Belgium)
HFR (Netherlands)
BIGR (Russia)
Osiris + CEA hot cells(LECASTAR, PELECI) (France)
MIR + hot cells (Russia)
RIAR-hot cells (Russia)

Notes* Hot cells also discussed under Integrity of Equipment and Structures Section.
Specify range: Low, Medium, High.
Operation cost
: Low is <1.0 MUS$/y; Medium is 1.0-2 MUS$/y; High is >2 MUS$/y.
Replacement cost : Low is <2 MUS$; Medium is 2-10 MUS$; High is >10 MUS$.

52

Table 3.1.2-3. Facilities in the fuel area
Issuescovered1
1

1, 2, 3Capability
Conducting simulated LOCA tests, on irradiated clad.
Cannot handle full length fuel assemblies.
Capable of conducting 25 pin bundle in-reactor LOCA
tests LECASTAR hot cells can conduct simulated
LOCA tests.

Planned durationof operation
Through 2008.

Relativeranking
1.0

In danger of shutdown
in 2007.

1.0

Indefinite.

2.4

Through 2014.

0.7

Indefinite.

2.3

Through 2010.

1.3

1,4,5

Conducting in reactor RIA tests. Good in-core
instrumentation capability.
Pulse reactor for conducting in reactor RIA
experiments.
19 MWt test reactor mostly for conducting in reactor
S.S. tests. Good in-core instrumentation capability.
135 MWt S.S. test reactor.

4,5
4,5
4,5

250 MWt S.S. test reactor.
50 MWt S.S. test reactor with power ramp capability.
30 MWt test reactor.

Indefinite.
Through 2008, Future beyond
2008
under discussion.
Indefinite.

0.7
0.7
0.7

4,5
4,5
2
4,5

60 MWt test reactor.
45 MWt test reactor.
In-reactor reactivity pulse testing.
70 MWt test reactor.

Through 2016.
Through 2015.
Indefinite.
Through 2014.

0.7
0.7
0.4
0.72

1
1

Can conduct DBA and beyond DBA - LOCA tests.
Can conduct simulated LOCA tests in hot cells.

Indefinite.
Indefinite.

1.0
0.6

2, 3
1,3,4,53.1.3 Reactor physics
Introduction
Reactor physics issues are becoming increasingly important as plants continue to seek improved
performance through power uprates, higher burn-up fuel, longer operating cycles, etc. Coreconfigurations are becoming increasingly heterogeneous in composition and distribution of power
generation. This makes prediction of core 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. 3-D neutronics) tobetter refine safety analyses and safety margins has emphasised the need for more detailed reactor
physics data and experimental confirmation of analytical methods. Also, thermal-hydraulic andneutronic codes are being coupled to address issues such as boron dilution and ATWS and to analyse
PHWR pressure tube reactors.
It is recognised that the NEA Nuclear Science Committee (NSC) has activities in the reactor
physics area and is also concerned about the status of key facilities. Accordingly, this section has been
written in co-operation with the NSC. However, to ensure an integrated approach to the preservationof critical facilities in this area, the NSC will take the lead to monitor facility status and recommend
appropriate actions for consideration by the NSC and CSNI.

53

ScopeThe scope of work in reactor physics covers the current and future needs of nuclear power plants
such as pressurised water reactors and VVER reactors, boiling water reactors, gas cooled (thermal)
reactors, light water and heavy water moderated reactors, gas cooled (fast) reactors, liquid metal fast
reactors, and molten salt reactors. The latter three are outside the scope of this report, yet experiments
and research on these systems contribute to a wider and more comprehensive validation of the models
and computer codes used and to their further development.
Reactor physics data (cross sections, neutron spectra, reactivity coefficients, etc.) and facilities tomeasure data for code assessment are the areas of interest in this section. These issues and facilities areshown in Tables 3.1.3-1 and 3.1.3-3, respectively.
Table 3.1.3-1. Current reactor physics issues
Issues and relevant reactors1) MOX fuel data:
PWR, BWR, VVER, ALWR,
PHWR

2) High burn-up fuel data:
PWR, BWR, VVER, ALWR,
APHWR
3) Coolant void coefficient:
PHWR, APHWR

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

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

6) Moderator coefficients:
PHWR, APHWR

Description
Reactor physics data to support the use of MOX fuel in current and future
reactors is essential to ensure safe operation. Such data includes crosssections, and their uncertainties, delayed neutron generation, power
distributions, decay heat production and power, temperature and void
coefficients. This issue relates to the use of weapons grade Pu in LWRs,
PHWRs and VVERs as well as the use of PWR recycle Pu. Advanced fuelcycles involving Pu and other actinides are also being studied for use in
LWRs and PHWRs.
Reactor physics data to support the use of high burn-up fuel in current andfuture reactors is essential to ensure safe operation. Such data includes
cross-sections, and their uncertainties, delayed neutron generation, power
distribution and power, temperature and void coefficients.
LOCA 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 the subsequent
reactivity effect) needs to be understood and included in PHWR safety
analysis.
The neutron flux and spectra on the reactor pressure vessel (RPV) internal
structures and the RPV wall are critical to determining material
embrittlement, component lifetime and the potential for RPV failure due
to pressurised thermal shock. Such data is 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.
The ability of materials inside the reactor vessel to shield key components
from irradiation induced damage is key to understanding their lifetime andability to withstand transients. Also, protecting operating personnel andpredicting the environment in which equipment must function depends
upon predicting shielding performance.
The coolant and moderator in PHWRs are separate. Thus the impact onreactivity of changes in the heavy-water moderator temperature, density,and poison concentration must be included in the safety analysis.

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

Reactor core and fuel-cycle physics issues at very high burn-up.
54

x
x
x

Minor actinides recycling in LWRs.
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 from current
operation and extended plant lifetime.

Support facilities for providing 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. Specificnew experiments are required, many of which can be covered by existing 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:
x
x
x
x
x
x
x
x
x
x

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 requirements fromadvanced reactors.
The Nuclear Science Committee together with the OECD/NEA Data Bank, in collaboration withthe member countries and other specialised institutions have developed data bases with evaluated andqualified experimental data shared internationally in addition to a large set of computer codes covering
the different needs in nuclear applications modelling. The databases cover:
x
x
x
x
x
x

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).

55

Table 3.1.3-2. Issues versus facilities (Reactor physics)
Safety relevanceof issueState of knowledge
on issue

PWR, BWR, PHWR,
APHWR, VVER, ALWRHigh

Medium2) High burn-up fuel data

PWR, BWR, VVER, ALWR,
APHWR

High

Medium3) Coolant void coefficient

PHWR, APHWR, PWR BWR,
ALWR

High

Medium4) Neutron flux and spectra

PWR, BWR, VVER, ALWR,
PHWR, APHWR
PWR, BWR, VVER, PHWR,
ALWR, APHWR
PHWR, APHWR

High

MediumMediumMediumMediumMediumIssue

Applicability of issue1) MOX fuel data

5) Shielding
6) Moderator coefficients

Table 3.1.3-3. Facilities in the reactor physics area
Facility name
MINERVE (France)
VENUS (Belgium)
ZED-2 (Canada)
PROTEUS (Switzerland)

Applicability(Type of reactor)
PWR, BWR, ALWR
PWR, BWR, ALWR, PHWR,
APHWR
PHWR, APHWR
PWR, BWR, VVER, PHWR, ALWR,
APHWR

LR-0 (CZECH Republic)

VVER (issue 4). All (issue 5)

TCA (Japan)
KUCA (Japan)

PWR, BWR, ALWR
PWR, BWR, ALWR

Notes:
Specify range
Operation cost

Cost/Year
operation
MediumHigh

Replacement cost
M US$
High
High

Low
MediumHigh
High

Medium

High

Low
Low

High
High

: Low, Medium, High.
: Low is <1.0 MUS$/y; Medium is 1.0 MUS$/y; High is > 2 MU$/y.

Replacement cost : Low is <2 MUS$; Medium is 2-10 MUS$; High is > 10 MUS$.

56

Table 3.1.3-2. Issues versus facilities (Reactor physics)
Facility
Name
ZED-2
TCA
PROTEUSVENUS
MINERVE
VENUS
PROTEUS
ZED-2
TCA
KUCA
PROTEUSMINERVE
VENUS
LR-0

Importance of facility toresolution of the issue?
High
MediumHigh
High
Medium
High
High
High
High
High
High
MediumHigh
High

Versatility
Critical assembly.
Reactor physics experiments.
Large test volume.
Zero power reactor.
Limited.
Extensive.
Extensive.
Critical assembly.
Extensive.
Extensive.
Extensive.
Limited.
Zero power reactors.
Extensive.

LR-0

High

Can mock-up various shielding configurations.

ZED-2
TCA
KUCA
MINERVEPROTEUS
VENUS

High
MediumMedium
Low
High
High

Critical assembly.
Zero power research reactor.
Zero power research reactor.
Critical assembly.
Extensive.
Extensive.

Table 3.1.3-3. Facilities in the reactor physics area
Issuescovered
2,3,6

Criticality safety.

Planned durationof operation
Indefinite

1,2,3,6

Zero power critical reactor.

Indefinite

2.2

1,3,6
1,2,3

Critical assembly.
Critical facility:
Large experimental volume (9 full length BWR fuel assemblies):
Driven test zone with capability for fresh and irradiated UO2 fuelcan be used for strongly poisoned lattices at high void conditions
and for non-LWR fuel.
Critical, VVER assembly with long experimental volume forneutron flux and reactivity measurements.
Reactor physics experiments, teaching and training.
Fundamental research and development, education and training.

Through 2008
Through 2011

1.6
2.2

Through 2010

1.0

Indefinite
Indefinite

1.2
0.8

4,5
1,3,6
3,6

Capability

57

Relativepriority
0.8

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 improvement of yields offission product isotopes in the fission of most heavy isotopes and decay schemes and energy yields ofradioactive isotopes, are required. In general, higher than current resolution cross-section
measurements, from thermal energies to several MeV are required for a number of important 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 critical facilities, andirradiation measurements in reactors, play an essential role in the qualification studies. Theinterpretation of experiments is a driving force for the continuous improvement of computational
methods and nuclear data.
Information on the activities and identified R&D needs by the NSC are provided.4 An expert
group addressing needs of research and test facilities in nuclear science has been set up and has held afirst meeting in May 2005. The report that is being prepared covering this topic will provide further
details on the issues discussed in the section.
3.1.4 Severe accidents
Introduction
Severe accidents (SA) are generally considered to be events beyond the traditional design basis ofcurrently operating nuclear power plants. The prevention or mitigation of SAs is the largest
contributor to reducing risk to the public from the operation of NPPs. SA scenarios involve an
initiating transient, such as a loss-of-coolant accident, accompanied by the postulated failures ofmultiple safety systems, thus compromising the capability to shut down the reactor or maintainadequate cooling of the fuel, resulting in significant damage to the fuel (core melting), possibly
leading to the release of significant amounts of radioactivity from the primary system into thecontainment. Under certain circumstances, the containment may also be postulated to fail or to be
bypassed (e.g. through steam generator tube failure in a PWR), resulting in a major radioactive releaseto the environment. Although generally not considered during initial licensing, SAs have beenassessed through specific plant reviews, generic analysis and the development of accident
management programmes.
For many years important national and international programmes have been undertaken in thefield of severe accidents and their results have been shared through international “networks.” CSNI
has played a major role in organising and administering co-operative research programmes in the areaof severe accidents. These programmes include RASPLAV (conducted in Russia to assess the thermalload on the RPV lower head under core melt conditions), SNL-LHF (conducted in the U.S. to assess
the mechanical behaviour of the RPV lower head under pressurised severe accident conditions), MCCI
(conducted in the U.S. to assess ex-vessel molten core debris coolability), MASCA (conducted in
Russia to measure the physical properties of molten core material) and SERENA (an analyticalprogramme assessing the state of knowledge related to fuel-coolant interactions). In addition, CSNI
sponsored an effort to assess accident management strategies and identify areas of consensus. The
effort (called senior Group of Experts on Severe Accident Management – SESAM) published its
report in the late 1990s. These programmes have contributed to the knowledge about severe accident
phenomena, the resolution of questions related to severe accidents and the potential for accident
management measures to successfully terminate or mitigate the accident progression. They have also
4.

NEA (2003), Research and Development Needs for Current and Future Nuclear Energy Systems, OECD,
Paris.

58

served to maintain certain key facilities from premature shutdown. However, important issues remain
and need to be studied to support the continued safe operation of nuclear power plants via severe
accident management and/or reducing the potential for SA scenerios, as well as supporting thelicensing of new LWR and PHWR designs.
Scope
The severe accident issues and phenomena that could benefit from additional research are related
to reducing the remaining uncertainties in accident progression and mitigation and to understanding
the safety implications caused by changes in plant design or operating characteristics (e.g. high burnup fuel, MOX fuel).
The issues in this section are listed in Table 3.1.4-1 and arranged according to the phases ofprogression of a severe accident and the phenomena present in each of those phases, as follows:
x
x
x
x
x

In-vessel phenomena (core heatup, clad/fuel melting and relocation, combustible gas
generation, FCI).
Ex-vessel phenomena (vessel failure, core-concrete interaction, DCH, FCI, combustible gas
generation).
Source term (quantity, chemical form, transport and timing of fission product release fromthe fuel, RCS and containment).
Containment integrity (capability of containment to withstand severe accident conditions
caused by combustible gas burning, decay heat, molten core attack).
Accident management (actions that can be taken to terminate or mitigate the consequences ofa severe accident).

Description
The prevention of severe (core damage) accidents and how to manage them if they do occur
remains an important objective for the continued safe operation of LWR and PHWR nuclear power
plants. Although in-vessel melt progression is fairly well understood, there remain significant
uncertainties in predicting whether or not molten core material will remain in-vessel, the consequencesof molten core material getting out of the reactor vessel (e.g., coolability, combustible gas, etc.), source
term generation and the best accident management strategies for preserving RPV and containment
integrity and reducing the amount of radioactive material available for release to the atmosphere.
Resolution of severe accident issues through prevention or mitigation is the goal of the remaining
research. This can be accomplished by design changes, analysis showing the issue is of low safetysignificance or developing strategies to terminate or mitigate severe accidents prior to their resulting in
the release of large quantities of radioactive material to the environment. To reduce uncertainties,
current research should be conducted at sufficient scale to investigate the important phenomena anduse real materials, whenever possible. PHWRs have similar severe accident issues as LWRs, however,the core melt progression in a pressure tube reactor presents additional challenges associated withpropagation of pressure tube failure, fuel-coolant or fuel-moderator interaction and the potential toover pressurise the calandria and cause calandria and additional pressure tube rupture.

59

Table 3.1.4-1. Current severe accident issues
Table 3.1.4-2. Issues versus facilities (Severe accidents)
Table 3.1.4-3. Facilities in the severe accident area60

Table 3.1.4-1. Current severe accident issues
Issues and relevant reactors

DescriptionA) In-vessel phenomena
1) Pre-core melt conditions: PWR,
BWR, VVER, PHWR, ALWR,
APHWR

2)

3)

4)

5)

6)

7)

8)

Understanding the conditions that can lead to core melt and thethermal-hydraulic conditions of the core prior to core melt are essential
to understanding whether or not implementation of accident management strategies will be successful in preventing core melt (e.g. hasflow blockage occurred?). Good knowledge of pre-core melt thermalhydraulic conditions in the core will also help to refine accident
management strategies so as to understand and be prepared for theoutcome of actions taken by the operator. This issue is closely coupledwith issue #17.
In-vessel melt progression:
The amount, composition, rate and timing of a core melt are important
PWR, BWR, VVER, PHWR,
to determining the effectiveness of accident management measures
ALWR, APHWR
and, the ability of the RPV or reactor calandria to maintain its
integrity. The type of fuel (UO2 or MOX), cladding material, burn-up
and other factors which affect the composition of the melt, are also
important in this determination. In-vessel melt progression includes
relocation in the core and to the lower portion of the RPV anddetermines the heat load on the RPV during a core melt accident.
Molten fuel contacting reactor coolant or moderator (PHWR,
In-vessel fuel-coolant
interaction: PWR, BWR, VVER, APHWR) may cause the rapid generation of steam and this is an
PHWR, ALWR, APHWR
important component of the load on the RPV or calandria.
Core melt accidents where air is present in the RPV (such as during
Effect of air on core melt
refuelling) could behave differently than those where no air is present.
progression: PWR, BWR,
This could include the dynamics of the melt progression and the FP
VVER, ALWR
release.
Effect of high burn-up and
The use of high burn-up or MOX fuel could change the dynamics of
MOX fuel: PWR, BWR,
melt progression and fission product release. Data on these effects isVVER, ALWR
needed to properly assess consequences and risk from accident
sequences involving high burn-up or MOX fuel.
RPV pressure: PWR, VVER
Depressurising the primary coolant system is important during the invessel melt progression phase to reduce stress on the RPV and to
facilitate water injection into the RPV. Accordingly, if the design does
not have the capability to depressurise the primary system. It is
important to understand the effect of high pressure on RPV and other
RCS components’ integrity and the subsequent effect on core melt
progression. This is primarily an analysis issue.
Maintaining the integrity of the RPV or reactor calandria vessel is
Maintaining RPV integrity:
important to terminating and confining a core melt accident, thus
PWR, BWR, VVER, PHWR,
eliminating ex-vessel severe accident phenomena and their challenge
ALWR, APHWR
to containment integrity. Cooling the RPV or reactor calandria bothinternally and/or externally are potential strategies for maintaining
RPV integrity in the event of a core melt accident. However, higher
core power densities will make it more difficult to maintain RPV
integrity due to the higher heat flux on the RPV. Knowledge of RPV
integrity as a function of heat flux is important in assessing the successof accident management strategies.
Pressure tube integrity: PHWR, Maintaining the integrity of the pressure tubes in a pressure tube
APHWR
reactor is important for maintaining cooling of the fuel in the tube andpreventing over-pressurisation and failure of the calandria due to highpressure water injection and/or molten fuel injection and FCI.

61

Table 3.1.4-1. Current severe accident issues (Cont’d)Issues and relevant reactors

DescriptionB) Ex-vessel phenomena
9) Ex-vessel melt progression anddebris coolability: PWR, BWR,
VVER, PHWR, ALWR,
APHWR

The amount, rate, timing and spreading of molten core material
released following RPV failure are important to determining the abilityof the concrete basemat to maintain its integrity and the ability of an
overlying pool of water or basemat cooling system to cool the debris
and terminate the core-concrete reaction (i.e. ex-vessel melt
coolability). Debris coolability can be affected by the amount of water
overlying the core debris and the porosity of the debris or the strengthof the crust formed on top of molten core debris. Obtaining theproperties of the crust and underlying debris is important tounderstanding debris coolability and its uncertainties. In addition, highpressure melt ejection could result in molten core material being
relocated to other parts of containment and a rapid pressure rise incontainment due to the sudden release of steam and combustible gases
from the RPV to containment.
10) Core-concrete interaction: PWR, When molten core material leaves the reactor pressure vessel, it will
likely come in contact with concrete. Depending upon the amount and
BWR, VVER, ALWR, PHWR,
depth of the molten core material and the composition of the concrete,
APHWR
various amounts of combustible and non-combustible gas will be
released into the containment, thus raising its pressure. These gases
can also be a source of additional energy if they ignite, thus causing
additional pressure and temperature rise in containment. If not stopped,
the core concrete interaction can potentially also penetrate the reactor
containment basemat, thus failing containment. Understanding the rate
and amount of gas generated from core-concrete interactions isimportant to understanding the potential for containment failure, thepotential for success of mitigation strategies and, in the case of newplant designs, selecting materials and configurations to minimise coreconcrete interactions.
11) Ex-vessel fuel coolant interaction: Upon failure of the reactor pressure vessel, molten core material may
PWR, BWR, VVER, PHWR,
fall or be ejected into water, if the reactor cavity has been partially or
ALWR, APHWR
fully flooded. Such contact with water has the potential to cause rapid
steam generation and, depending upon the amount, rate, fragmentationand mixing of the molten material, release a large amount of energy
should be taken into account in assessing, which structural integrity ofcontainment.
12) Combustible gas control: PWR,
Combustible gas (H2 and CO) generated from metal-water reactions or
BWR, VVER, PHWR, ALWR,
core- concrete reactions in the in-vessel and ex-vessel phases of a core
APHWR
melt accident can ignite heat and/or pressurise containment, thus
challenging containment integrity.

62

Table 3.1.4-1. Current severe accident issues (Cont’d)Issues and relevant reactors

DescriptionC) Source term13) Fission product chemistry andrelease: PWR, BWR, VVER,
PHWR, ALWR, APHWR

The amount, composition, chemistry, and timing of fission productrelease from the fuel through the reactor coolant system and into
containment results in 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 isalso affected by the type of fuel (UO2 or MOX) and burn-up level. In
addition, the effectiveness of source term attenuation measures (e.g.
sprays, water chemistry, filters) needs to be understood.
14) Post containment failure FP
Containment failure can lead to additional FP release due torelease to the environment: PWR, revolitisation under depressurised conditions and/or due to air ingress.
BWR, VVER, PHWR, APHWR, Understanding these phenomena is necessary for predicting the
ALWR
consequences and risk from accident sequences that fail anddepressurise the containment.
D) Containment integrity15) Containment integrity: PWR,
Understanding the conditions which could lead to containment failure
BWR, VVER, PHWR, APHWR, or bypass (e.g., pressure, temperature, and equipment failure) is
ALWR
important. Therefore, knowledge of the integrated effects of designbasis and severe accident loads is necessary input to containment
design. This issue provides input for structural analysis andcontainment failure modes testing (see Section 3.1.5 on integrity ofequipment and structures). This is primarily an analysis issue.
16) Containment bypass-overheating During core heatup under severe accident conditions, significant
and failing steam generator tubes: amounts of heat are transferred by natural circulation of steam and
PWR, ALWR
hydrogen to the hot legs, surge line, and steam generators. This may
cause one or more steam generator tubes to fail prior to failure of the
hot leg, surge line, or vessel lower head. Understanding the behaviourof steam generator tubes is important to preventing containment bypass scenarios. This is primarily an analysis issue.
E) Accident management17) Coolability of overheated core:
Accident management (which deals with the plant when core
PWR, BWR, VVER, PHWR,
conditions are in excess of peak cladding temperature) has the
APHWR, ALWR
potential to prevent as well as mitigate core melt accidents. In thisregard, to develop and verify core melt prevention strategies, thesymptoms, timing, and consequences of various strategies need to beunderstood to assess their effectiveness and enable the operator to
anticipate the resultant consequences. This issue is closely coupledwith issue #1. The effects of adding water to overheated cores stillneeds experimental verification.
18) Accident management strategies: Once core melt begins, operator actions should be taken to terminate
the accident as soon as possible. Therefore, it is important to
PWR, BWR, VVER, PHWR,
understand the symptoms and implications of the various AM actions
PHWR, ALWR
in order to develop the optimum AM strategies and actions. These
actions can include depressurising the primary coolant system,
injecting water, actuating containment sprays, etc. Specific AM actions
depend upon the reactor design and accident scenarios. Analysing thepotential for success and potential consequences of the various AM
strategies/actions would rely on data developed in the resolution ofmany of the issues stated above. Relevant facilities are primarily thelarge integral thermal-hydraulic facilities discussed in Section 3.1.1.1.

63

Table 3.1.4-2. Issues versus facilities (Severe accidents)

Issue

Applicabilityof issue

Safety relevance State of knowledgeof issueon issue1) Pre-core melt conditions.

PWR, BWR, VVER, PHWR,
ALWR, APHWR.

High

High

2) In-vessel melt progression.

PWR, BWR, VVER, PHWR,
ALWR, APHWR.

High

Medium3) In-vessel fuel coolant interaction.

PWR, BWR, VVER, PHWR,
ALWR, APHWR.

MediumMedium4) Effect of air
on core-melt progression.

PWR, BWR,VVER, ALWR

Low

Medium5) Effect of high burn-up and
MOX fuel.

PWR, BWR, VVER, ALWR.

MediumMedium

7) Maintaining RPV integrity.

PWR, BWR, VVER, PHWR,
ALWR, APHWR.

High

Medium8) Pressure tube integrity.
9) Ex-vessel melt progression anddebris coolability.

PHWR, APHWR.
PWR, BWR, VVER, ALWR,
PHWR, APHWR.

MediumHigh

MediumMedium10) Core-concrete interaction.

PWR, BWR, VVER, ALWR,
PHWR, APHWR.

High

MediumMediumMedium11) Ex-vessel fuel coolant interaction. BWR, PWR, VVER, PHWR,
ALWR, APHWR.

64

Table 3.1.4-2. Issues versus facilities (Severe accidents)
FacilityName
PHEBUS

Core disassembly test facility
Fuel channel safety facility
QUENCHPHEBUS
LIVE-FZKQUENCHKROTOSMFMI
TROI
PHEBUS
VERDONPHEBUS
VERDON (LECA-STAR)
MASCA/RASPLAV
LIVE-FZK
Fuel ch. safety facility
MCCI
VULCANOARTEMISMCCICOMET-FZK
VULCANOARTEMISKROTOSMFMI
TROI

Importance ofVersatility
facility to resolutionof the issue?
High
Can be used for various small bundle tests up toand beyond melting and tests on coolability ofover-heated core.
High
Can be used to assess PHWR fuel bundle
behaviour up to melting.
High
Can be used to assess integrity of PHWR fuelchannel.
High
Can be used to assess effectiveness of coremelt prevention strategies.
High
Capable of small bundle in-core melt tests.
MediumUses simulant material.
MediumCan assess partially degraded core.
High
Capable of using prototypic materials.
High
High and low pressure PHWR melt ejection.
MediumUses prototypic materials (20 kg).
High
Capable of small bundle in-core melt tests withair.
MediumCan conduct hot cell experiment with irradiatedfuel or air.
High
Can conduct in-reactor experiments with highburn-up or MOX fuel.
MediumCan conduct hot cell experiments with
irradiated fuel.
High
Can obtain material properties using realmaterials.
MediumUses simulant material.
High
Fuel channel thermal- mechanical behaviour.
High
Can use prototypic materials.
High
Can use prototypic materials.
MediumUses simulant materials.
High
Large-scale test (1 m2) with real materials,
simulated decay heat and with or w/o overlaying
water cooling.
MediumUses simulant material.
High
Uses prototypic materials (oxide and metal).
MediumUses simulant material.
High
Can test with real materials.
MediumFacility for PHWR test configurations Uses
simulant material.
MediumUses prototypic materials.

65

Table 3.1.4-2. Issues versus facilities (Severe accidents) (Cont’d)

Issue

Applicabilityof issue

Safety relevance State of knowledgeof issueon issue12) Combustible gas control

PWR, BWR, VVER, PHWR,
ALWR, APHWR

High

Medium13) Fission product release

PWR, BWR, VVER, PHWR,
ALWR, APHWR

High

Medium14) Post containment failure-FP
release to the integrity
environment15) Containment

BWR, PWR, VVER, PHWR,
APHWR,ALWR
PWR, BWR,
VVER, PHWR,
ALWR, APHWR

MediumHigh

Low
Medium16) Cont bypass overheating andfailing S.G. tubes17) Coolabillity of overheated core

PWR, ALWRHigh

Medium

PWR, BWR, VVER, PHWR,
APHWR, ALWR

High

High

18) Accident management
strategies

PWR, BWR, VVER, PHWR,
ALWR, APHWR

High

Medium66

Table 3.1.4-2. Issues versus facilities (Severe accidents) (Cont’d)
Facility

CTF
LSVCTF
PANDA

Importance offacility to resolutionof the issue?
High
High
MediumMISTRA

Medium

H2-TecRUT
THAI

MediumHigh
High

TOSQANPHEBUS
VERDONARTIST

Low
High
High
High

THAI
EPICUR
MAESTRO
CHIP
THAI
PANDA

High
MediumHigh
MediumHigh
Medium for current
LWR. High forpassive ALWRs
MediumHigh
MediumName

MISTRA
LSCF
THAI
No facility identified. This is
primarily an analysis issue
RD-14MQUENCHPHEBUS
Uses data generated in the resolutionof other SA and T/H issues. Facility
needs primarily filled by otherfacilities, such as the large integral
thermal-hydraulic facilities.

High

Versatility
Separate effects combustion studies.
Large-scale combustion studies.
Large-scale-multi compartment capability for
3-D mixing.
Large-scale tests in multi-compartment
configuration for mixing and distribution of H2.
No compartments.
Large-scale tests in multi-compartment.
Addresses H2, I and aerosols and their combined
effects. Instrumentation can measure aerosol
distribution and movement. Uses H2.
Small scale, uses He as a simulant of H2.
Could be used for other tests.
FP behaviour in containment.
Unique S.G. configuration: bundle, separator
and dryer.
FP behaviour in containment.
I chemistry under SA conditions.
Can test effect of B4C and air.
I chemistry in RCS.
FP resuspension.
Large-scale tests in multi-comportment
configurations for decay heat removal fromcontainment.
Detailed 3-D instrumentation.
1 700 m3 containment T/H facility.
Capability for testing with H2.

Full height, full pressure PHWR T/Hexperiments.
Can test core melt prevention scenerios.
Can simulate various SA conditions.

High
High

67

Table 3.1.4-3. Facilities in the severe accident area
Applicability(type of reactor)
PWR, BWR, VVER, PHWR,
ALWR.
PWR,
BWR, VVER, PHWR,
ALWR.
PWR, BWR, VVER, ALWR,
PHWR, APHWR.

Cost/year,
operation
Low
MediumReplacement
cost
High
High

MediumHigh

ARTIST (Switzerland)
MISTRA (France)

PWR, PHWR, ALWR, APHWR.
PWR, BWR, VVER, ALWR.

MediumMediumMediumHigh

RUT (Russia)

PWR, BWR, VVER ALWR,
APHWR.
PWR, BWR, VVER, ALWR,
APHWR.

Low

High

MediumHigh

MediumMediumMediumMediumFacility name
KROTOS (France)
MCCI (MACE) (USA)MASCA and RASPLAV (Russia)

PANDA (Switzerland)*TROI (Korea)
Hs-Tec (Germany)

PWR, BWR, ALWR, VVER.
BWR, PWR, ALWR, VVER.

CTF containment test facility (Canada)

ALL.

Low

Medium

LSVCTF large-scale vented
combustion test facility (Canada)
LSCF-large-scale containment facility(Canada)Core disassembly test facility (Canada)ALL.

MediumMediumALL.

Low

Medium

PHWR, APHWR.

Low

Medium

PHWR, APHWR.

MediumMediumHigh
MediumMediumMediumMediumMediumHigh
MediumHigh
MediumMediumMediumVULCANO (France)

PHWR, APHWR.
PHWR, APHWR.
PWR, BWR, VVER, ALWR.
PWR, BWR, VVER, ALWR.
PWR, BWR, VVER, ALWR.
PWR, BWR,VER, PHWR, ALWR,
APHWR.
PWR, BWR, VVER, ALWR.

MediumHigh

QUENCH, FZK (Germany)
COMET-FZK (Germany)

PWR, BWR, ALWR, VVER.
PWR, BWR, ALWR, VVER.

MediumLow

MediumMediumTOSQAN (France)

PWR, BWR, VVER PHWR, ALWR,
APHWR.
PWR, BWR, VVER PHWR, ALWR,
APHWR.

MediumMediumLow

MediumMFMI (molten fuel moderator
interaction) test facility (Canada)
RD-14M* (Canada)
Fuel channel safety facility (Canada)
VERDON (LECA-STAR)(France)
EPICUR (France)
CHIP (France)
MAESTRO (France)

ARTEMIS (France)68

Table 3.1.4-3. Facilities in the severe accident area
IssuesPlanned duration Relative
Capability
coveredof operation
ranking
3,11
FCI tests using prototypic materials.
Through 2009.
0.7
9,10
Large-scale ex-vessel core concrete interaction and cooling using Through 2009.
1.2
prototypic materials.
7
RASPLAV is a one-tenth scale (slab-geometry) facility to model Through 2010.
0.6
RPV and molten core (using real materials). MASCA measuresmaterial properties with prototypic materials in small facilities.
13
12,15

Retention of aerosols and fission products in a PWR steam generator. Through 2007.
100 m3 PWR containment. (0.1 linear scale) using He as a simulant. In danger of
Flexible free and compartmented volumes. Capability for steam and shutdown in 2007.
gas injection. Spray system, 3-D instrumentation.

0.60
0.7

12

100-1 600 m3 facility, with compartments for combustion experiment Shutdown in 2005.
with or without steam
Flexible large-scale facility for 3-D effects, multi-compartment Until the end of thecontainment behaviour, containment mixing and stratification studies SETH project inand passive heat removal, Integral system behaviour. Has extensive 2006.
3-D instrumentation.

0.6

0.4
0.4

1

Can test ex-vessel SE with up to 20 kg of prototypic material.
Through 2015.
Vessel for deflagration and detonation and recombiner testing. Through 2012.
(100 bar capability).
10 Mpa – Hydrogen combustion tests (6 m3 and 10 m3 connected To be put invessels). Heated.
standby in 2007.
Hydrogen combustion large-scale vented facility (120 m3) with To be put inheating, can be subdivided into 2-3 compartments.
standby in 2007.
3
1 625 m facility for gas mixing and thermal hydraulics testing (no Indefinite.
combustion testing). Uses He as a simulant.
Facility for testing fuel bundle behaviour up to fuel melting.
Indefinite.

3,11

Facility to inject molten fuel simulant material into moderator fluid.

Through 2008.

0.4

17
1,8
4,5,1313
13
13

Full height, full pressure PHWR T/H experiment.
Facility to test the integrity of fuel channels.
FP release from irradiated fuel heated up to 3 000qK.
Iodine chemistry under high radiation and high temperature.
Source term: iodine chemistry in primary system.
Source term under air oxidation, B4C degradation, fuel degradation.

2010.
Indefinite.
Indefinite.
Through 2009.
Through 2010.
Through 2010.

0.3
0.7
0.9
0.4
0.4
0.6

9,10

Core-concrete interaction test and debris coolability tests using Through 2007.
prototypic materials.
Reflood of over-heated core.
Through 2007.
Core-concrete interaction using simulant corium and decay heat Through 2006.
simulation. Can add water.
7 m3 f facility to simulate H2 (using He) mixing under SA conditions, Through 2008.
including steam, sprays and aerosols.
Debris coolability and core-concrete interaction using simulant Through 2008.
material (molten salt).

1.2

12,15

3,11
12,
12
12
15

1,2,17
10
12
9,10

69

1.0

0.6
0.6
0.6
0.3

1.0
0.4
0.2
0.7

Table 3.1.4-3. Facilities in the severe accident area (Cont’d)Applicability(type of reactor)
PWR, BWR, VVER, PHWR, ALWR,
APHWR.

Cost/year,
operationHigh

Replacement
cost
High

THAI (Germany)

PWR, BWR, VVER, ALWR,
APHWR.

MediumMediumLIVE-FZK (Germany)

PWR, VVER, BWR, ALWR.

MediumMediumFacility name
PHEBUS (France)

Notes:
Specify range: Low, Medium, High.
Operation cost
: Low is <1.0 MUS$/y; Medium is 1.0-2 MUS$/y; High is >2 MUS$/y.
Replacement cost : Low is <2 MUS$; Medium is 2-10 MUS$; High is >10 MUS$.
* Also discussed in thermal-hydraulic section.

70

Table 3.1.4-3. Facilities in the severe accident area (Cont’d)IssuesPlanned duration Relative
Capability
coveredof operation
ranking
1.2,4.5, A 20 MWt pool type reactor, with a large central cell, connected to a In danger of
2.3
13,17
Fission Products Bldg. (1/5000 scale containment) for conducting in- shutdown in 2007.
reactor fuel melt tests. Facililty is also capable of conducting othertypes of tests (e.g. thermal-hydraulic).
12,13,1 60 m3 facility can be used to test hydrogen burning, affects of Through 2009.
4, 15
containment sprays on mixing, iodine and aerosols. FP resuspension
and recombiner performance.

2.2

2,7

0.7

Test on formation of molten pools and relocation. 1/5 scale using Through 2012.
simulant corium.

3.1.5 Integrity of equipment and structures
Introduction
Many of the current problems with operating reactors are related to materials issues. While plants
were initially designed for a 40 or more year lifetime, a number of unanticipated material 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 will continue to need to beinvestigated and solved. Accordingly ensuring the condition of equipment and structures is monitored
and known becomes increasingly important. The ageing effects can include cracking, corrosion,
erosion, cable insulation cracking, fatigue, embrittlement, etc., and can affect most plant equipment
and structures. Identifying, monitoring and controlling the ageing effects are important to continued
safe plant operation.
Scope
The issues addressed in this section are related to identifying the phenomena that are causing the
problems, improving techniques for detecting and repairing problems and anticipating and preventing
future problems before they become safety issues. The safety issues that could benefit from additional
research are those that are associated with the ageing of existing plants and those that are associatedwith initiatives to improve plant performance or develop new designs and are listed in Table 3.1.5-1.
Description
The timely detection and mitigation of ageing degradation of plant systems, structures andcomponents (SSC) are important to safety, so as to ensure their integrity and functional capability
throughout plant service life. It means that managing the safety aspects of NPP ageing requires theimplementation of effective programmes. General guidance for NPP activities which are relevant to
the management of ageing, i.e. operation, inspection, testing, examination, maintenance andsurveillance, are given in the frame of the International Atomic Energy Agency (IAEA) Nuclear
Safety Standards (NUSSs) Code on the Safety of NPP. The largest activity in this area is related to theintegrity of the primary pressure boundary, including reactor vessel and the associated piping. The
integrity of the secondary side components is included also. The extension of the operating lifetime of
NPPs is also important from the point of view of ageing.

71Research programmes need to focus on the ways in which the reactor environment and operating
conditions degrade the strength and integrity of equipment and structures over their operational
lifetime. Therefore in the area of integrity of equipment and structures the research should address:x

The state of the material of the equipment or structures which may be affected by material
composition, manufacturing processes or operational parameters.

x

The loads imposed on the equipment and structures during operation (normal and transient
operation, incidents, accidents and events) which are combined with the initial state of thestress and environmental conditions.

x

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

x

The safety margins available in the design.

x

The sensitivity of the examination and testing methods applied.

Although the research needed to address the issues in Table 3.1.5-1 involves reactors, hot cells,
autoclaves and, in some cases, other facilities, the focus of this section is on the reactors and otherunique facilities needed. Due to the large number of hot cells and autoclaves and the fact that mostmember countries have those facilities, the SESAR group decided to not make specificrecommendations on these facilities, but rather to recommend that each member country monitor the
status of these facilities and bring to CSNI’s attention any concerns regarding loss of criticalinfrastructure. Key hot cells and autoclaves are listed in Table 3.1.5-4.
Table 3.1.5-1. Current integrity of equipment and structures issues
A) Plant ageing issues/and relevant reactors
1) Erosion/corrosion:
PWR, BWR, VVER, PHWR

As plants age, environmental conditions can cause some materials
to corrode or erode more than was originally anticipated in thedesign. This can lead to cracks, leaks or even large failures. The
causes and corrective actions for these conditions needs to beunderstood and implemented to support continued operation
and/or plant life extension. The corrosion can be both internal tothe component or external due to leakage.

2) Embrittlement:
PWR, BWR, VVER

Embrittlement of steels, particularly reactor pressure vessel steel,due to exposure to fast neutrons can reduce their ability towithstand thermal and mechanical stresses. The properties ofembrittled materials and the ability to predict the amount of theirembrittlement need confirmation by experimental data. Also, theeffectiveness of any corrective actions needs confirmation.

3) Cracking and crack propagation:
PWR, BWR, VVER PHWR

Cracking of materials and crack propagation (both steels andconcrete) has caused problems at operating reactors. Cracking andcrack propagation may be due to environmental conditions, fatigue
or poor design. The causes and corrective action for cracking andcrack propagation need experimental confirmation.

4) In-service inspection:
PWR, BWR, VVER, PHWR

Inspection techniques to look for cracks, erosion or other ageing
effects is important for early detection and correction, before theylead to a safety concern. Testing and validation of inspection
techniques is essential.

72

Table 3.1.5-1. Current integrity of equipment and structures issues (Cont’d)A) Plant ageing issues/and relevant reactors5) Cable insulation cracking:
PWR, BWR, VVER, PHWR

Cable insulation (both power and I and C cables) can crack andbecome brittle over time. Environmental conditions affect the rate
at which this happens. This can lead to shorts, fires or unexpectedbehaviour under conditions of high moisture (e.g., coolant leak,fire suppression). Detection and correction techniques need to beverified.

6) Pressure tube integrity:
PHWR.

Corrosion and irradiation of PHWR pressure tubes during reactor
operation can change material properties, in some cases making
them more susceptible to failure. This could lead to random
failures or common cause failures (e.g. due to seismic events),
which in either case, could pressurise the calliandra and lead tomore severe damage.
Understanding the PHWR pressure tube condition, limits andfailure modes is important to safety.

7) Long-term behaviour of concrete
structures:
PWR, BWR, VVER, PHWR

As plants age, concrete properties change and/or cracks develop.
The safety implications of concrete ageing need to be understoodto support the continued safe operation of existing plants and thereview of requests to extend plant lifetime.

8) Containment integrity:
PWR, BWR, VVER, PHWR

The conditions under which containments fail, and the timing andmodes of failure, are important to understand to assess safetymargins, consequences (i.e. FP release) and risk. Therefore, thestructural analysis methods need experimental confirmation due tothe complex nature of containment designs and penetrations.
Currently, this issue is primarily an analysis issue using previous
experimental data to assess analytical methods.

9) Flow induced vibrations:
BWR, PWR

As current plants pursue power increases, the flow distributions,particularly in-vessel, and their contribution to mechanical loads
and vibration of equipment needs to be understood. Predicting
such flow distributions needs experimental data to validate
analytical tools.

B) Performance improvement/new design issues10) New materials - existing plants:
PWR, BWR, VVER, PHWR

To respond to materials problems (e.g. cracking, corrosion) on
existing plants, new materials are being used in replacement
components. The performance of these materials needs to beunderstood.

11) New materials:
ALWR, APHWR

Future designs will try to use materials less susceptible to the
problems occurring on existing plants. Qualification of thesematerials needs to be addressed.

73

Table 3.1.5-2. Issues versus facilities (Integrity of equipment and structures)

Issue

Applicabilityof issue

Safetyrelevanceof issueStateof knowledge
on issueMediumMediumHigh

Medium1)

Erosion / Corrosion

PWR, BWR, VVER, PHWR

2)

Embrittlement

PWR, BWR, VVER

3)

Cracking and crack
propagation

PWR, BWR, VVER, PHWR

MediumMedium4)

In-service inspection

PWR, BWR, VVER, PHWR

MediumMedium5)
6)

Cable insulation cracking
Pressure tube integrity

PWR, BWR, VVER, PHWR
PHWR

High
High

Low
MediumLong-term behaviour ofconcrete structures
8) Containment integrity9) Flow induced vibrations
10) New materials.
Existing plants

PWR, BWR, VVER, PHWR

High

Medium

PWR, BWR, VVER, PHWR
PWR, BWR
PWR, BWR, VVER, PHWR

High
Low
MediumMediumMediumMedium11) New materials.
New designs

ALWR, APHWR

High

Low

7)

74

Table 3.1.5-2. Issues versus facilities (Integrity of equipment and structures)
FacilityName
Hot Cells*
Autoclaves*
LVR-15
HaldenATR
LVR-15JMTR
Hot Cells*
HaldenATR
LVR-15JMTR
Hot Cells*
Autoclaves*
IQ
EPRI-N.C.
PANOZANRU
Hot Cells*
No applicable facilities identified
Primarily an analysis issue
No applicable facilities identified
LVR-15JMTR
Hot Cells*
Autoclaves*
SKODA
Halden
NRU
ATR
SKODA
Autoclaves*
Hot Cells*
LVR-15JMTR
HaldenNRU
ATR

Importance offacility to resolutionof the issue?
MediumHigh
MediumHigh
High
MediumMediumHigh
High
Medium
High
High
High
High
MediumMediumMediumHigh
High

There is a large number with varying capability.
There is a large number with varying capability.
Low power.
Many uses and versatile instrumentation.
High flux capability.
Low power.
Medium power.
There is a large number with varying capability.
Many uses and versatile instrumentation.
High flux capability.
Various material testing loops (BWR, PWR).
Medium power.
There is a large number with varying capability.
There is a large number with varying capability.
Corrosion measurement.
Many capabilities for ISI testing.
Gamma irradiation facility.
PHWR materials irradiation.
There is a large number with varying capabilities.

High
High
High
High
Medium
High
High
High
Medium
High
High
High
High
High
High
High

Various materials testing loops (BWR, PWR).
Medium power.
There is a large number with varying capabilities.
There is a large number with varying capabilities.
Mechanical testing.
Versatile uses and instrumentation.
High power capability.
High flux capability.
Mechanical testing.
There is a large number with varying capabilities.
There is a large number with varying capabilities.
Has various material testing loops.
Medium power.
Versatile uses and instrumentation.
High power capability.
High flux capability.

Versatility

Note:* In assessing issues versus hot cells and autoclaves, only a generic entry is included in Table 3.1.5-2, due tothe large number of such facilities and the general similarity of their importance to issue resolution. Eachcountry should monitor the status of their hot cells and autoclaves and bring to CSNI’s attention any
concerns regarding loss of critical infrastructure. Key hot cells and autoclaves are listed in Table 3.1.5-4 for
information.

75

Table 3.1.5-3. Facilities in the area of integrity of equipment and structures
Applicability(type of reactor)

Facility name
Reactors
NRU (Canada)
Halden Reactor (Norway)
LVR-15 (Czech Rep.)
JMTR (Japan)

Cost/year, Replacement
operation
cost

PHWR
BWR, PWR, PHWR, ALWR

High
High

High
High

All types
PWR, BWR

High
High

High
High

ATR (USA)
PWR, BWR, ALWRHigh
Other facilities
Cobalt irradiation units
PWR, BWR, VVERLow
PANOZA, PRAZDROJ (Czech. Rep.)
SKODA (Czech. Rep.) ZZ800 Robertion
BWR, PWR, VVER
Low
Mechanical testing machine
Inspection qualification (IQ) (Czech. Rep.)
VVER
Low
EPRI-North Carolina (USA)
PWR, BWR
MediumNotes:
Specify range: Low, Medium, High.
Operation cost
: Low is <1.0 MUS$/y; Medium is 1.0-2 MUS$/y; High is >2 MUS$/y.
Replacement cost : Low is <2 MUS$; Medium is 2-10 MUS$; High is >10 MUS$.

High
MediumHigh
MediumHigh

Table 3.1.5-4. Key hot cells and autoclaves
Hot cell facilities*
PSI-hot cells (Switzerland)
FHL (NDC) and NFD hot cells (Japan)
IMEF (Korea)
Chalk River Lab (Canada)
RIAR (KI) (Russia)
Argonne National * Laboratory (USA)Oak Ridge National laboratory (USA)Hot cells (Czech. Rep.)
Idaho National Lab. (USA)LECI (France)
EC-JRC (Germany)
VTT – Hot Cell (Finland)

Cost/year, operation
High
High
High
High
High
High
High
High
High
High
High
MediumReplacement cost
High
High
High
High
High
High
High
High
High
High
High
High

Autoclave facilities
PSI autoclaves (Switzerland)
Chalk River Lab. (Canada)
LECSI (France)
VTT Autoclaves (Finland)
Pacific Northwest National Laboratory (USA)

Cost/year, operation
MediumHigh
MediumMediumHigh

Replacement cost
High
High
High
High
High

Note:
* Hot cells also discussed in fuel section (3.1.2)

76

Table 3.1.5-3. Facilities in the area of integrity of equipment and structures
IssuescoveredPlanned durationof operation

Capability

135 MWt test reactor.
19 MWt test reactor capable of IASCC, corrosion /hydridling testing.
1,2,3,10,11 Materials test reactor (10 MWt).
2,3,10,11 Materials test reactor (50 MWt).

Relativeranking

6, 10, 11
2,3,10,11

Through 2010
Indefinite

2.0
2.3
2.3
2.1

2,3,0,11250 MWt test reactor.

Through 2018
Through 2008 (under discussion
to extend beyond 2008)
Indefinite

5

Gamma irradiation facility, including cable ageing tests. Indefinite

0.6

10,11

Mechanical testing laboratory.

Indefinite

0.8

4
4

Corrosion measurements.
ISI qualification and training facility.

Thru 2010
Indefinite

0.2
0.2

2.2

Table 3.1.5-4. Key hot cells and autoclaves
Issues covered
1,2,3,6,10,11
1,3,6,10
1,3,6,10
1,3,6,10,11
1,3,6,10,11
2,3,6,10,11
2,3,6,10,11
2,3,4,10,11
2,3,6,10,11
2,3,6,10,11
2,3,6,10,11
2,3,10,11
Issues covered
1,10,11
1,3,10,l11
1,3,10,11
1,3,10,11
1,3,10,11

Capability
Hot cells with diagnostic equipment.

Hot cells with material and fracture testing capability.
Hot cells for examination of small samples of irradiated materials.
Hot cells for examination of large samples of irradiated materials.
Examination of irradiated structural materials.
Can handle large items.

Fracture toughness radiation embrittlement, mechanical properties and microstructuralcharacteristics.
Issues covered
Autoclaves for irradiated materials testing.
Autoclaves, with corrosion loop.
Autoclaves for irradiated materials testing.
Corrosion, fatigue, stress corrosion cracking testing.
Examination of non-irradiated materials.

773.2 Issues and facilities not unique to the nuclear industry
3.2.1 Human and organisational factors
Introduction
The importance of the human and organisational factors in nuclear reactor safety has beenrecognised for a long time. These are issues that transcend engineering and involve social,
psychological and other non-engineering factors and disciplines; for this reason, they are difficult toquantify and analyse. It is important, however, to address these issues in order to assess the root causesof human and organisational performance problems and attempt to remedy these by appropriate
design, procedures, training, etc. It is also important to assess the degradation of safety related to suchcauses in order to establish the degree of severity of the problems and prioritise remedial actions.
Scope
Human and organisational issues apply to both currently operating and future plants. For
currently operating plants, design solutions may be limited, whereas for future plants consideration ofhuman factors early in the design process can contribute to good human performance. The human andorganisational factors safety issues that could benefit from additional research are shown in
Table 3.2.1-1.
Description
Currently operating plants
The assessment of operating experience can identify many of the root causes of human andorganisational performance in operating plants. Solutions can then be identified. Also, upgrading I&C
systems can lead to new human factors issues.
New reactor designs
New reactor systems may have different human interfaces, may respond in different ways toreactor operator actions. The continuing introduction of digital and computer controlled systems andadvanced instrumentation also introduces some new human man-machine interface questions.
Organisational factors
Organisations, social systems and priorities and ways of operating within organisations are also
evolving; an example is the liberalisation of the electricity markets in most countries. Globalisation of
corporate activities and increased interactions with different cultures also introduce new organisational
behaviour patterns, structures, etc. Such factors need to be considered and it is necessary to assess
their impact on reactor safety.

78

Table 3.2.1-1. Current human and organisational factors issues
Issues and relevant

Description

1) Staffing
ALWR, APHWR, HTGR

New designs are incorporating passive safety features and employing more
automation. They are also being designed with longer response times.
Accordingly, the role of the operator and the number of operational staff ischanging. How to decide on correct staffing levels remains an issue. In
addition the analysis of new tasks and the qualification of the staff are to beconsidered. Also, in both, new and existing plants, the effect of staffing cuts
should be investigated. Many or most plants have cut staff as a result ofmarket liberalisation and deregulation.
As more plants upgrade or introduce advanced instrumentation, issues related2) Human-machine interface:
to how humans interface with the system must be addressed. This includes
PWR, BWR, VVER,
issues such as:
PHWR, ALWR, APHWR,
HTGR
x Role of the human vs. automation.
x Navigation through software controlled displays.
x Inputting commands.
Testing of new human-machine interfaces will be useful. This should include
verification and integrated system tests.
3) Organisational Influences:
Organisational factors, market liberalisation, and deregulation can influence
PWR, BWR, VVER,
the performance of socio-technical systems and have negative effects on
PHWR, ALWR, APHWR,
safety, factors such as communication, organisational learning, safety
HTGR
culture, etc., influence employee’s knowledge and behaviour and have been
fundamental factors in actual accidents. Although facilities are not relevant
for this issue, understanding, monitoring and addressing these factors canimprove safety. Accordingly, developing and maintaining expertise is of key
concern. Objective measures of organisational performance would also be
useful.
4) Human performance model: Developing reliable models of human performance will greatly enhance theaccuracy of risk assessments and the ability to evaluate human related issues
PWR, BWR, VVER,(e.g., procedures, training). Expertise, rather than facilities is the critical
PHWR, ALWR, APHWR,
need.
HTGR
Many events at operating reactors have as their initiator or as an important
5) Review of operating
element in the event human and/or organisational factor contributors. The
experience: PWR, BWR,
review of operating experience to identify and correct those human and
VVER, PHWR
organisational contributors is key to maintaining/improving safety.

Table 3.2.1-2. Facilities in the area of human and organisational factors
Facility nameCost/year,
operation

Replacement
cost

IssuescoveredCapability

Plannedduration ofoperation

Halden Reactor
Project

High

High

1, 2

Ability to simulate control
room environment and conduct
with plant operators

Indefinite

Notes:
Specify range: Low, Medium, High.
Operation cost
: Low is <1.0 MUS$/y; Medium is 1.0-2 MUS$/y; High is >2 MUS$/y.
Replacement cost : Low is <2 MUS$; Medium is 2-10 MUS$; High is >10 MUS$.

79

3.2.2 Plant control and monitoring
Introduction
All nuclear power reactors require plant control, and monitoring and protection systems(commonly referred to as instrumentation and control (I&C) systems), and there is a growing use ofdigital instrumentation and control 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, andthat replacement systems meet reliability goals. For the former, the challenges are to ensure that
degradation mechanisms are understood and mitigated, and systems are appropriately qualified for postaccident 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 that regard, the increasing use of digital
I&C presents both opportunities and challenges. The primary opportunity is to replace systems with new
equipment with enhanced functionality. The challenges are to ensure that the new systems perform withequal 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 useof 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.
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):xxx

Safety (or protection) systems, primarily responsible for mitigating against consequences offailure 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.

The plant control, and monitoring and protection system safety issues that could benefit fromadditional research are listed in Table 3.2.2-1.
Description
The challenges are to ensure that the new systems, primarily digital 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 theuse 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.
x

LWRs (including VVERs). LWRs traditionally relied on analogue systems for plant safety
and control. As these systems age, it is important to demonstrate that they continue to meettheir original performance specifications, particularly when exposed to harsh environmental
conditions. When the current systems reach the end of their design life, they are increasingly
being replaced with digital systems. These digital systems must be shown to meet the same
performance specifications, including qualification for expected environments. This will
require demonstration of hardware and software reliability. In addition, increasing use isbeing made of advanced on-line monitoring and diagnostic systems to help manage the plant
more reliably. The complex failure modes for these systems need to be investigated.
80

x

ALWRs. The issues for LWRs are magnified for ALWRs. Advanced reactor designs will
make increasing use of complex digital systems to lower costs, simplify operations andimprove reliability, placing an increased burden on demonstrating that the digital systems arecorrectly deployed and will meet performance specifications. Advanced signalling
techniques including multiplexing and the use of wireless technology will reduce cablingcosts, while presenting challenges for assuring signal integrity.

x

PHWRs. PHWRs started out using more digital systems for plant safety and control, but are
now pretty much on par with LWRs, and therefore face the same issues.
Table 3.2.2-1. Current plant control and monitoring issues

Issues and relevant reactors
1) Software quality and reliability:
PWR, BWR, VVER, PHWR,
ALWR, APHWR, HTGR

2) Environmental qualification:
PWR, BWR, VVER, PHWR,
ALWR, APHWR, HTGR

3) Digital system reliability:
PWR, BWR, VVER, PHWR,
ALWR, APHWR, HTGR

4) Wireless communication:
PWR, BWR, VVER, PHWR,
ALWR, APHWR, HTGR

5) On-line monitoring andadvanced instruments:
PWR, BWR, VVER, PHWR,
ALWR, APHWR, HTGR

Description
With the increasing use of digital instrumentation, control (I&C) andprotection systems in currently operating plants, the extensive plansfor complete control room retrofits using digital systems and the plans
for their use in future plants, how to ensure the quality and reliabilityof the software used to perform safety functions is a growing concern.
Software verification and validation methods, as well as, qualitativeand qualitative software testing methods need to be assessed and their
attributes and effectiveness established to aid in the review andregulation of software based systems important to safety.
The environment in which they operate (e.g. temperature, humidity,
radiation, smoke, electro-magnetic/radio frequency interference, etc.)
can affect the performance (reliability, failure rate, and failure mode)of digital I&C systems. Several standards developed by IEEE and IEC
are used to guide qualification testing. For future plants with different
environmental conditions new methods and tests will be needed toestablish failure thresholds and modes of failure.
To understand the performance of digital systems, the integration ofsoftware, hardware and humans is needed. Although checks of thevarious system components individually are also required, they are
not sufficient to confirm overall system performance and reliability.
Facilities where such testing can be done in prototypical fashion areneeded. Additionally methods are needed to support integration ofdigital system reliability models into current generation PSAs.
The use of wireless communication for monitoring and control in
nuclear power plant is expanding rapidly. The technology that supports
the current generation of wireless applications were not designed for thechallenging environments in nuclear plant that have the potential todisrupt signals. There are needs to do be research in order to understandpossible effects before such communication is used for safety functions.
Testing is needed to confirm design and performance. Additionally the
security aspects of wireless communication needs to be explored.
The use of advanced on-line monitoring and systems diagnostics innuclear power plant instrumentation and control systems has added ahigher level of complexity to the current generation instrumentation,
control and protection systems, in addition to the complexity already
added by the use of digital systems. Although these systems have thepotential to reduce operator workload and increase system reliability,
the new and complex failure modes for these systems needs to beinvestigated. Facilities where these systems can be tested andreviewed in a prototypical fashion are needed.

81

Table 3.2.2-2. Facilities in the area of plant control and monitoring
Facility name
Sandia National Laboratory – Environmental qualification lab (USA)
University of Virginia – Centre for Safety Critical Systems (USA)Ohio State University – INL Academic Centre of Excellence in Nuclear
Instrumentation and Control and Safety Analysis (USA)
Halden Reactor Laboratory – HAMMLAB and Software Engineering
Laboratory (Norway)
Notes:
Specify range
Operation cost
Replacement cost:::

Operational
cost/year
Low
Low

Replacement
cost
MediumMediumLow

MediumMediumMediumLow, Medium, High.
Low is <1.0 MUS$/y; Medium is 1.0-2 MUS$/y; High is >2 MUS$/y.
Low is <2 MUS$; Medium is 2-10 MUS$; High is >10 MUS$.

82

Table 3.2.2-2. Facilities in the area of plant control and monitoring
Issuesaddressed
2, 5
1, 3
1,3,5,
1, 2, 5

Capabilities
Smoke temperature humidity and radiation testing
Integrated systems testing. Simulates operating conditions, faults,
instrumentation and equipment failures.
Dynamic reliability analysis, advanced and high temperature
sensor design and analysis
Ability to simulate advanced control room and computer based
control and diagnostic systems

Planned durationof operation
Currently in standby.
Indefinite.
Has non-nuclear funding.
Indefinite.
Indefinite.

3.2.3 Seismic effects
Introduction
The seismic behaviour of components and structures has the potential to impact plant safety bysimultaneously affecting all plant systems structures and components (i.e. common mode failure). To
ensure plants are designed for such events, data to confirm seismic design and seismic safety
evaluation methods need to be obtained.
Scope
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 for which plants
must be designed varies across member countries and with plant age, since seismic concerns are site
specific and methods are evolving. For future plants (which are likely to be standard designs to bemarketed worldwide), it is expected they 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 plantseismic safety (e.g., below ground structures, seismic isolation devices) which will need experimental
confirmation. The safety issues associated with the seismic behaviour of components and structuresthat could benefit from additional research are shown in Table 3.2.3-1.
Description
Data to validate seismic design methods is essential to ensuring plant seismic safety. The dataneeded is generally in one of the following two categories:xx

Ground motion characteristics (e.g. frequency, magnitude, direction, etc.).
Structural response to ground motion.

Seismic simulation facilities (i.e. shake tables and reaction walls) are needed to obtain data forthe second category and are the subject of this section. Such facilities have the capability to simulate awide variety of seismic motions (which are not specific to the nuclear industry) with scaled or full size
equipment. Data generated from such simulations can be used to compare against analytical results or
establish failure modes and thresholds, thus allowing quantification of margins to failure.
In July 2004, CSNI issued a report titled Experimental Facilities for Earthquake Engineering
Simulation Worldwide, NEA/CSNI/R(2004)10, that summarised the capabilities of shake tables andreaction walls in member countries. Although the CSNI report concluded that none of the facilities
were currently in danger of being closed, this section looks at the longer term prospects of facility
value and use. Only large shake table (> 100 ton capacity) and large reaction wall (> 15 m tall)
facilities are included, since these have the greatest versatility and capability.
83

Table 3.2.3-1. Current seismic effects issues
1)

Issues and relevant reactors
Confirmation of seismic
design:APHWR, ALWR, HTGR

2)

Below ground siting:
ALWR, HTGR

3)Continued safe operation:
BWR, PWR, VVER, PHWR

4)

Seismic isolation devices:
ALWR

Description
New plant designs are incorporating new safety features (e.g., passive
ECCS, passive containment cooling) that need to be designed towithstand seismic events. Also, some designs may incorporate seismic
isolation features to limit the transmission of ground motion to plantstructures and equipment. In both cases, experimental confirmation of
the design’s ability to withstand seismic events and data to validate
analytical tools will be necessary.
Some future designs may locate all or some critical systems, structuresand components below ground to protect them from external events.
The response of below ground structures to seismic events needs
experimental data to confirm analysis methods.
As seismic events continue to occur and plants continue to age, data to
confirm continued safe operation may be necessary. This data could bein the form of simulating the earthquake and the aged plant structure.
Some future designs may incorporate seismic isolation devices into thedesign to reduce the seismic ground motion transferral to vital plantstructures. Experimental confirmation of the performance of thesedevices should be obtained.

Table 3.2.3-2. Facilities in the area of seismic effects
Facility nameShake tables
CEA TAMARIS Facility(France)
NIED (Japan)
NIED (Japan)
Public Works Research Facility(Japan)Reaction walls
EC-JRC (Italy)
Building Research Facility(Japan)Cost/year,
operation

Replacement
cost

IssuescoveredCapability

Planned durationof operation

High

High

1,3,4

100 tons

Indefinite

Low
High
Low

High
High
High

1,3,4
1,3,4
1,3,4

500 tons
1 200 tons
100 tonsHigh
Medium

High
High

1,3
1,3

15 m
25 m

Indefinite

Notes:*
There are no facilities currently that can mock up below ground structures.
Specify range: Low, Medium, High.
Operational Cost
: Low is <1.0 M US$/Y; Medium is 1.0-2 M US$/Y; High is >2 M US$/Y.
Replacement Cost : Low is <2 M US$; Medium is 2-10 M US$; High is >10 M US$.

3.2.4 Fire assessment
Introduction
Fires present a very demanding generic problem to plant safety, which has been demonstrated bysome serious incidents of the past and by several plant specific safety analyses. The long experience offire 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. Theoretical modelling of fireis very demanding because of its multiple effects and because some key parameters are not well84

known. Modelling experience of fires has, however, gradually proceeded hand-in-hand withexperimental work. In the past years zone models and multi-compartment zone models gained
promising results. Zone models continue to be used and validated, due to their ease of use. During
recent years, progress in computing power caused CFD models to become everyday tools forengineering applications. Zone models have been complemented by CFD codes, since CFD models
are much more versatile, have no compartment size or configuration limitations, and allow predictionof all physical variables of fires if the fire size is known. Remaining problems include determinationof fire size on solid fuels and development of reliable models for distributed fire loads like cables.
Brute force methods like solving Navier-Stokes equations numerically are not within the foreseeable
future. Promising theoretical and experimental results in micro-gravity and other aerospace relatedindustries have indicated, that new analytical modelling for the flame spread is possible. 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 the development of deterministic and stochastic
fire modelling. Comparisons of code predictions with relevant experiments, benchmarking, and othersimilar international comparisons of the codes during past years have given good reliance on thetechnology in general, and also selected the most useful numerical codes. The development ofcomputational tools and accumulating experience is gradually enabling fire PSA on the same realistic
level as in other branches of PSA.
However, additional experimental data is needed to assess the codes and new safety issues havearisen 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. The fire assessment safety issues that could benefit from additional
research are shown in Table 3.2.4-1.
Description
Because of the generic nature of fire, nuclear power plant specific experimental facilities are notnecessary. Most of the experimental work has been conducted using inexpensive small scaleequipment.
Facilitated by the increase 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 isneeded to fix crucial parameters of the modern fire models. The major problem is to calculate fire size,in connection with active air-solid interface of distributed fire loads like cables. Actions should betaken to test various proposed models of the emerging technology at all relevant scales, and implement
the promising models in the best CFD codes.
Quantitative assessment of a fire scenario needs calculation using Monte Carlo techniques. Mostof the needed deterministic fire models already exist, as well as some calculation platforms. Efforts ofcreating needed input data banks should be started soon.
While the computing and simulation models tools are, to some degree, able to utilise alreadyavailable knowledge and tools, in practice several parallel development lines are needed; (a) databases from the most safety relevant fire scenarios, (b) ignition and flame spread data measurements forthe relevant materials, c) models of automatic fire protection, (d) and quantitative assessment ofmanual fire protection.
85

Table 3.2.4-1. Current fire assessment issues
Issues and relevant reactors

Description

1) Fire growth and propagation:
PWR, BWR, VVER, PHWR,
ALWR, APHWR, HTGR

Accurate modelling of fire growth and propagation is the key todetermining the time and extent of equipment affected. While the best
CFD simulation models are able to predict consequences for given fires
rather satisfactorily common efforts should be taken to simulate mostsafety relevant typical scenarios as benchmarks, and make the results
available in data bases as example and study material for plant specificwork.
Efforts should be taken to utilise the emerging technology of flame spread
modelling on solids. Actions should be taken to test various proposedmodels at all relevant scales, and implement the promising models in thebest CFD codes. Special efforts are needed to select the most suitable
testing methods from existing or new concepts, which are needed todetermine flame spread parameters for practical commercial products. For
example for cables, none of the available methods are able to determine
them at present. Benchmarking efforts are needed to transfer thetechnology from laboratories to industrial practice.

2) Hot shorts:
PWR, BWR, VVER, PHWR,
ALWR, APHWR, HTGR

Fires in cable trays can 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 well understoodor modelled in safety analysis. Experimental data is needed. There aresome data already available for simple basic scenarios. Theoreticalmodelling is needed for assessing effects on systems performance.

3) Smoke propagation:
PWR, BWR, VVER, PHWR,
ALWR, APHWR, HTGR

The spread of smoke during a fire is generally not modelled, althoughmost of the needed technology exists. Smoke can affect the operability ofcertain equipment and inhibit human fire fighting efforts by limiting
access and visibility. However, this is not unique to the nuclear industry
and steps should be taken to implement existing technology.

4) Equipment vulnerability:
PWR, BWR, VVER, PHWR,
ALWR, HTGR

When and how equipment fails under fire conditions is essential to fire
assessments. This includes failures from heat, smoke, suppression system
activation, shorts, etc. Experimental data will likely be needed to address
this issue.
There are some data and basic calculation models available on heat andsmoke effects for some equipment. Establishing a data bank withbenchmarking examples would be a good way to educate utilities to usethat information. For Monte Carlo analyses establishing these data banks
is mandatory.

5) High energy arcing faults:
PWR, BWR, VVER, PHWR,
ALWR, HTGR

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

86

Table 3.2.4-2. Facilities in the area of fire assessment
Cost/year, Replacement Issues
operation
cost
coveredFacility name
GALAXIE (France)

High

High

1,3,4

Sandia Labs (USA)MediumMedium

1,3,4

Omega Point (USA)MediumMedium2

U.S. National Institute ofStandards and Technology(NIST, USA)
VTT (Finland)
DIVA (France)

MediumMedium1,3

MediumMediumMediumMedium1,4
1,3,4

Notes:
Specify Range
Operational Cost
Replacement Cost:::

Capabilities

Planneddurationof operation

Facility composed ofcompartments (from 1 to 680 m3 )and needs (up to 30 000 m3/hr for
fire intonsities up to 2 MW).
Fire growth, propagation, effects In-stand byof temperature and smoke onequipment.
Issues covered include spurious
Indefinite
actuation of equipment.
Issues covered include fire growth Indefinite
and propagation.
Experimental/fire modelling
Part of GALAXIE (5 rooms
connected by a ventilation
system) to investigate electricalcabinet fires, heat, and smoke
propagation from room to room.

Through2010

Low, Medium, High.
Low is < 1.0 MUS $/yr.: Medium is 1.0-2.0 MUS $/yr.: High is > 2 MUS $/yr.
Low is < 2.0 MUS $: Medium is 2-10 MUS $: High is > 10 MUS$.

3.3 HTGR unique safety issues and research needs
Introduction
In February 2002, CSNI hosted a workshop to discuss safety issues and research needs associatedwith advanced reactors. The results of this workshop were documented in an NEA publication entitled
Advanced Nuclear Reactor Safety Issues And Research Needs, published in 2002. The scope of thisworkshop covered ALWRs, HTGR, and LMRs.
The SESAR/SFEAR group used the results of the CSNI workshop as a starting point in assessing
ALWR, APHWR and HTGR safety issues and research needs. This was supplemented with morecurrent information, when available, and reflects the views of the SESAR/SFEAR participants. The
purpose of the assessment by the SESAR/SFEAR group is to identify those safety issues and researchneeds judged to be important to the safety and licensing of future HTGRs. The results of thisassessment are for use by designers and regulators in planning and conducting programmes in support
of future HTGRs. Due to the early stage of development of future HTGRs, no CSNI action isnecessary or recommended at this time. It is also recognised that many safety issues associated withcurrent plants and ALWRs/APHWRs also pertain to (e.g., human and organisational influences).
These issues have been covered under current plant issues in Section 3.2 and are not repeated here.
Scope
This section addresses those issues unique to HTGRs. They fall in the same technical areas ascovered in Section 3.1. The technical areas covered in this section are as follows:x

Thermal-hydraulics.
87

x
x
x
x

Reactor fuel.
Reactor physics.
Severe accidents.
Integrity of equipment and structures.

Description
Discussed below are the safety issues and research needs associated with each of the above
technical areas. Where appropriate, facility needs are also discussed.
3.3.1 Thermal-hydraulics
A key safety attribute of HTGRs is their ability to remove decay heat, even in the event of loss of
coolant. In addition, the large amount of graphite used as a moderator serves as a heat sink which
provides for slow core heat-up during loss of coolant events. However, during normal and off-normal
operation structural steel materials need to be protected from the high temperatures at which HTGR
cores operate. Accordingly, the thermal-hydraulic area presents challenges for designers andregulators to ensure analytical tools can model the various modes of operation and accident conditions.
This will require understanding and modelling of heat conduction, convention and radiation heattransfer and ensuring sufficient experimental data to validate the models. Discussed below are safety
and research needs in the thermal-hydraulic area.
Data will be needed to evaluate the accuracy of codes and assess margins of safety. Test data canbe obtained from facilities ranging in size and complexity from small-scaled component tests to scaled
representations of the entire system. Past and ongoing HTGR research has been conducted at suchreactor facilities as the AVR, Thorium Hochtemperaturreaktor (THTR) in Germany, the HighTemperature Engineering Test Reactor (HHTR) in Japan, and the 10-MWe High-Temperature Reactor(HTR-10) in China. These and other experimental programmes, such as the air-ingress tests done in
the NACOK facility at FZ-Jülich and in a similar facility at Japan Atomic Energy Research Institute(JAERI), as well as the pebble-bed fluid-flow and heat-transfer tests performed in the SANA facility at
FZ-Jülich, provide significant sources of measured T/Hs data. However, additional data is needed toinvestigate issues including pebble-bed hot spots inferred from the melt-wire test results at AVR, theincomplete mixing of reactor outlet helium and thermal stratification natural circulation under loss offorced circulation accidents, air and moisture ingress accidents with oxidation, and reactor cavity
cooling.
Existing test reactors may be useful for obtaining some data for code assessment, particularly forsteady state and some transient operation. However, scaled facilities testing may be needed to confirm
plant performance under more severe accident conditions such as:xxx

Loss-of-coolant.
Station blackout.
ATWS.

Also, since power, decay heat and flux distribution are important to thermal-hydraulics
calculations, coupled T/H and reactor physics codes may be necessary.
3.3.2 Reactor fuelThe safety of HTGRs is tied more closely to fuel performance than is the case for LWRs,
ALWRs or PHWRs. The coated fuel particles are intended to retain fission products, even at hightemperatures experienced during many accident conditions (e.g. up to 1600(C), and thus avoid
88conditions representative of widespread core damage. However, our current understanding of HTGR
fuel safety limits and the factors that can affect them are not as well understood as those for LWR and
PHWR fuel.
The safety claims of the HTGR design are inherent in the assumption of predicted performance of
the TRISO-coated fuel particles (CFPs) under potential accident conditions. The HTGR fuel uses
higher enrichment and operates at higher temperatures than conventional LWRs. The value of 1600(Cis the proposed maximum permissible fuel temperature beyond which some degradation of the siliconcarbide protective coating occurs. However, the integrity of the coated fuel particles is also dependent
upon their manufacturing process and upon the environment within the reactor core.
Discussed below are the safety issues and research needs associated with HTGR fuelperformance. Due to the importance of fuel performance to HTGR safety, the long lead time and thecost of fuel testing, this is one area where international collaboration should be strongly considered.
Virtually all of the past and ongoing worldwide irradiation testing research of HTGR fuel designs
with TRISO CFPs include accelerated irradiations in MTRs. Although there subsequently wassignificant large-scale operating experience with these fuels in plants such as the AVR in Germany,
accident simulation tests (i.e., fuel heatup test following irradiation) to qualify the fuel involved
accelerated irradiations in MTRs. A well-established and thorough understanding of the mechanics
and properties (e.g., creep) of CFP behaviour, failure, and FP release does not exist to allow one toconclude, with certainty, that fuel accident simulation tests following accelerated irradiations areconservative as compared to the rate of fuel irradiation in a power reactor. Accident simulation heatup
tests, either after real time MTR fuel irradiations or after fuel irradiations in a power reactor, would beneeded to resolve this issue.
Virtually all of the accident simulation tests for TRISO CFPs involved so called “ramp and hold”
temperature increases. These typically consist of increasing fuel temperature at about 50(C/hr. up to a
set temperature (e.g., 1600(C, 1700(C or 1800(C) and then holding the fuel at the set temperature forseveral hundred hours while FP release measurements are taken. The results of ramp-and-hold tests up
to 1600(C, for qualified fuel, show that no additional CFP failures occur. However, in the FRG, there
was at least one test in which the temperature was controlled to closely simulated the predictedaccident heatup curve to 1600(C for a design-basis reactor coolant pressure boundary failure. For thistest, CFP failures were observed to occur. Additional post-irradiation accident simulation tests thatclosely simulate the predicted temperature curve for a design-basis reactor coolant pressure boundary
failure would be needed to determine if the traditional ramp-and-hold test accident simulation
approach is conservative with respect to establishing CFP failure rates for postulated accidents.
Among the most limiting events that could challenge HTGR CFP integrity are those involving
large-scale chemical attack, such as air intrusion following a large pipe break in the reactor coolant
pressure boundary and moisture intrusion for a postulated heat exchanger tube failure with the reactor
helium pressure falling below the heat exchanger tube pressure. Experiments on unirradiated HTGR
fuel in air and water at HTGR accident temperatures have been conducted. These experiments have
involved measurements of fuel oxidation due to air or moisture impurities in helium during fuelexperimental irradiations. However, few experiments have been conducted on fully irradiated HTGR
fuels that simulate the effects of large air or water ingress events. Additional post-irradiation accident
simulation tests that closely simulate air or water intrusion events and take the fuel to the onset of CFP
failures would be needed to fully assess the adverse effects of air and water corrosion on HTGR fuels
and the margins to failure for such events.
Very limited testing has been conducted on fuels with TRISO CFPs to assess the capabilities andthe margins to CFP failure for reactivity events involving a large energy deposition in the fuel over a89very short time interval (less than 1 second). Some limited testing was conducted in Japan for a
postulated control rod ejection accident in support of the HTTR licensing, this scenario was one of thelimiting licensing basis events. Accordingly, in order to fully understand the margins to failure forreactivity events, fuel irradiation experiments involving such reactivity insertion events may be useful.
Only limited worldwide testing has been conducted on previously qualified FRG or U.S. HTGR
CFP fuel for conditions that go beyond the maximum qualification operating temperature andmaximum qualification fuel burn-up. In order to fully understand the margins to CFP failure and FP
release for fuel operations beyond the maximum allowed operating temperature and design fuel burnup limits, fuel experiments involving irradiation conditions beyond such limits would need to be
conducted.
Accordingly, much experimental data will need to be generated to validate fuel performance.
This will include:x

Data which explores the safety limits for fuel performance and FP release for conditions thatare beyond the design basis for parameters important to fuel performance. These conditions
involve fuel operating temperature, maximum fuel accident temperature, fuel oxidising
environment, fuel burn-up, energy deposition and deposition rate in the fuel (due toreactivity accidents), beyond those that are expected to be examined by the fuel vendor orapplicant. Such data will help established margins in fuel performance and establish
appropriate operating and licensing limits.

x

In-core hot spots. The results of melt-wire experiments conducted in the German AVR test
reactor demonstrated the existence of unpredicted local hot spots under normal operating
conditions in pebble bed cores. Such hot spots can be used to determine the maximum
normal operating temperatures of the fuel. These hot spots may arise from a combination ofhigher local power density (e.g. due to moderation effects near the reflector wall or fromchance clustering of lower burn-up pebbles), lower local bed porosity due to locally tightpebble packings, and reduced local helium flow due to the increase of helium viscosity withtemperature. Whereas the slow evolution of loss-of-cooling heatup transients will tend towash out any effects of pre-accident local flow starvation on subsequent peak fueltemperatures, the effects of higher local fission power densities will be retained throughout
the heatup transient in the form of higher local decay heat powers. Therefore, data on theeffect of decay-power hot spots, in particular, may be needed in evaluating the maximum
fuel temperatures arising in pressurised or depressurised LOCAs.

x

Physics of TRISO fuel irradiation in test reactors versus HTGRs. The extensive use ofvarious test reactors for the irradiation testing of HTGR TRISO fuels raises questions about
the non-prototypicality of the neutron energy spectra, accelerated fuel burn-up rates, and fueltemperature histories in the test reactors. Reactor-specific calculations of neutron fluxes andnuclide generation, depletion, and decay may need to be performed to provide a basis foranalysing the sensitivity of computed fluences and fuel nuclide inventories to the neutronic
differences between the test reactors and HTGRs. Of interest are the potential effects of suchdifferences on TRISO fuel performance (i.e., FP retention) under normal and accident
conditions. Such differences include the variations in irradiation temperature histories, burnup rates, and neutron energy spectra that result in different neutron fluences, different ratesof plutonium production and plutonium fission versus uranium fission, and, thus, different
yields of important FPs. It is known, for example, that 236U and 239Pu give substantiallydifferent yields of various FPs that potentially affect TRISO fuel performance.

x

Data on the effect of fuel fabrication parameters on fuel performance.
90

To obtain the experimental data needed to confirm fuel performance and validate fuelperformance codes will require irradiation facilities and hot cells capable of steady state and transient
testing. Prototypic irradiation and transient test conditions will need to be demonstrated. The effects ofvariations in fuel fabrication will also need to be tested to understand their impact on fuelperformance. In this regard, test reactors capable of testing HTGR fuel at steady status and transientconditions are essential to generate data to establish fuel performance.
Existing test reactors, such as CABRI, NSRR and ATR will be import to maintain due to theirability to also test HTGR fuels.
3.3.3 Reactor physics
Since controlling fuel temperature is key to ensuring the safe operation of HTGRs (see Reactor
Fuel below), reactor physics modelling and analysis is important in understanding power distributions
and responses to off-normal events. Also, some future HTGR designs are attempting to incorporate
passive shutdown capability (using Doppler feedback from 238U as fuel temperatures increase) which
will require a coupling of thermal-hydraulic and reactor physics analysis methods. Additionally, somereactivity insertion mechanisms are different in HTGRs than in LWRs (e.g. compaction of pebble fuelduring a seismic event, water ingress acting as a neutron moderator) and these needs to be modelled
and analysed.
All of the above will require good knowledge and data on cross sections, power distributions,reactivity coefficients and other reactor physics parameters. Discussed below are safety and researchneeds in the reactor physics area.
Nuclear analysis infrastructure development will be necessary to analyse reactor and fuelperformance under a variety of normal and off normal conditions. Items needing research are:x

Temperature coefficients of reactivity. Validated analytical tools will be needed to confirm
that the reactivity feedback effects from temperature changes in the fuel, moderator graphite,
central graphite region, and outer reflector graphite are appropriately treated in safety
analyses. Sensitivity analyses and validation against representative experiments and should
be used to assess and account for computational uncertainties in the competing physicalphenomena, including for example, the positive contributions to the fuel and moderator
temperature coefficients associated with 135Xe and bred fissile plutonium.

x

Reactivity control and shutdown absorbers. The reactivity worths of in-reflector control andshutdown absorbers may be sensitive to tolerance in the radial positioning of the absorbers
within the core. Analytical evaluations for reactivity control and hot and cold shutdown willneed to account for absorber worth variations through burn-up cycles and the transition frominitial core to equilibrium core loadings. Modelling uncertainties, and absorber worth
variations caused by temperature changes in the core and reflector regions, xenon effects,
variations or aberrations of pebble flow, and accidental moisture ingress will need to beunderstood to ensure valid predictions of absorber worth.

x

Moisture ingress reactivity. Although the absence of high-pressure, high-inventory water
circuits in closed Brayton cycle systems makes this issue less of a problem than in earlier
steam cycle HTGRs, the effects of limited moisture ingress will nevertheless need to beevaluated for depressurised and pressurised accident conditions. Effects to be evaluatedinclude the moisture reactivity (i.e. from adding hydrogenous moderator to the undermoderated core), the effects of moisture on temperature coefficients (e.g. from spectral
softening), shortened prompt-neutron lifetimes (i.e., faster thermalisation), and reduced
worths of in-reflector absorbers (i.e. fewer neutrons migrating to the reflector).
91

x

Reactivity transients. T/H-coupled spatial reactor kinetics analyses may be needed to assess
axial xenon stability, as well as reactivity transients caused by credible events, such as
overcooling, control rod ejection, rod bank withdrawal, shutdown system withdrawal orejection, seismic pebble-bed compaction, and moisture ingress. Of particular importance isthe need to identify any credible events that could produce a prompt supercritical reactivity
pulse. Should any such prompt-pulse events be identified as credible, their estimatedprobabilities and maximum pulse intensities should be considered in establishing any relatedplans or requirements for pulsed accident testing and analysis of HTGR fuels. For loss-ofcooling passive shutdown events with failure of the active shutdown systems (i.e. anticipatedtransient without scram – ATWS), the delayed recriticality that occurs after many hours of
xenon decay may also require spatial kinetics analysis models to account for the uniquespatial power profiles and feedback effects caused by the higher local reactivity near theaxial ends and periphery of the core where temperatures and xenon concentrations are lower.

To address the above, critical facilities capable of experimental confirmation of HTGR reactor
physics analysis results are essential. Examples of such facilities are HTR-PROTEUS (Switzerland)and ASTRA (Russia).
3.3.4 Severe accidents
HTGR designers are attempting to reduce, as much as possible, the potential for and consequencesof severe accidents. This is being done by incorporating passive safety features into the design (e.g.,reactor shutdown, decay heat removal), by ensuring high quality fuel and by removing the potential for
air and water ingress into the core. Likewise, understanding fission product release and transportmechanism is key to designing appropriate plant features to limit the release of radioactive material.
Discussed below are specific safety issues and research needs pertaining to HTGR severe
accident behaviour, with emphasis on understanding the potential for the release of fission products(FP) from the fuel into the environment.
For HTGRs, both the types of sequences and the process by which FPs may be released from thefuel are different than current generation LWRs. In HTGRs, FPs may be released as a result ofdiffusion during normal operation, by rupture of coated fuel particles as a result of accidents, any byvaporisation during high-temperature degradation of the fuel.
The risk from HTGR operation is the risk from releases during normal operation and, from
accidents involving rupture of coated fuel particles. Technical expertise and technical capability in thearea of FP transport and behaviour during high-temperature fuel degradation is needed in order toassess the risk from HTGR operation. Because FPs released from the fuel are transported through theprimary system and containment predominantly as aerosols, the offsite releases and offsite
radiological consequences may be significantly reduced by FP deposition in the primary system andcontainment. Aerosol deposition occurs through a variety of mechanisms, such as gravitational
settling, thermophoresis, and diffusiophoresis. Therefore, research activities should focus on FP
transport and behaviour in the primary system and containment of other structural buildings.
HTGR designers will propose the accident source term to be used in their safety analysis basedon models and methods that mechanistically predict FP release from the fuel. Should this be the case,it would differ from the traditional deterministic licensing approach to source term used by LWRs,
which involves a pre-determined conservative upper bound for the accident source term. HTGR
designers may also likely propose that HTGR plants utilise a non-leak-tight “confinement” structure,92

rather than a traditional leak-tight and pressure retaining containment structure. Accordingly, thesafety analysis for modular HTGRs will largely hinge on the capability to confirm fuel FP release andassociated uncertainties.
The qualification of HTGR fuels will be based on a wide range of technical areas and specificfactors that are known to influence fuel performance, such as FP release and particle failure rates. The
technical areas include fuel design; fuel manufacturing process, including process specifications; andstatistical product specifications; design-specific core operating conditions; design-basis accident
conditions, and postulated accident conditions beyond design basis. Key factors within the designspecific plant operating conditions that are known to affect fuel (particle) performance include fueloperating temperature, fuel burn-up, particle fast fluence, particle power, and fuel residence time in thecore. The key factor affecting fuel particle performance during an accident (following the priordegrading effects of the operating conditions) is the peak particle temperature during the accident.
Temperature increases can occur due to heatup events, which are caused by the loss of normal cooling,core power increases, or significant local reactivity insertion events. Other factors potentially affecting
fuel (CFP) performance during accidents can include the effects of chemical attack (e.g., oxidation) on
the fuel element and (possibly) the CFPs.
To predict CPF performance and a deterministic approach to the source term, capabilities in a
number of interfacing technical areas will be needed. These include: (1) nuclear analysis for fuel burnup, fast fluence (for particle coating behaviour), thermal fluence (for particle power and fuel kernel
behaviour), and fuel particle power during reactivity events and (2) T/H analysis of normal operatingcore temperature distributions, accident core temperature distributions, and core temperature and flow
distributions (for fuel oxidation during postulated air intrusion events). The FP release rates from thefuel during normal operation and postulated accidents are key inputs to the accident source term
calculation. Accordingly, a range of significant fuel design, fuel manufacture, fuel quality, and fuelperformance issues exist which will require research to resolve.
It should also be stated that understanding fission product release from the fuel will also require
understanding the performance of other plant safety features, such as shutdown mechanisms, graphite
behaviour, metallic structure behaviour, decay heat removal and FP transport. Accordingly there is aninterrelationship among this section and other sections in this chapter.
3.3.5 Integrity of equipment and structures
During operation, various HTGR core internals, reactivity control elements and structuralelements as well as system components will be exposed to higher temperatures than those inconventional LWRs. Issues that need further consideration would include a) applicability of theexisting database of currently qualified high temperature materials, including the effect of various
coolant impurity levels to the specific HTGR applications, b) the adequacy of procedures forevaluating material properties for HTGRs, and c) in-service inspection examination and surveillanceplans and techniques.
There is also a need to establish a database related to the long-term performance and behaviour ofgraphite material under high temperature and irradiation levels expected during normal operation andaccident conditions in HTGRs. This includes understanding the affect of the graphite manufacturing
process and impurities on material properties and their changes due to temperature, irradiation, etc.
Properties of interest are strength, irradiation creep, shrinkage, swelling, thermal conductivity, andfatigue. The issue of the loss of structural integrity of graphite material also needs carefulconsideration because it is one of the key issues which would impact the long-term performance ofgraphite structures, including the top- and bottom-reflector as well ad the end-of-life behaviour of all
93

graphite structures. It is also important to understand graphite oxidation behaviour under accident
conditions, such as air ingress.
Discussed below in more detail are the safety issues and research needs associated with graphite,
metallic and concrete structures for HTGRs.
Graphite structures
In HTGRs, graphite acts as a moderator and reflector, as well as a major structural component,
providing channels for the fuel and coolant gas, and control and shutdown rods, and acting as athermal and neutron shield. Additionally, graphite components are employed as supports. Graphite
also acts as a heat sink during reactor trip and transients. During reactor operation, many physicalproperties of graphite are significantly modified as a result of temperature, environment, andirradiation. Significant internal shrinkage, bowing, and stresses develop which may cause component
failure, and/or loss of core geometry. Additionally, when graphite is irradiated to a very high radiationdose, swelling occurs, which also causes rapid reduction in strength, reducing the components
structural integrity. In the event of an accident causing air ingress, subsequent graphite oxidation
causes further changes in its physical and mechanical properties.
Some irradiation studies have been conducted on older graphites that are no longer available due
to loss of raw materials supply and/or manufacturers. In addition, limited results are available at highlevels of irradiation exposure. Thus, two key issues are the lack of data on irradiated properties ofcurrent graphites and the lack of data at higher doses of irradiation. Irradiated material properties areheavily dependent on the particular make-up of the graphite and the manufacturing process; therefore,
at issue is whether the irradiated materials properties of the “old graphites” can be assumed to be thesame as the “new graphites”. Irradiation affects, and in many cases degrades, the physical andmechanical properties of the graphite. Important properties that change with irradiation are density,
thermal conductivity, strength, and dimensions. Some of these changes are not linear with irradiation
dose. Graphite strength initially increases with irradiation dose, then, at higher levels, it begins todecrease. With respect to dimensional changes, graphite initially begins to shrink with increasing dose,then, beyond turn-around, graphite beings to swell with increasing dose. During operation, thermal
gradients and irradiation-induced dimensional and strength changes can result in significant
component stresses, distortion, and bowing of components. These can lead to loss of structural
integrity, loss of core geometry, and potential problems with insertion of control rods. At still higher
doses, beyond turn-around, where the swelling makes the volume considerably greater than theoriginal volume, graphite structures and fuel balls will start to disintegrate and experience total loss of
integrity.
Accordingly, there is a need to conduct confirmatory research to establish an information baserelated to the long-term performance and behaviour of nuclear-grade graphite under the temperatures,radiation, and environments expected during normal operating and accident conditions. Potential lossof strength and of resistance to fatigue and creep, shrinkage, swelling, cracking, and corrosion during
operation could impact the performance and function of the graphite core components. Various
variables, including coke source, size, impurity, and structure; manufacturing processes; density; grainsize; and crystallite size and uniformity determine the as-received and irradiated properties of thegraphite component and need to be correlated to the irradiated graphite properties.
To evaluate the suitability of a particular graphite for HTGR application, property change data
due to irradiation is needed in addition to the as-received properties. Development of irradiation dataon graphite is difficult, expensive, and time consuming. Therefore, reactor designers/vendors haveproposed to use radiation data from studies conducted on older graphites and attempt to use graphites
94

produced in a similar manner. However, the as-received and irradiated graphite properties depend
strongly on the raw materials and manufacturing processes. Small variations in these may have strong
effects on the graphite properties. Since the exact raw materials and processes have changed and may
continue to change in the future, there is a need to independently confirm whether a particular graphite
will behave the same as the old graphites under operating irradiation conditions. To accomplish thiswithout irradiation testing every time a change occurs in the graphite raw materials or processing,correlations are needed for predicting irradiated graphite properties and changes from the as-received
graphite raw materials characteristics, composition, processing, and properties.
Graphite corrosion and oxidation can occur in HTGRs from oxidising impurities in (or added to)the helium coolant, from in-leakage during normal operation, or from air or water ingress during
accidents. The oxidation of graphite is an exothermic reaction, and it is important to know the rate ofheat generation, particularly during accidents. Oxidation also will remove the surface layers ofgraphite components resulting in loss of structural integrity. Further, oxidation will change the thermal
conductivity and reduce the fracture toughness and strength of graphite components. The loss instrength may be due to attack of the binder. The oxidation rates vary for different graphites, and can begreatly affected by the impurities in the original graphite. Therefore, oxidation rate data is needed forthe graphites proposed for new reactors.
The graphite structures will consist of thick and relatively thin pieces. The relatively thin
structures may be manufactured differently from large structural blocks of graphite, and themechanical and other properties may be different. Furthermore, the properties of the large block
graphite will vary through the thickness of the block. The difference in properties between the sleevesand large blocks and through-thickness variations need to be established. The potential for different
irradiated properties of sleeve graphite and large block graphite also needs to be evaluated.
There is a lack of standards for nuclear grade graphite. To ensure consistency, nuclear graphites
should meet certain minimum requirements with respect to important properties, such as strength,
density, and thermal conductivity, as is the case for materials used in other reactor systems. Specificimpurities in the graphite might be detrimental to irradiation properties of the component, and theyshould be limited in nuclear graphites. Other elements, such as halides, which can be released during
operation and cause degradation of other components in the reactor, should also be limited in nuclear
grade graphite. Thus, standards need to be developed to establish the acceptable physical, thermal, andmechanical properties, composition, and manufacturing variables for nuclear grade graphite.
Metallic structures
National codes and standards for the design and fabrication of metallic structures and components
for HTGR service conditions are needed. Although methodologies could be assembled from existingknowledge for calculating fatigue, creep, and creep-fatigue lives of components in high-temperatureapplications, appropriate data bases are needed for these calculations. Based on past experience andresearch, environmental effects play an important role in reducing fatigue lives and in enhancing
degradation of materials. For example, small levels of impurities, such as less than 1 part per millionof oxygen in the high-purity water coolant of LWRs, can greatly decrease fatigue life and resistance tostress corrosion cracking of metallic components. Because helium is inert, there has been a tendency
top obtain design data in pure helium; in impure helium (but not all impurities were included) or in air.
The effects of all important impurities, such as oxygen, in helium need to be taken into account withrespect to reductions in fatigue and creep life and such data and understanding need to be developed.
To address degradation and ageing of metals in HTGRs, the effects of high-temperature helium
with impurities, including oxygen, at levels present in HTGRs need to be evaluated with respect to95

stress corrosion crack initiation and growth rate, crevice corrosion crack initiation and growth rate, andcyclic crack growth rate. Low levels of impurities in high-temperature, high-purity aqueous
environments are know to cause these types of degradation and to accelerate the crack growth rates.
The potential exists for these phenomena to occur in a high-temperature helium environment with lowlevels of impurities.
Many alloys undergo solid state transformation and precipitation during elevated temperature
exposures. These transformation reactions are known as ageing and can lead to embrittlement of thealloy. Ageing and embrittlement occurs, for example, in cast stainless steel components undertemperatures and time conditions experienced in operating LWRs. At the operating temperatures of
HTGRs, the reaction rates can be much higher (i.e., the ageing and embrittlement would occur
sooner). The different alloys and higher temperatures of HTGRs would indicate potentially different
ageing reactions and mechanisms, some of which could occur relatively rapidly and render thematerial embrittled and susceptible to cracking. The ageing reactions, as a function of time andtemperature, in the different alloys used in important components of HTGRs need to be studied toestablish the potential for material property degradation and embrittlement during the operating
lifetime of HTGRs.
Another solid state reaction that occurs in stainless steels (and austenitic alloys) is called
sensitisation. Sensitisation is caused by the precipitation of chromium carbides at the grain boundariesof the stainless steel. This precipitation normally occurs during slow cooling of the metal through hightemperatures, such as when cooling from the high temperatures associated with welding. Formation of
the carbides depletes the chromium from the grain boundary areas, rendering the stainless steel
susceptible to intergranular stress corrosion cracking (cracking along the grain boundaries) inoxidising and impurity environments. The sensitisation rate is exponential with temperature, and at thehigher operating temperatures of HTGRs, there is a potential for sensitisation during the lifetime ofthese plants, thus rendering the stainless steel components susceptible to stress corrosion cracking.
In some HTGR designs, the connecting pipe which carries hot helium from the core to the power
conversion system is treated as a vessel because this pipe is designed, fabricated, and inspected to thesame rules as a reactor pressure vessel. The consequence of this assumption is that a design-basis
double-ended break is not considered for the connecting pipe, and therefore no mitigating systems areincorporated in the design. Considering this pipe as a vessel will require further investigation, because
the pipe is of a much smaller diameter and therefore possesses a much thinner wall than a reactor
pressure vessel designed to the same working pressure. If an unexpected degradation mechanism
should initiate in the pipe, because of the thin wall, it can propagate through the wall in a relatively
short time and possibly not be detected by ISI. Conversely, if an unexpected degradation mechanism
were to initiate in a pressure vessel, it would require a long time to propagate through the greater wall
thickness, allowing enough time to be detected by ISI.
Carburisation, decarburisation, and oxidation of metals in HTGRs are other phenomena that canlead to degradation caused by the operating gaseous and particulate environment. Carburisation is aphenomenon where carbon, either as a particulate or from carbon containing gases, diffuses into steel
to form a surface layer with high carbon content. This surface layer may be hard and brittle, and havehigher strength than the substrate. Differences in strength and other physical properties between thesurface layer and substrate may lead to high stresses in the surface layer when the component is underload. In addition, carbides may form in the high carbon surface layer of stainless steel leaving thematrix depleted of chromium and susceptible to stress corrosion cracking and oxidation. Cracking,stress corrosion cracking, and oxidation can more easily develop in the surface layer which could thenpropagate into the component.

96Decarburisation is a process whereby carbon is depleted from the steel depending on thecomposition of the gaseous environment. Depletion of carbon results in a softer steel and in reduced
fatigue and creep lives. The presence of oxygen results in the formation of scale and general corrosionof metallic components. More importantly it can oxidise the graphite and render metallic components
susceptible to stress corrosion cracking. To control the phenomena of carburisation, decarburisation,and oxidation, a very careful control of the level of different impurities in the coolant is required.
Further, conditions that lead to avoidance of one of the above phenomena can lead to development ofanother. For example, to avoid carburisation, some HTGRs might use slightly oxidising conditions
created by the addition of oxygen to the gas stream. However, this can lead to oxidation of graphite,
general corrosion of metals, and an increased susceptibility to stress corrosion cracking. Some
research has been conducted to study the phenomena described above; however, additional
confirmatory research is needed to better define the conditions under which the phenomena occur forimportant metallic components of HTGRs. In addition, much of the previous research did not include
oxygen in the gaseous environment. Since oxygen may be present in HTGRs at high enough levels toaffect the progression of the above phenomena and to rescue fatigue life, creep life, and resistance tostress corrosion cracking, oxygen needs to be included n new experimental studies.
A number of potential issues related to the inspection of HTGR reactor components exist.
Because some of these reactors are designed to operate for long periods of time between scheduled
shutdowns for maintenance or refuelling, intervals between ISI may be long and the amount ofinspection conducted limited. Therefore, the effectiveness of various ISI programmes as a function of
the frequency of inspections and the number and types of components inspected needs to be evaluated.
Additionally, many internal components are not easily accessible for ISI, and the impact of notinspecting these components needs to be assessed. An alternative to conducting periodic ISIs during
reactor shutdowns is to conduct continuous online, nondestructive monitoring for structural integrity
and leakage detection of the entire reactor or reactor components during operation. Techniques forcontinuous monitoring have been developed, validated, and codified for use in LWRs. If ISI for
HTGRs cannot be conducted on a frequent enough basis and certain components cannot be inspected,
then continuous monitoring may become necessary. The continuous monitoring techniques need to beevaluated and validated for the materials, environments, and degradation mechanisms of the HTGR.
In summary, the technical issues that need to be addressed for metallic structures are:
(1) Availability and applicability of national codes and standards for design and fabrication of
metallic components for service in HTGR high-temperature helium environments.
(2) Lack of appropriate databases for calculating fatigue, creep, and creep-fatigue interaction
lifetimes of components in high-temperature helium on degradation of components.
(3) Ageing behaviour of alloys during elevated temperature exposures.
(4) Sensitisation of austenitic alloys.
(5) Treatment of pipe as a vessel.
(6) Degradation by carburisation, decarburisation, and oxidation of metals in HTGRs.
(7) Issues related to inspection of HTGR reactor components.
For information related to recent and on-going work in the field of high temperature materials,
the OECD Proceedings from a September 1999 conference on Survey on Basic Studies in the Field ofHigh Temperature Engineering are a good reference.
Concrete structures
In HTGRs, concrete structures may be subjected to sustained high temperature. Research isneeded to accumulate and expand existing data on the effects of high temperatures on the properties of97

concrete. The objective of additional research would be to investigate the change in concrete
properties when it is subjected to sustained high temperatures. In the current American Concrete
Institute (ACI) Code, the temperature limits specified for concrete are 150(F for long term, 200(F fornormal use, and 300(F for abnormal conditions.
The operating temperatures of HTGR reactor vessels may be greater than those for currently
licensed nuclear power reactors. Therefore, depending on the effectiveness of the reactor vessel
insulation and cooling system, the concrete reactor building could experience a high-temperatureenvironment. Elevated temperatures can reduce the strength of concrete due to additional shrinkage
effects, as well as cause degradations such as cracking and spalling.
Additional research would include data accumulation and expansion of existing data bases.
Significant information regarding high-temperature effects is available in the literature, including
journals, conference transactions, and proceedings. Earlier research on LWR severe accidents
conducted by Sandia National Laboratories also accumulated significant data on the effects of hightemperatures on the properties of concrete. Oak Ridge National Laboratory has also assembled
information on concrete subjected to high temperature. Lessons learned from facilities at whichconcrete was found to be subjected to high temperatures for long periods of time should also be
investigated and used.

98Chapter 4
CONCLUSIONS AND RECOMMENDATIONS

4.1 Introduction
This chapter provides the SFEAR group’s conclusions and recommendations regarding key
facilities unique to nuclear safety research in danger of being lost in the short term (next 1-2 years)

and those that should be monitored by CSNI in the longer term (> 2 years) to ensure a minimum
facility infrastructure to support operating LWRs and PHWRs in member countries and the licensingof ALWRs and APHWRs. This chapter also provides some general conclusions and recommendations(independent of short or long term) for CSNI consideration.
The overall strategy employed in developing recommendations was to identify minimum research
infrastructure needs and the facilities that should be maintained to ensure that infrastructure isavailable. The group took an approach that focused on those facilities with unique capabilities andrepresented 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.
This approach was considered the most practical given the large number of facilities discussed in
Section 3.1 and the limited resources available for research.
Specifically, the following factors have been considered in determining whether or not torecommend CSNI action to preserve a facility or to recommend longer term monitoring of status.
x

The importance of the facility to resolving the identified safety issues, based upon therelative ranking of a facility within a given technical area using the ranking approachdescribed in the introduction to Chapter 3.5x

The versatility of the facility.

x

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 cost and high relative rankings were consideredcandidates for CSNI action. The conclusions and recommendations are summarised below, organisedby the 5 technical areas discussed in Section 3.1. No conclusions or recommendations are provided for
Sections 3.2 or 3.3, since the information therein is for information only.
4.2 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 SFEAR activity and5.

It is important to note that the relative ranking is only valid within a given technical area. Comparison ofrelative ranking values between technical areas should not be done.

99desire to develop a practical set of recommendations with facility preservation being a coordinatedeffort among the NEA standing committees. Specific general conclusions and recommendations arelisted below:x

CSNI efforts at facility preservation should focus on large facilities, whose loss would mean
the loss of unique capability as well as the loss of substantial investment that in the current
climate of tight resources, would not likely be replaced. Such preservation also includes
maintaining the expertise, knowledge, capabilities and personnel essential to infrastructure
preservation. In this regard, it should be noted that due to previous CSNI efforts, several
large facilities (i.e. PANDA, PKL, MACE, ROSA) have been kept active over the past
5 years, thus helping the current SFEAR effort. However, many large, expensive and uniquefacilities are projected to close over the next 1-5 years. Examples include thermal-hydraulicand severe accident facilities. In addition, many of the test reactors are old and will reach
their end of life without substantial refurbishment. The loss of such facilities would severely
detract from the nuclear safety research infrastructure. Additional discussion on a strategy
for long-term facility preservation is discussed in item c) below.

x

The NEA-Nuclear Science Committee (NSC) should take the lead to monitor the status ofand make recommendations for actions to preserve key facilities in the reactor physics area.
The facilities and information in this report in the reactor physics area represent the SESAR
group’s views on the safety issues and facilities important to nuclear safety research and are
for NSC use in carrying out this responsibility.

x

To help stimulate industry interest in facility and infrastructure preservation, it isrecommended that both CSNI and CNRA take steps to encourage industry co-operation byemphasizing 1) the responsibility of industry to develop sufficient data to support theirapplications, 2) the benefits of co-operative research and 3) the value of preserving criticalresearch infrastructure.

x

Hot cells and autoclaves are essential to nuclear safety research. However, due to the large
number of hot cells and autoclaves, it is impractical for CSNI to monitor their status.
Accordingly, each country should monitor the status of these facilities and bring to CSNI’s
attention any concerns regarding loss of critical infrastructure.

x

Certain safety issues have no large-scale facilities identified for the conduct of relevant
research. The appropriate CSNI Working Groups should evaluate whether or not large-scale
facilities are needed to support resolution of these issues. The issues that fall in this category
are:
–
–
–
–

ECCS strainer clogging (Thermal-Hydraulic issue #6).
3-D core flow distribution (Thermal-Hydraulic issue #12).
Long-term behaviour of concrete structures (Structural Integrity issue #7).
Flow induced vibrations (Structural Integrity issue #9).

4.3 Short-term conclusions and recommendations
The following recommendations are directed toward those actions that CSNI could take in the shortterm (2006-2007) to prevent the loss of key facilities in imminent danger of closure. To assess short-term
concerns, the facilities in Section 3.1 were examined as to which ones are in danger of being shut downin the next 1-2 years. For the purposes of this report, that time period includes through year 2007. Thesefacilities are shown in Table 4.1, by technical area, along with the issues and reactors they support. Also
shown are their replacement cost, and relative ranking scores for that technical area. As can be seen from
Table 4.1 there are a large number of facilities in danger of being shut down in the next 1-2 years.
Discussed below, by technical area, are the short-term conclusions and recommendations for CSNI
100

consideration. These are based upon trying to preserve those facilities having high relative ranking andversatility that would be hard to replace, if lost.
In the thermal-hydraulics area, four facilities are in short-term danger. Two of these facilities
support PWR thermal-hydraulic work (PKL and APEX). However, there are other facilities for PWRT/H work not in short-term danger (e.g., ROSA). Thus no recommendation for short-term action isneeded for PWR T/H facilities. For BWR T/H facilities, both existing large integral BWR thermalhydraulic test facilities (PANDA and PUMA) are in danger of being closed in the next 1-2 years.
These facilities are unique and expensive and at least one should be maintained to be available forsupporting research related to current or future BWR safety issues. Accordingly, preservation of one
integral BWR thermal-hydraulic test facility (either PANDA or PUMA) is considered essential forpreserving a BWR thermal-hydraulic research infrastructure. SESAR is of the view that PANDA is thepreferred facility for preservation due to its scale, replacement cost and versatility (i.e. it is useful inthe severe accident as well as thermal-hydraulic area). Accordingly, CSNI action is recommended in
the short term to support a co-operative research programme in PANDA. It should be noted that CSNI
actions resulting from the SESAR/FAP report played a major role in the preservation of PANDA overthe past 5 years.
In the severe accident area, most facilities supporting the resolution of the following safety issues
for BWRs, PWRs, VVERs and ALWR are in danger in the short term:xxx

Pre-core melt conditions.
Combustible gas control.
Coolability of over-heated cores.

Based upon a review of the facilities in short-term danger, listed in Table 4-1, the groupconcluded that the following facilities should be preserved due to their importance to resolution of theabove issues (as illustrated by their high relative ranking), replacement cost, versatility, and value inlong-term infrastructure preservation.
xx
x

PHEBUS.
QUENCH.
MISTRA.

Each of these is discussed further below.
PHEBUS is a unique facility representing a substantial financial investment, capabilities andexpertise. Due to the high cost of its operation and the long timeframe necessary to plan and conduct
experiments, it is not considered practical to propose that the CSNI organise a co-operative research
programme in PHEBUS. Accordingly, it is recommended that PHEBUS be treated as a special case,with the French authorities taking the lead to propose and organise a future research programme using
PHEBUS. In this regard it should be noted that a PHEBUS Expert Group has been organised to assess
future experimental programmes in PHEBUS in the areas of LOCA (fuel response to LOCAs) and ofsevere accidents (fuel degradation and fission product release and transport). Both separate affects andintegral tests are included in the assessment. The recommendations from this group should provide
valuable input for justifying and planning future programmes in PHEBUS.
The QUENCH facility has been used extensively in the past to investigate pre-core melt
conditions in LWRs. Although it is a unique facility in near-term danger, the group has concluded thatany effort to preserve it for the long term should be dependent upon identifying a future experimental
programme that can provide useful information beyond what has already been done in QUENCH.
Accordingly, it is recommended that QUENCH be treated as a special case, with the German
101authorities taking the lead to propose a future research programme in QUENCH capable of generating
useful new information. In this regard, it is also recommended that the appropriate CSNI Working
Group (WGAMA) be requested to consider future uses for QUENCH and provide a recommendation
to CSNI that can be factored into deliberations on the future of QUENCH.
The MISTRA facility has the capability for conducting experiments on combustible gas mixing
and transport in multi-compartmental configurations with detailed instrumentation and helium as asimulant for H2. As such, it can measure 3-D effects useful for assessing 3-D analytical tools.
MISTRA complements the THAI facility which uses H2 and can conduct experiments on H2
combustion and aerosol distribution. THAI is not in near-term danger and is recommended forlong-term preservation (see below). Accordingly, it is recommended that CSNI take action to reserve
the MISTRA facility (so as to maintain the complementary infrastructure and expertise) by organisingand conducting a co-operative research programme in MISTRA.
In the other technical areas (fuels, and integrity of equipment and structures) no short-term CSNI
actions are recommended (i.e., note, PHEBUS was being addressed under severe accidents).
It should be recognised that implementation of the above recommendations are dependent uponinterest and commitment of the “host countries” to provide sufficient resources to attract participationof other interested parties and the ability to propose experimental programmes relevant to resolution of
the issues and of interest to member countries.
4.4 Longer-term conclusions and recommendations
In the longer term (beyond 2007), it is recommended that CSNI adopt a strategy for thepreservation of a research facility infrastructure, based upon preserving unique, versatile and hard to
replace facilities. The number and nature of these facilities should be based upon supporting currently
operating LWRs and PHWRs and the licensing of future ALWRs and APHWRs. The strategy should
include consideration of short and long-term priorities, cost of preservation (e.g., would the cost ofpreservation detract substantially from other programmes/facilities) and contingency plans in case of
facility loss.
In this regard, many of the factors used in the report to arrive at conclusions andrecommendations could be useful in developing a long-term strategy for assessing and initiating future
co-operative research projects. These factors are:
x
x
x
x
x

x
x

Cost of facility operation and replacement (i.e., limit CSNI involvement to large facilities
needing multi-national 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).
Relative priority, if there are multiple co-operative programmes proposed.
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 programme on
resources available for other projects) and the host country’s long-term plans for the facility.
Industry participation.
Host country commitment.

102

It is recommended that CSNI consider factors such as the above in developing a strategy forfacility preservation and in assessing and initiating future co-operative research programmes.
In addition, critical research capabilities and expertise are defined qualitatively in the
OECD/NEA 2004 publication Collective Statement Concerning Nuclear Safety Research. Using thispublication and the safety issues contained in Section 3.1 of this report, a table of critical researchfacility infrastructure needs was developed along with a list, by reactor type, of existing facilities that
could fulfil those needs. These are shown as Tables 4-2 and 4-3, respectively. The facilities listed inTable 4-3 are those considered unique, hard to replace and identified as having high relative
importance in their technical area, as discussed in Chapter 3 of this report. Accordingly, it isrecommended that CSNI focus on these facilities in developing a strategy for long-term infrastructure
preservation. CSNI should monitor the status of these facilities in the longer term with a goal of taking
action, as appropriate, to ensure that critical facilities are available for each reactor type to meet thecritical research infrastructure needs. In addition, for investigating safety issues associated with newreactors and technologies, CSNI should take an active role in encouraging and organising co-operative
research efforts. This will also contribute to infrastructure preservation. Similar to the short-term
recommendations above, host country interest will be an important factor in determining whichfacilities to preserve.
It should be noted that the SESAR/FAP report recommended that in the long-term thermal
hydraulic facilities for each major reactor type should be maintained in North America, Europe andAsia. However, given the current situation with respect to safety research programme funding, theSFEAR group is of the opinion that this recommendation is no longer practical and recommends thatthe long-term strategy for facility preservation focus on ensuring at least one thermal-hydraulic facility
for each reactor type be maintained worldwide.
Finally, it should be noted that test reactors (TRs) have for several decades supported thedevelopment and safety of nuclear power plants. However, most existing TRs have been in operation
for a considerable period of time. This is the case for all relevant TRs in Europe, for which theoperational lifetime is currently in the range 40-50 years. With the exception of the HANARO test
reactor in the Republic of Korea, which has been in operation for about 10 years, the situation in
Japan, in the U.S. and Canada, is practically the same as in Europe, i.e., the current TRs have,considerable age. Although one foresees that some of these reactors might continue operation forseveral years ahead, it is apparent that there will be a need to gradually update or supplement theexisting ageing test reactors in the years to come. Due to the critical nature of test reactors to nuclear
safety research and the fact that they are the most costly to replace, particular attention should be paid
to test reactor preservation.
Table 4.1. Facilities in danger in the short term (2006-2007)
Facility

Applicable
reactors

Thermal-hydraulics
PKL
PWR(Germany)

Issues addressed
1.
2.
3.
4.
8.
9.
13.

Boron dilution.
Passive safety system performance.
Non-pipe breaks.
SGTR
Two phase natural circulation.
Thermal stratification.
Flow distribution in cold legs anddowncomers.
14. Accidents initiated during shutdown.

103

Replacement Relative
cost
rankingHigh

2.9

Table 4.1. Facilities in danger in the short term (2006-2007) (Cont’d)
Facility

Applicable
reactors

Thermal-hydraulics
APEX
PWR, ALWR
(USA)

PUMA
(USA)

BWR, ALWR

PANDA(Switzerland)

PWR, BWR,
ALWR

Issues addressed
1.
2.
4.
8.
9.
14.
2.
5.
8.
9.
14.
2
8.
9.
10.
14.

Boron dilution.
Passive safety system performance.
SGTR.
Two phase – natural circulation.
Thermal stratification.
Accidents initiated during shutdown.
Passive safety system performance.
Stability and power oscillations.
Two-phase natural circulation.
Thermal stratification.
Accidents initiated during shutdown.
Passive safety system performance.
Two-phase natural circulation.
Thermal stratification.
thermal cycling.
Accidents initiated during shutdown.

Fuels
PHEBUS

PWR, BWR,
1. Response to LOCAs.
VVER, ALWR
Reactor physics: NSC lead no near-term concerns
Severe accidents
CTF (Canada)All
12. Combustible gas control.
LSVCTF (Canada) All
12. Combustible gas control.
ARTIST
PWR, PHWR, 13. FP chemistry and release.
APHWR,ALWR
VULCANO
PWR, BWR,
9. Ex-vessel melt progression.
VVER ALWR 10. Core-concrete interaction.
PHEBUS
BWR,PWR,VV 1. Pre-core melt conditions.
(France)
ER,ALWR,PH 2. In-vessel melt progression.
WR,APHWR 4. Effect of air on core melt progression.
5. Effect of HB and MOX Fuel.
13. FP chemistry and release.
17. Coolability of overheated cores.
MISTRA
PWR, BWR,
12. Combustible gas control.
(France)
VVER, ALWR 15. Containment integrity.
PANDA
BWR, ALWR 12. Combustible gas control.
(Switzerland)
15. Containment integrity.
QUENCH
PWR, BWR,
1. Pre-core melt conditions.
(Germany)
VVER, ALWR 2. In-vessel melt progression.
17. Coolability of overheated cores.
COMET-FZK
PWR, BWR,
10. Core concrete interaction.
(Germany)
VVER, ALWR
Integrity of Equipment and Structures – No near-term concerns
Notes:
Replacement cost:High
Medium

10 Million USD.
2-10 Million USD.

104

Replacement Relative
cost
rankingHigh

1.6

Medium

1.6

High

1.9

High

1.0

MediumMediumMedium

0.6
0.6
0.6

High

1.20

High

2.3

High

0.7

High

1.0

Medium1.0

Medium0.4

Table 4-2. Critical research facility infrastructure needs
Technical expertise needed
Thermal-hydraulics:
modelling and analysis

Fuels:
performance andphenomena

Reactor physics:
modelling, cross sections,parameters and analysis
Severe accidents:
phenomena, progression,modelling and analysis
Integrity of equipment andstructures:materials behaviour,
structural design

Facility capability needs
Large-scale integral test facility foreach reactor type.

Important factors for facilities

Scale, temperature and pressure capability
are key factors. Also, the completeness ofthe facility with respect to factors such as:
auxiliary systems, number of loops andinstrumentation capability are important.
Test reactor for steady state andAbility to achieve representative values ofreactivity insertion testing. Hot cell energy deposition in transient testing.
for PIE and simulated LOCA testing. Adequate linear heat rating, burn-up andin-core instrumentation for steady state
testing. Ability to do experiments with
MOX and high burn-up fuel. Hot cells forfull length and pin segment PIE.
Ability to do experiments with MOX andCritical facility for measuringhigh burn-up fuel.
physics parameters and performing
benchmark experiments.
In-reactor or ex-reactor testing of FP Use of prototypic materials and large
release and transport, core debris
scale are important.
cooling, combustible gas control andAM strategies.
Test reactors for irradiating material Ability to achieve fluence and othersamples under controlled conditions. prototypic conditions (e.g. temperature,
simulate impurities, stress, etc.). Hot cells
Hot cells for examining large andand autoclaves for ex-reactor testing of
small irradiated material samples.
irradiated materials.
Autoclaves for material testing.

105

Table 4.3. Critical facilities to be monitored in the long term
Technical area

BWR
1

PWR

VVER

Thermalhydraulics

PANDA

LSTF/ROSA
PKL
ATLAS

PSB-VVERPACTEL

Fuels2

HaldenNSRR
PHEBUS

HaldenNSRRCABRIPHEBUS
Proteus
Venus

HaldenMIRCABRIPHEBUS
Proteus
LR-O

Reactor physics* Proteus
Venus

Severe accidents Integral testing
PHEBUS1
PHEBUS1
In-vessel phenomena– QUENCH1– QUENCH1– VERDON– VERDON– KROTOS– KROTOS

Integrity ofequipment andstructures3
*

PHWR/
APHWR
RD-14-M

HaldenNRU

Proteus
ZED-2
Venus

ALWR
LSTF/ROSAPKLPANDA1ATLAS
HaldenNSRRCABRIPHEBUS
ProteusVenus

PHEBUS1

PHEBUS1

PHEBUS1– QUENCH1– VERDON– KROTOS– Fuel Channel
Safety Facility– VERDON– KROTOS– QUENCH1– VERDON– KROTOS

Ex-vessel phenomena– MCCI– MCCI– MCCI– MCCI– MCCI– VULCANO – VULCANO – VULCANO – VULCANO– VULCANO– THAI2– THAI2– THAI2– THAI2– THAI2
44
44– KROTOS– KROTOS– KROTOS– KROTOS– KROTOS4
Containment mixing/combustion– PANDA1– PANDA1– PANDA1– PANDA1– PANDA1– LSCF– LSCF– LSCF– LSCF– LSCF– THAI2– THAI2– THAI2– THAI2– THAI2
1
1
1
1– MISTRA– MISTRA– MISTRA– MISTRA– MISTRA1
Accident management
Uses data generated in the resolution of other severe accident and thermal-hydraulic issues.
No unique facility needs
HaldenHaldenHaldenHaldenHalden
JMTR
NRU
JMTR
JMTR
JMTR
LVR-15
LVR-15
LVR-15
LVR-15
LVR-15ATR
ATR
ATR(Included for completeness; NSC to monitor status).

Notes:
1. Assumes actions will be taken in the short term to preserve these facilities.
2. Assumes on-going effort to initiate a co-operative research programme will be successful.
3. Due to the large number of hot cells and autoclaves, each country should monitor the status and identify
concerns.
4. Experimental programme under discussion in the CSNI-SERENA programme.

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