Operational Safety

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22 févr. 2014 (il y a 3 années et 3 mois)

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1


EU Research in “Operational Safety of Existing Installations”

under the Nuclear Fission Programme 1998
-
2002


G. Van Goethem, A. Zurita, P. Manolatos, J. Martin Bermejo and S. Casalta


European Commission, DG Research
-

Dir. J: Energy

Unit 4: Nuclear Fis
sion and Radiation Protection // 1049 Brussels, Belgium

Phone: +32
-
2
-
2951424, Fax: +32
-
2
-
2954991, E
-
mail:georges.van
-
goethem@cec.eu.int



A
BSTRACT

In this paper an overview is given of the most important aspects of the research activities
organised by the
European Union (EU) in the area of reactor safety under the current 5th
Euratom Framework Programme 1998
-
2002 (FP
-
5). This area is focussing on "Operational
Safety of Existing Installations". The fundamental safety objective for nuclear power plants
(NPPs)

consists in protecting the public and the environment from the harmful effects resulting
from ionising radiations. Community research with this objective is carried out through both
"indirect actions” (organised by DG Research) and “direct actions” (carri
ed out by DG Joint
Research Centre / JRC). The mid
-
term achievements of this area were discussed at the
symposium FISA
-
2001 (EC Luxembourg, November 12
-
14, 2001 / 750 pages, EUR 20281 EN,
OPOCE Luxembourg 2002). This research area is actually part of the
FP
-
5 Key Action
NUCLEAR FISSION, which consists of the following 4 areas: reactor safety; waste
management (including partitioning and transmutation); future systems (including high
temperature reactors); and radiation protection.


More specifically, this

paper deals with the strategy, the organisation and the main achievements
of the 73 multi
-
partner projects cosponsored by the European Union as "indirect actions"
(shared
-
cost and concerted actions). These projects are organised in 3 clusters, each devote
d to
one key safety issue. Each cluster is treated in a separate section of this paper, namely: (1) plant
life extension and management (PLEM cluster); (2) severe accident management (SAM cluster)
and (3) evolutionary safety concepts (EVOL cluster). The to
tal cost of the “indirect actions” of
this Community research area is approximately € 82.5 million, out of which € 43 million is
contributed by the EU budget. At FISA
-
2001, only the “indirect actions” that started before
January 1, 2001, were formally pres
ented, i.e. a total of 41 projects


the 32 more recent multi
-
partner projects were discussed whenever it was felt appropriate.


INTRODUCTION / Operational Safety of Existing Installations


As defined by the IAEA (1993), the “General Nuclear Safety Object
ive”, underlying all reactor
design and safety activities, remains that of “protecting individuals, society and the environment
from harm by establishing and maintaining in nuclear installations effective defences against
radiological hazards”. However, as

with all engineering constructions having an inherent
capacity to cause hazards to people and environment, the safety of nuclear installations does not
rely only on “effective defences” of the technical type (T) but also on the interaction of these
techni
ques with men’s attitudes (M), and organisational measures (O). Therefore the so
-
called
MTO factor has become increasingly important in plant safety management: it is nowadays
recognised as the very basis of the reactor safety culture.





2

In the past Eura
tom research activities devoted to operational safety, the emphasis was put on
the T component of the MTO factor, i.e. the identification and solution of technological
problems. Traditionally the technological problems of nuclear reactor safety are related

to the 3
basic safety functions, namely: controlling the power, cooling the fuel and confining the
radioactive material. These problems have received a standard solution. As a result from long
standing industrial and theoretical research programmes, inclu
ding operational feedback, the 3
-
levels defence
-
in
-
depth strategy was developed against accidental radioactivity releases, i.e.: (1)
prevention of abnormal operation and failures (quality control), (2) control of abnormal
operation and detection of failure
s (surveillance and protection), and (3) control of accidents
within the design basis (safeguards systems). Usually associated with this approach are the
following concepts: (1) the multiple barrier design for the confinement of radioactive material,
(2) t
he protection and safeguard systems to ensure the integrity of the barriers, and (3) the
regulatory procedures (for example, the safety analysis reports) to ensure the health and safety of
the plant workers and of the population.



The EU co
-
sponsored “ind
irect actions” discussed here are related more specifically to the
above
-
mentioned level (1) of the defence
-
in
-
depth and to a new level, say (4), in line with the
increasingly stringent safety requirements, especially for the next generation of reactors. A
s far
as level (1) is concerned, a cluster of thirty
-
three FP
-
5 projects, called
PLEM

(“Plant Life
Extension and Management”) is focusing on the following items: integrity of equipment and
structure; on
-
line monitoring and maintenance; organisation and man
agement of safety. As far as
level (4) is concerned, another cluster of twenty
-
two FP
-
5 projects, called
SAM

(“Severe
Accident Management”), is focusing on the following items: assessment of severe accident risks;
and severe accident management measures. F
inally, the activities dealing with evolutionary
concepts are put together in the cluster of eightteen FP
-
5 projects, called
EVOL

(“Evolutionary
Concepts”), focusing on the following items: evolutionary safety concepts (using, for example,
passive safety s
ystems); and advanced fuel technologies such as high burn
-
up and MOX fuel.


Knowledge management (databanks and eurocourses)


It is worth presenting first of all three projects of general interest that are dealing with
management of knowledge, stretching
across the three clusters PLEM, SAM and EVOL.




JSRI
(Joint Safety Research Index), a concerted action concerning the dissemination of
information about European reactor safety research programmes and their main
achievements in all above mentioned areas (
http://w2ksrvx.ike.uni
-
stuttgart.de/jsri/
).




ENEN

(European Nuclear Education Network), an

accompanying measure dealing with the
creation of a European higher area space in the field of nuclear engineering (homepage
http://www3.sckcen.be/enen/workplan.html
). It is aimed at proposing a global network
strategy and at performing pilot education sessions, thereby contributing to the conservation
of the nuclear knowledge and expertise, in co
-
operation wit
h other international organisations
such as IAEA (homepage

http://www.iaea.org/km
)
and OECD/NEA (homepage
http://www.nea.fr
)




CERTA,
a network seeking to provide a consolidated framework for the preservation of the
int
egral system experimental data bases for reactor thermal
-
hydraulic safety analysis acquired
in the context of research carried out by European institutional and industrial research
organizations. It includes the main experimental programmes and databases r
elevant to
reactors in operation in Europe (homepage
http://lunar.jrc.it/stresaWebSite
/)
.





3

Worth mentioning also are a series of Eurocourses, co
-
organised by DG Research and national
hosting organisations, with the aim to disseminate specific research r
esults and/or to discuss the
latest achievements in a given area, thereby contributing to the objectives of “education and
training”, that are becoming increasingly important. Here is the list of Eurocourses organised
under FP
-
5:

-

in the area of PLEM :



M
ASC: “Use and application of the master curve method for determining fracture
toughness” (co
-
organised by VTT, Helsinki, 12
-
14/06/02)



IPC: “Integrity of pressurised components of nuclear power plants” (co
-
organised by GRS,
Cologne, 17
-
21/09/01)



CRACHS: “Co
nsensus on Reconstitution Techniques and Fracture Toughness Analysis of
Charpy
-

Type specimens” (co
-
organised by SCK
-
CEN, Mol, 5
-
7/09/01)

-

in the area of SAM :



CORIUM : “European training course on corium” (co
-
organised by CEA, Cadarache, 27
-
31/01/03


home page
http://www.cad.cea.fr/Eurocourse/EUROCOURSE
-
en.htm
)

-

in the area of EVOL :



PSARID : “Probabilistic safety assessment and risk
-
informed decision making” (co
-
organised by GRS, Garching, 5
-
9/03/01)



SMIRT
-
17 : “17th International conference on str
uctural mechanics in reactor technology”
(co
-
organised by Brno University of Technology, Czech republic, Prague, 18
-
22/08/03


homepage
http://www.teris.cz/SMiRT17
).


A total of 3 individual Marie Curie grants f
or post
-
PhD students are also part of the programme.


Besides the "indirect actions" described above and co
-
ordinated by DG Research, it is worth
recalling the "direct actions" in nuclear fission carried out by DG Joint Research Centre under
the common Eur
atom umbrella (recall that JRC consists of 5 establishments in 5 EU countries).
The following web site provides information about JRC activities:
http://www.jrc.cec.eu.int/
.


One of the main actions of DG
JRC in reactor safety research consists in operating European
networks in the field of structural integrity for nuclear components, namely

: the network on
Ageing Materials European Strategy (AMES


homepage http://www.jrc.nl/ames), the European
Network fo
r Inspection Qualification (ENIQ


homepage http://www.jrc.nl/eniq) and the
Network for Evaluating Structural Components (NESC


homepage
http://nesc.jrc.nl/
). Each of
these networks deals with a specific aspect of fitness for service of materials in struct
ural
components. These JRC networks have developed co
-
operative programmes, bringing together
several European institutions and organisations including utilities, manufacturers, engineering
companies, research and development laboratories, regulatory bodie
s and the JRC Institute for
Energy (Petten, NL). For more information about the DG JRC «

direct actions

», the reader is
referred to the above web site.

1

P
LANT
L
IFE
E
XTENSION AND
M
ANAGEMENT
(C
LUSTER
PLEM



SEE
T
ABLE
1)

At the end of 2000, in the EU, in 8

of the 15 Member States, a total of 143 reactor units were in
operation with a total capacity of 123 net GWe and a total gross generation of 863 TWh (i.e. 35
% of the EU electricity generation), representing a cumulated experience of 3886 reactor
-
years.
W
orth mentioning, in particular, is the large number of nuclear reactors that have been operating
for longer than 20 years. In the EU only, a total of 65 reactor units were put in commercial
operation before 1980 (and a total of 78 after this date). As a co
nsequence, the nuclear industry
is increasingly interested in research activities aimed at better understanding and managing
aging phenomena (i.e. changes in microstructural and mechanical properties due to irradiation,
etc.). More importantly, the optimis
ation of the operational conditions of aged reactors (using,
for example, appropriate prediction tools for evaluating the safety margins) and the decision

4

process about plant life management (involving, for example, replacement of equipment) are
becoming k
ey issues for those in charge of plant safety and performance.


Despite the historical safety record of NPPs of Western design, the key players believe that
research is still needed to further increase both the safety and the performances of these power
pl
ants in line with the steadily growing pressure of regulatory and market forces as well as of the
public opinion. Looking at the future of these plants, there is a natural drive to extend their
lifetime where this can be achieved safely, bearing in mind th
at the lifetime of a nuclear plant is
definitely limited by the aging of non
-
replaceable components like the nuclear reactor pressure
vessel (RPV). Knowing that operating license for most countries will broadly expire by 2015
-
2020, many utilities and vendo
rs of equipments nowadays are very active in improving
techniques to ensure both the performances and the safety of their plants until this expiration
date and even beyond it. Indeed extending the life of units is usually considered a worthy
investment, de
pending on license renewal obtention as well as replacement equipment and
operational safety requirements.


In the thirty
-
three projects belonging to the cluster PLEM (
“1. Plant Life Extension and
Management”
), the emphasis of the work programme is on the

reactor’s very high radioactive
inventory which is perceived as a high potential danger for people and environment. Under FP
-
5, the following three key issues have been identified:




(1)

harsh environmental conditions

put on structures and equipments (for

example, reactor
pressure vessels with coolant at temperatures of 290 °C
-

325 °C, pressures up to ~ 15.5 MPa
and end
-
of
-
life irradiation doses of ~ 10
19

n/cm
2
). Research in preventive measures to ensure
plant integrity is being performed in networks, usu
ally run by the Joint Research Centre (in
particular the Energy Institute of Petten, NL), and is becoming increasingly important in FP
-
5, under the section
“1.1 Integrity of Equipment and Structures
”.




(2) radiation zones when carrying out
inspections, te
sts, maintenance and supervision
.
Radiation protection is an important aspect and is naturally a key item in FP
-
5, both in the
Key Action and in Generic Research. The reactor safety relevant aspects of radiation
protection as well as on
-
line monitoring to
ensure the integrity of components and structures
are the subject of the section
“1.2 On
-
line Monitoring and Maintenance”

.




(3)
safety culture
. Routine operations that are regulated by operational procedures, are related
to the M and O components of the
above
-
discussed TMO factor. Special attention is also
devoted to harmonisation of best practices through the production, as project output, of a
number of European handbooks for the benefit of both industry and safety authorities. These
aspects are investi
gated in the section
“1.3 Organisation and Management of Safety”
.


No wonder that the main actors in the area of “plant life extension and management” are the
European utilities, as it is also shown on Table 4, that is: EDF in France; E.ON AG and RWE
AG i
n Germany; British Energy in the UK; ENDESA and IBERDROLA in Spain,
VATTENFALL AB in Sweden; ELECTRABEL in Belgium; and FORTUM in Finland.


Under FP
-
5, the total EU budget to be spent for the 33 projects in the cluster PLEM amounts to
approximately EUR 18
million, which represents roughly half of the total value of these
projects. This might be compared to the total EU budget spent under the 4
th

framework
programme 1994
-
1998 (FP
-
4) for the 11 projects in the AGE cluster, which was EUR 2.1
million.


TABLE 1



5

1.1

Integrity of Equipment and Structures

To prevent any in
-
service failure of the RPV and in general of the reactor cooling system,
stringent operational rules are necessary, based on deep understanding of irradiation
embrittlement and of various types o
f fluid
-
structure interaction effects (chemical as in
corrosion processes or structural as in waterhammer impacts), as well as on accurate evaluation
of safety margins.


Understanding of irradiation embrittlement and EU strategy for materials testing react
ors


One of the most important material properties for ensuring structural integrity is the fracture
toughness, which measures the resistance of the material to the propagation of a hypothetical
sharp crack which in the safety case is (conservatively) assu
med to be present in the pressure
vessel from the time of its construction.


In unirradiated steel, there is a transition, as steel is cooled down, from high toughness to low
toughness. This transition is accompanied by a change in the mode of fracture fro
m ductile to
brittle. As the irradiation dose increases, the fracture toughness decreases, with a shift of the
ductile
-
to
-
brittle transition towards the higher temperatures. At typical operating temperatures, it
is necessary to be in the upper region, away

from any risk of brittle failure. By predicting the
extent of the irradiation shifts it is possible to modify the operating rules, and the temperatures
at which the vessels actually operate, to ensure that they always operate in the ductile region.


Hence
, during operation of a nuclear power plant, changes to the tensile and fracture properties
of the steels (i.e. in particular, the fracture toughness of the pressure vessel material) are
followed as a function of the dose rate received by testing at regula
r intervals surveillance
specimens, which are exposed, inside the pressure vessel in locations close to the vessel wall.
Irradiation embrittlement is the subject of a number of different projects, focusing on reactor
pressure vessel aging aspects.


Within
FRAME,

research is conducted to improve the assessment of the most important
parameters used to measure the embrittlement conditions of the RPV. Currently this is done
through indirect measurements in a rather conservative way (the so
-
called reference tem
perature
methodology, which makes use of Charpy
-
V notch impact testing). The work focuses on the
development of a method, which allows to measure directly the fracture toughness. This should
result in a better accurate estimation of the embrittlement condi
tions of the RPV material.


RETROSPEC

develops procedures and guidelines for retrospective dosimetry, to evaluate the
neutron doses induced in reactor structural materials in those cases where no or unreliable data
from surveillance specimens are availabl
e (e.g. the older generation of VVER
-
440 type
reactors).


The objective of
PISA

is to improve the predictability of the impact that phosphorus segregation
to internal grain boundaries, assisted by irradiation at elevated temperatures, can have on
embrittl
ement in RPV steels.


Most of the materials testing reactors (MTRs) will be more than 40 years old by 2010. The
objective of the thematic network
FEUNMARR

is to determine the future European irradiation
needs in MTRs. The following items are be covered: m
aterials and fuel for current and future
reactors, medical applications and productions, neutron beams for research, and other
applications as neutron radiography and isotope production.



6

As support to the transnational access to research infrastructures,
the project
RENION

(homepage
www.nri.cz
) offers the Czech experimental LR
-
0 reactor to researchers to conduct
experimental projects related to VVER and PWR reactor physics in order to extend their
experimental databases and

to validate computer codes.


Optimisation of operational conditions focusing on corrosion issues and on thermal
-
hydraulics


A review of plant concerns has revealed that besides embrittlement of the RPV, corrosion in
synergy with irradiation becomes an im
portant degradation mechanism especially for internals.
For example, aging of reactor internals is in almost all cases associated with irradiation assisted
stress corrosion cracking (IASCC). For BWRs, in particular, the 2 major corrosion concerns at
prese
nt are cracking of the core shroud/plate and possible cracking in the lower plenum region.


Two projects deal with irradiation assisted corrosion cracking of austenitic steels for reactor
internals. Within
INTERWELD,

the radiation induced damages that prom
ote cracking in the
heat affected zones of PWR and BWR core internal components is studied, focusing on
parameters such as neutron fluence/irradiation conditions, microstructural and microchemical
conditions and giving particular emphasis on the residual s
tresses. Further work is performed in
PRIS

to produce materials data (in particular J
IC
, J
-
R, tensile properties and irradiation induced
microstructural changes) for LWR internals as a function of fluence up to 70 dpa. Within
CASTOC,

environmentally assis
ted corrosion (EAC) of low alloy steels, for RPV, under static
and cyclic conditions is studied with the aim to improve service operation and code
implementation.


It is worth recalling that decisions about repair/replacement/back
-
fitting and operational
p
erformances under the usual high safety standard conditions depend not only upon the
challenges of irradiation or corrosion effects on materials integrity but also upon the dynamic
loadings during operation (generated, for example, by condensation induced
waterhammers in
pipes and open networks). The
WAHALOADS
project

deals with these issues (homepage
http://www.meca.ucl.ac.be/waha)
.



The project

FLOMIX
-
R

aims at performing a set of mixing experiments that are supported by
CFD calculations (homepage
http://www.fz
-
rossendorf.de/FWS/FLOMIX/

). Emphasis is on
slug mixing phenomena relevant for local boron dilution scenarios, and mixing phenomena of
interest for operational issues and thermal fatigue. Impr
oved measurement techniques with
enhanced resolution in time and space are tested.


Prediction of structural safety margins


Bimetallic welds (BMW), connecting ferritic components with austenitic piping are used in
safety class systems of all PWR and BWR
plants. For PWRs, the BMWs of particular interest
are those attaching the piping systems (made of stainless steel) to the various nozzles of the
RPVs, SG and pressurizer. Because of their metallurgy, these weldments are particularly prone
to localised crac
king. The integrity of the BMWs without and with hypothetical cracks has to be
justified in all conditions for the life of the plant. Classical fracture mechanics methods (such as
J
IC

values for the initiation fracture toughness and fracture resistance cur
ves J
-
R for the ductile
crack growth) are difficult to apply to this specific case due to a number of complicating factors
such as the prevailing mixed
-
mode loading conditions, the variation in material constitutive
equations across the weld zone and the p
resence of a large residual stress field. A number of
different projects are focusing on the development of predictive tools.





7

The project
ADIMEW

extends to an industrial scale the work done in BIMET (under FP
-
4) to
predict safety margins for bimetallic
welds. It aims at providing recommendations for codes and
standards on flaw evaluation for Dissimilar Metal Welds (DMW). Dissimilar metal welds,
representative of a 16’’ diameter girth weld of a French pressuriser, have been fabricated with
artificial def
ects, and are tested under operating temperature (300°C) and bending conditions.
Different types of testing and analytical tools (fracture mechanics, residual stress measurement
and calculation, finite element analysis, etc…) are used in order to evaluate
the crack driving
force.


Defect assessment techniques are further improved in
VOCALIST

to better predict safety
margins, in particular with respect to the constraint effect (i.e. the pattern of crack
-
tip stresses
and strains causing plastic flow and fract
ure), which gives rise to an effective toughness for
components higher than that measured on test specimens.


The effect of the Warm Pre
-
Stress (WPS) is studied by
SMILE
. Data are gathered or created by
experimental and numerical work. The objective is to
propose a harmonised way of taking into
account this generic phenomenon in the safety studies.


Events as the Civaux
-
1 incident (May 1998) indicate that certain piping system Tee’s are
susceptible to turbulent temperature mixing effects that cannot be adeq
uately monitored by
common thermocouple instrumentation, putting the reliability of integrity evaluation in doubt.
THERFAT

proposes to review field data and to perform advanced thermohydraulic flaw
simulations and stress and fracture analysis. Critical ele
ments of the procedure are investigated
by targeted verification tests. Proposals are made for improved thermal fatigue assessment
procedures, screening criteria and for establishing a European Methodology on Thermal
Fatigue.


Concrete containment aging


The
MAECENAS

project is developing an advanced analysis tool that allows the structural
integrity of aged, reinforced, pre
-
stressed concrete NPP structures to be assessed in a rational
manner. Using data from laboratory tests under arbitrary multi
-
axial s
tress states, a generalised
thermo
-
mechanical constitutive model for concrete is constructed. A simplified safety
-
cost
analysis tool also is developed to determine appropriate repair or strengthening strategies.


The
CONMOD

project is aimed at improving th
e interaction of NDT and finite element
calculations in order to optimise maintenance activities for concrete containments (homepage
http://
www.conmod.dk
). Emphasis is placed on the identification of possible critical
defects and
damage mechanisms, using laboratory
-
scale and full
-
scale experiments (for example, in
Sweden's Barsebäck
-
1 reactor). It aims also at standardisation of the concrete testing procedures
throughout the EU.



8

1.2

On
-
line Monitoring and Maintenance

In FP
-
4, a concerted action, AMES
-
NDT, was aiming at verifying to what extent non
-
destructive
testing (NDT) techniques can be used to assess material damage. Under FP
-
5, this work is
continued as a full
-
scale shared cost action within
GRETE
. Capability and

reliability of
innovative inspection techniques (ultrasonic, magnetic, thermoelectricity and dynamic indent)
are assessed by means of round robin testing. These non standard techniques aim to detect
microstructural changes in the material leading to degra
dation of the mechanical properties of
the component long before macroscopic cracks are initiated. Applications are focussed, on
irradiation damage in RPVs and thermal fatigue in pipings.


In
SPIQNAR,

attention is devoted to improve performance of ultrason
ic inspection aimed at
detecting and sizing of cracks in structural components. Specific issues addressed are
development of signal processing techniques and a reliable methodology for producing synthetic
defects and “virtual defect signals” to improve ins
pection qualification methodologies.


The
REDOS

project (homepage
http://www.ie.jrc.cec.eu.int/ames/relproj/redos.htm
) aims at
improving dosimetry for irradiated steel and qualifying a methodology for radiation field
parameters monitoring. Benchmarking as
well as combined experimental and computational
techniques are used taking in account deep penetration and space energy dependent radiation
field, complex geometry, as well as gamma irradiation.


In

LIRES
robust

reference electrodes

are under development

to allow on
-
line monitoring of the
corrosion potential (important parameter for IASCC) within the harsh operational conditions of
reactors.


Under FP
-
4, a Risk Based Management Philosophy was proposed by the thematic network
EURIS, which was driven by uti
lities. For example, guidelines for a European framework for
risk informed inspection were developed. These guidelines should be able to identify safety
-
significant categories for power plant components and to optimise the targeting of costly
inspections.
It includes feedback from plant operation and must indicate the specific components
and the locations to be inspected, the defects to be detected and the performance in detection and
sizing to be achieved. The methodology integrates actions or mitigation m
ethods other than
inspection, in order to manage the risk. As a consequence, risk informed in
-
service inspection
(ISI) should reduce the cost and efforts whilst maintaining safety at its currently high level or
above. In the area of inspection qualificatio
n, actually, both regulatory and industry
organisations are taking advantage of the work performed by the JRC operated network ENIQ
(see the series of “handbooks of recommended practices” on their web site).


The
NURBIM

project aims at developing improved
procedures to identify where the highest
likelihood of damage/failure of passive components is located in plant and provide quantitative
measures of the associated risk. It focuses on the definition of best practice methodologies for
performing risk
-
based
analysis and establishing a set of criteria for the acceptance of risk
quantities that can help regulatory bodies in Europe to accept risk based inspection (RBI) as a
valid tool for managing plant safety.


Development of advanced weld repair and assessmen
t technologies will reduce unplanned
outage time and thereby increase the economic performance and safety of the nuclear power
industry.
ENPOWER

aims to develop weld repair procedures and alternative post weld
treatments that minimise residual stresses and

shorten repair time scales. Assessment
methodology for treating defects in residual stress fields will be refined to give more accurate
and informed sentencing of defects in aging plant. Guidelines for optimising weld repair
procedures will be developed t
hat will also profit to other sectors of the industry.


9


1.3

Organisation and Management of Safety

Digital instrumentation

From a technological point of view, one important challenge in NPP modernisation is the
implementation of digital instrumentation an
d control tools (substituting the original analog
systems) and the subsequent training required for the operating staff. Two projects are devoted
to software modernisation for plant safety.
BE
-
SECBS

is dealing with computer
-
based systems
embedded in a nucl
ear installation to support I&C functions important to safety
. CEMSIS

is
aiming at developing a safety justification framework for the refurbishment of systems important
to safety (SIS) that is acceptable to different stakeholders (especially licensing bod
ies and
utilities).


Organisational factors


Under FP
-
4, issues on organisational matters were investigated in several projects, in
cooperation with some of the JRC operated networks. One example is the concerted action
ORFA that looked at organisational f
actors and how they influence nuclear safety. In many
studies it is recognised that organisational factors are often the root cause of incidents and
accidents. However, there is unfortunately no agreed and validated method for their assessment.
Important

issues for short
-
term research are related to the identification and description of those
factors which define good practice, the development of organisational self
-
assessment tools, the
inclusion of organisational factors in incident analysis, the defini
tion of methods of how to
maintain the corporate knowledge, etc.


The main objective of
LEARNSAFE
is to create methods and tools for supporting processes of
organisational learning at the NPP. This has become increasingly important for the nuclear
industr
y in its adaptation to a changing political and economic environment, changing regulatory
requirements, changing work force, changing technology and changing organisation of NPPs
and power utilities. The focus is on the management of change (homepage
http:
//www.vtt.fi/virtual/learnsafe/contact_info/index.htm
).


The objective of the concerted action
SPI

(homepage
http://domino.grs.de/spi/spi.nsf
) is to
review and evaluate the application of safety performance i
ndicators


in combination with other
tools, like PSA


in order to maintain and improve safety of NPPs. It will also seek methods that
can be used in a risk
-
informed regulatory system and environment.


As far as the decision process related to inspection,

maintenance, operation and repair of NPPs
goes, special attention is devoted within the shared
-
cost action
VRIMOR

to innovative support
tools based on virtual reality.

More generally, the objectives of the accompanying measure
EUROSAFE

are to support the

convergence of nuclear and radiological safety practices (safety culture) in Europe, while
developing the idea of a European scientific and technical pool in the fields of reactor safety and
radiation protection.


Emphasis on VVER reactors safety


As the
EU enlargement process is now well engaged, a number of Central and Eastern European
Countries (CEECs) become of particular interest, as far as nuclear power production is
concerned. The Czech Republic, Hungary, Slovakia and Slovenia, which are particularl
y active
in FP
-
5, are operating all together 16 nuclear units (15 VVERs
-
440 and 1 PWR) with a total
capacity of 7.3 net GWe providing 30 % of their electricity.


10


The project
VERSAFE
brings together utilities from some of the Central and Eastern European
C
ountries with the aim to produce common guidelines for the implementation of techniques for
both plant modernisation and severe accident management. The

IMPAM
-
VVER

project is
addressing a safety relevant issue identified in recent studies on VVER safety. I
t investigates
effective means and criteria for primary depressurisation during small loss of coolant accident
(SBLOCA) including feed and bleed operation. The issue was raised using analytical tools but
the resolution requires experimental investigation a
s well as specific computer code validation.
The resulting knowledge will effectively contribute to the safety in all VVER countries.
VERLIFE

is creating a “unified procedure for lifetime assessment of components and piping in
VVER type nuclear power plant
s” based, in a first step, on former Soviet rules and codes. Later
on, a critical analysis of possible application to some PWR type components will be done, with
the aim to harmonise VVER and PWR Codes and Procedures.


Finally,

ATHENA
, the AMES thematic ne
twork on aging, aims, within the enlarged EU, at
reaching a consensus on important issues that have an impact on the life management of nuclear
power plants. ATHENA creates a structure enhancing the collaboration between European
funded R&D, national progr
ams and TACIS/PHARE programs. This will greatly increase the
return from the individual projects and maximise the European added value.

2

S
EVERE
A
CCIDENT
M
ANAGEMENT
(
CLUSTER
SAM



SEE
T
ABLE
2)

The fission products constitute the principal health hazard to
the public, resulting from a severe
accident. Therefore, the amounts and physico
-
chemical forms of those materials released from
the reactor (the source term) are of great safety significance. As a result, the regulatory
authorities in some EU countries ar
e requiring to take into consideration as much as possible the
very unlikely severe beyond design
-
basis accidents (BDBAs). In the German licensing process,
for example, BDBA evaluations are necessary since 1 January 1994 to ensure that even
extremely unlik
ely events involving core melt
-
down would not require radical actions to ensure
protection against the damaging effects of ionising radiations outside the fence of the installation
site. BDBAs are also a concern expressed by the utilities and by the design
ers/vendors, as it is
shown in the discussions around the European Utility Requirements (EUR Document, last
release in April 2001) and in the MICHELANGELO initiative (started in December 1996).


As a result, it is envisaged by design to “practically elimi
nate” situations and phenomena which
could lead to early failure of the containment system and subsequent uncontrolled large releases
of fission products into the environment. Examples of such situations are high
-
pressure ejection
of molten core (possibly
leading to direct containment heating) and energetic in
-
vessel core
debris interactions with water (possibly leading to hydrogen generation). For example, for such
situations in the containment, a hydrogen strategy is proposed, based principally on passive

autocatalytic recombiners (PARs) aimed at keeping the hydrogen concentration far from the
critical deflagration
-
to
-
detonation transition (DDT) conditions. Other situations, then, such as
low pressure core melt, should be dealt with


or “controlled”


by
ensuring in the design that
the decay heat of the molten core can be removed and that the vessel or containment can
withstand the associated loads.


To better understand the source term behaviour and to develop appropriate prevention and
mitigation measure
s, appropriate research is needed which combines experimental investigations
and numerical modelling activities, supported by a robust scaling up strategy to extrapolate from
simulant to prototypical materials and from small
-
scale laboratory to full
-
scale
reactor
conditions. Historically, since the accidents of TMI
-
2 (March 1979) and Chernobyl
-
4 (April
1986), many international RTD programmes have been focusing on the development of a kind of
4
th

level to be added to the 3 “standard” levels of the defence
-
i
n
-
depth strategy, mentioned in the

11

Introduction. The international PHEBUS FP programme, in particular, that was launched by
IPSN (France) and is co
-
sponsored by the EC and other partners, is the largest and most
successful in
-
pile experimental programme de
voted to the source term behaviour: it is aimed at
bringing essential contributions to the knowledge on melt progression and fission product
release. Based on a series of integral in
-
pile experiments using real core materials, the PHEBUS
FP programme eval
uates the amount and nature of radioactive products that could be released
into the environment by occurrence of a core melt
-
down accident.


In the twenty
-
two projects belonging to the cluster SAM (“
2. Severe Accident Management
”),
the emphasis is on the

development of mitigative measures for defence
-
in
-
depth in the very
remote case of severe accidents (BDBAs). Under FP
-
4 a particular effort was devoted to the
understanding of BDBAs through the 45 projects of the 5 clusters INV (= IN
-
Vessel core
degradati
on), EXV(= EX
-
Vessel accident progression), ST (= radiological Source Term), CONT
(= accident progression in the CONTainment building) and AMM (= Accident Management
Measures). This effort on severe accident analysis is continued under FP
-
5 in the cluster

SAM.
Under FP
-
5, the following two key issues have been identified:




(1)
core degradation, corium formation in the reactor pressure vessel

and its behaviour inside
and outside the vessel (in particular upon a core
-
catcher). Research is needed with the aim

of
evaluating the coolability of the melt and ensuring the containment integrity. Criteria for
deflagration and detonation processes in hydrogen/air/steam/dust mixtures also are needed to
improve engineered safety systems and to better understand the capa
bilities of structures to
withstand dynamic loads. Finally, understanding the release of radioactive materials from a
degrading core into the cooling circuits and the containment, using in particular the
PHEBUS
-
FP results, will enable to optimise mitigati
on measures and to better predict the
source term. The above issues are the subject of section “
2.1 Assessment of Severe Accident
Risks
”.




(2)
improved methods and tools for severe accident management and operator training

that
make use of modern informat
ion and control systems and can handle uncertainties associated
with man
-
machine interfaces in a structured way. Research is needed to develop safety
systems for present and future reactors, which enable to extend the grace period, i.e.: the
period during

a severe accident when no active intervention is needed. The above issues are
the subject of section “
2.2 Severe Accident Management Measures
”.


No wonder that the important actors in the area “severe accident management” are the European
technical safet
y organisations, working for the national regulatory bodies, as it is also shown on
Table 5, that is principally: IPSN in France; GRS in Germany; universities and CSN in Spain;
NNC and HSE in the UK, SCK
-
CEN and AVN in Belgium, universities and SKI in Swed
en and
VTT in Finland.


Furthermore, the risks associated to those phenomena can be reduced through appropriate SAM
measures that could be implemented through the improvement of new plant specific designs
(e.g. ex
-
vessel core catchers) and for existing pla
nts the development of both, engineered
systems and backfitting measures (e.g. techniques for removing the hydrogen risk in the
containment or mitigation processes against radiological releases). Development of specific
operating emergency procedures is an
other expected result for current and future NPPs.


Generally, the results from experimental investigations and analytical studies on severe accident
(SA) phenomena contribute to improve the phenomena understanding (e.g. corium behaviour,
hydrogen explosi
ons or radiological releases) and to validate SA models and integral codes,
which have an impact on the quality of safety assessments, reduce uncertainties in the
quantification of safety margins and maintain readiness to respond to emerging issues.


12


Under

FP
-
5, the total EU budget to be spent for the 22 projects in the cluster SAM amounts
approximately to EUR 14.5 million, which represents roughly half of the total value of these
projects. This might be compared to the total EU budget spent under FP
-
4 for
the 45 projects in
the clusters INV, EXV, ST, CONT and AMM, which was EUR 29 million.


TABLE 2


2.1

Assessment of Severe Accident Risks

Core degradation processes in the presence of high burn
-
up and MOX fuel are investigated in
COLOSS

with emphasis on co
re degradation knowledge of PWR, BWR and VVER for
implementation in the various severe accident codes, including in particular tests on high burn
-
up and MOX dissolution, clad rupture, oxidation of U
-
O
-
Zr mixtures, and bundle experiments
concerning B
4
C cont
rol rod degradation and oxidation. Comprehensive phase diagrams of the
elements and systems present in both, in
-

and ex
-
vessel corium, are developed by
ENTHALPY,

based on one unique thermodynamic database and its coupling with severe accident codes.


The e
xperimental demonstration of the technical feasibility of ex
-
vessel mitigation measures
(core
-
catcher approach) and the validation of spreading codes are addressed in
ECOSTAR
.
Those experimental activities are addressing issues related to melt dispersion,

jet erosion, large
-
scale spreading, corium solidification, interaction with structural materials and coolability
approaches as top and bottom flooding.


Damage models and failure strain criteria of essential reactor components are examined in
LISSAC,

with

emphasis on the ultimate deformation capacity by applying uniaxial and biaxial
static and dynamic loads. The creep behaviour of prototypic reactor vessels, timing and modes
of its failure with and without penetrations, as well as the effects of the melt p
ool stratification
are investigated in
ARVI
.


The hydrogen combustion behaviour and the corresponding loads in complex multi
-
compartment geometries are investigated in
HYCOM
, using the large experimental programme
in the Russian RUT facility, with combusti
on modes ranging from slow to fast turbulent
deflagration.


A common code validation strategy for the integral code ASTEC is developed by
EVITA

with
the aim to optimise SAM strategies in a variety of NPPs: as a result, accident management
measures are opti
mised, such as filtered venting systems to limit the pressure in the containment
and/or PARs to reduce the hydrogen concentration.


The long
-
term behaviour of a solidified core immersed in a water pool is examined by
LPP,

which provides useful kinetics dat
a concerning the release of fission products and core materials
from in
-

and ex
-
vessel molten corium.


The PHEBUS
-
FP results are applied to SAM strategies in
PHEBEN
-
2
, which enables to
improve the safety margin calculation tools by developing detailed mode
ls for separate
-
effect
tests and integral codes for full
-
scale plant analysis. The radiological materials database
ASTERISM
-
2

continues to be developed with the aim to collect phenomenological data
relevant to advanced mitigation measures for the source te
rm.


EURSAFE
is an expert network which has as the main objective to establish within the EU an
ample consensus on SA issues where large uncertainties still subsist, and to propose a structured
approach to address them by appropriate R&D. The thematic net
work
THENPHEBISP

is

13

connected to an OECD/NEA/CSNI International Standard Problem exercise (ISP
-
46) devoted to
the PHEBUS FPT
-
1 test and is aimed at validating SA codes with emphasis on the identification
of numerical model uncertainties. In
SCACEX,

a Europ
ean network of experts is established in
order to identify the applicability and requirements of existing scaling up theories and methods
to reactor conditions, in particular, on thermal
-
hydraulic processes and materials behaviour.


Finally, there are two
actions on supporting transnational access to research infrastructures. The
action
PLINIUS
(homepage
www.cad.cea.fr
)

has as

main objective to provide support for
researchers to conduct specific experiments with prototypic corium in a platform composed by
t
he French facilities VULCANO, COLIMA, KROTOS and VITI, while the action
LACOMERA

is focused to provide support to researchers to conduct large
-
scale experiments
mainly on core degradation and melt retention aspects at the German facilities QUENCH, LIVE,
DI
SCO and COMET.


2.2

Severe Accident Management Measures

The scope is to contribute to the development of techniques, for example, to “practically
eliminate” some of those phenomena or to develop mitigation strategies to “control” some of
them. In addition
, the progress in numerical techniques as well as the availability of powerful
and cost
-
effective information technology systems will assure more reliable information and will
help to improve the diagnostic means as well as the implementation of some accid
ent
management measures.


Further assessment of various retention concepts, i.e. the pros and cons of internal and external
reactor vessel cooling, is the subject of the concerted action
EUROCORE

aimed at achieving
consensus on feasible and reliable indus
trial corium recovery concepts connected to in
-
vessel
retention, the corium
-
concrete interaction with water addition and ex
-
vessel spreading
techniques.


Experimental activities are developed by
SGTR
to generate experimental data and to validate
transport
models to support accident management interventions in steam generator tube rupture
sequences leading to severe accident conditions in PWRs and VVER
-
440. A better exploitation
of volatile iodine mitigation processes is investigated by
ICHEMM
, which provide
s useful
kinetics data for destruction and transmutation reactions of volatile forms of iodine.


Experimental activities are developed by
THINCAT

to develop a new hydrogen mitigation
concept based on catalytic surfaces as a coating on thermal insulation e
lements of the main
coolant loop components, as a complementary strategy to the installation of separate PARs. The
feasibility and reliability of PARs from an industrial prospect are examined in the concerted
action
PARSOAR,
which is comparing qualificatio
n tests and licensing procedures, relevant to
the hydrogen risk, under various severe accident conditions.


The radiological source term for operating reactors across Europe is examined in the project
OPTSAM
, with the aim to evaluate qualitatively the impa
ct of various SAM strategies on
radiological effects, and to establish a technical basis for the definition of realistic source terms.


Finally the thematic network
SAMOS

is investigating the feasibility of a computerised operator
supporting tool for eleme
nts in SA scenarios (e.g. severe accident management guidelines).







14

3

E
VOLUTIONARY
C
ONCEPTS
(
CLUSTER
EVOL

-

SEE TABLE
3)

It is worth recalling that there are currently 20 reactor units in construction in Eastern Europe
and 25 reactor units in constructi
on in the rest of the world, with total planned capacities of 16
and 21 net GWe, respectively. Some of these reactor units are of the evolutionary type, i.e. with
emphasis on design simplification and enhanced man
-
machine interface, with the aim to further

reduce any (severe) accident risk. In evolutionary LWR designs, with their emphasis on design
simplification and enhanced man
-
machine interface, the severe accident risk will be further
reduced. Some of the innovative reactors rely mainly on passive preve
ntion and mitigation
features and systems.


Part of EC co
-
sponsored research is devoted to the investigation of phenomena associated with
the use of passive systems in some evolutionary LWR designs for decay heat removal (from the
core region and from the
containment building) and for other safety measures (e.g.
depressurisation and injection). The potential advantages of passive safety systems (e.g.
independence from external energy sources, simpler design, less complex instrumentation and
control) should
be weighted against their potential disadvantages (e.g. reliance on small driving
forces, limited operational flexibility, reduced in
-
service testing capability, difficult diagnosis of
status, etc). In addition, the coupling of different neutronics and the
rmal
-
hydraulics computer
codes, needed for this purpose, enables to improve and/or validate the numerical models used in
the existing thermal
-
hydraulics computer codes, and to extrapolate the results of the small
-
scale
experiments towards full
-
scale reacto
r conditions.


Another area of interest for EC co
-
sponsored research in evolutionary reactors is the use of
MOX fuel which has been used on industrial level since 1982 in a number of EU power plants,
for example: up to 50 % core loading in 9 German reactor
s, up to 30 % in 17 French reactors and
up to 25 % in 2 Belgian reactors. High burn
-
up fuel is another matter of increasing interest


for
countries like France who have achieved burn
-
up targets of nearly 50 GWd/t for standard fuel
and 40 GWd/t for MOX fue
l on an industrial scale.



In the eightteen projects belonging to the cluster “EVOL (“
3. Evolutionary Concepts
”) the
emphasis is on a new generation of reactors which should be cheaper, safer and more simple to
operate, using, for example, passive (self
-
a
cting) safety systems. Under FP
-
5, the following two
key issues have been identified:


(1) cost and safety advantages of evolutionary improvements

in currently used nuclear
technologies, in particular those with the potential to significantly reduce the ri
sk and
consequences of human error and public concerns about nuclear technology (e.g. passive safety
systems). The above issues are the subject of section “
3.1 Evolutionary Safety Concepts
”.


(2) understanding of the performance of
high burn
-
up and MOX fu
el

under transient and
accident conditions as a basis for lowering fuel costs, whilst maintaining or improving safety
margins. The above issues are the subject of section “
3.2 High Burn
-
up and MOX Fuel
”.


No wonder that the important actors in the area “e
volutionary concepts” are the European
vendors and manufacturers, as it is also shown on Table 6.


As far as passive safety systems in evolutionary LWRs are concerned, the emphasis is on
understanding decay heat removal processes (both from the core region

and from the
containment building) and developing safety measures (e.g. depressurisation and injection). The
use of different neutronics and thermal
-
hydraulics computer codes (ATHLET, CATHARE,
TRAC, RELAP5, etc.) for pre
-

and post
-
test calculations is an
important part of the work
programme.


15


Research is also conducted to improve some

technologies (of interest for both present and next
generation reactors) that have potential cost and safety advantages.

Under the economic pressure
of the manufacturing ind
ustry in a deregulated electricity market, many efforts are naturally
devoted to the improvement of evolutionary safety concepts adapted to the new operational
conditions (e.g. related to reactor backfitting, power upgrades, high burn
-
up, MOX fuel, etc).


Under FP
-
5, the total EU budget to be spent for the 18 projects in the cluster EVOL amounts to
approximately EUR 10.5 million, which represents roughly half of the total value of these
projects. This might be compared to the total EU budget spent under FP
-
4 for the 11 projects in
the INNO cluster, which was EUR 4.8 million.


TABLE 3


3.1

Evolutionary Safety Concepts

The shared cost projects in this area have all a strong “numerical” flavor as progress in
numerical modelling, especially CFD (computational fl
uid dynamics) codes, is needed to catch
up with the progress made recently in experimental investigations, especially for passive
systems. The further use, improvement and applications of CFD codes for a wide range of
conditions were highly recommended in
the conclusions of some final reports of FP
-
4.


Four projects are strongly related to the use, development and improvement of three
-
dimensional
(3
-
D) and computational fluid dynamics (CFD) codes. The project
ASTAR

aims at the
development of advanced numeri
cal methods for 3
-
D two
-
phase flow simulation tools that might
lay the scientific and technical basis for a new generation of thermo
-
hydraulics codes. This
should improve the modelling of safety relevant phenomena related to the next generation of
evolutio
nary LWRs. The primary objective of the
TEMPEST

project is to validate (against
existing and new experimental data) and to improve advanced modelling methods for evaluating
pressure safety margins of the containment of BWRs (homepage
:http://www.nrg
-
nl.com/
extranet/tempest/index.html)
. Accurate prediction of containment pressure transients
during severe accidents requires capabilities for modelling effects such as 3D mixing and
stratification, since these strongly affect the performance of passive cooling sy
stems. For that
reason, the applicability of CFD will be particularly investigated. The expected outcome of the
ECORA

project (
http://domino.grs.de/ecora/user_forum_ecora.nsf)

is a comprehensive
evaluation of CFD software for applications in the primary sy
stem and the containment of
nuclear reactors, resulting in recommendations for Best Practice Guidelines and for necessary
CFD software improvements. The project aims also at establishing a Network of European
Centres of competence for applications of CFD c
odes to reactor safety. In the concerted action
EUROFASTNET

a critical assessment is made of the needs, in nuclear engineering, for
thermal
-
hydraulics R&D, with emphasis on a coherent balance between advanced CFD
modelling and experimental validation using

innovative instrumentation.


Three cost
-
shared actions are addressing operational practices and design improvement of
LWRs. The goal of the
NACUSP

project is to enhance the basic understanding on BWRs
thermal
-
hydraulics stability issues, under both forced

and natural
-
circulation conditions, through
generation of new experimental data, elaboration of guidelines, development of efficient models
and validation of computer codes. The results of this project should improve operational
flexibility and increase t
he confidence level on the safety margins of the operating BWRs and
future designs. The
DEEPSSI
project proposes to develop and test an innovative high
-
pressure
steam injector design and to assess its potential application in an Emergency Feedwater System
(EFWS) of PWRs steam generators. Applications to both western and eastern (i.e. VVER
-
440)
types PWRs are considered. The purpose of the
FABIS

project to develop and test a diverse fast
-

16

acting boron injection system that can be used in case of an anticipat
ed transient without scram
in the existing and the future advanced BWRs. The main innovation of FABIS compared to
similar existing designs is the use of steam for the pressurisation of the boron solution tank
instead of nitrogen. In this project, the SWR 1
000 concept from Framatome ANP has been
selected as the reference reactor


Four projects address the improvement of other analytical tools. Two of them address the issue
of coupling neutronics and thermal
-
hydraulics codes. The main objective of
CRISSUE
-
S

i
s to
elaborate a state of the art report about the coupling of neutronics and thermal
-
hydraulics codes
for LWRs with emphasis on the so
-
called RIA (Reactivity Initiated Accidents


homepage
www.ing.unipi.it/crissue_s).
The project should provide recommendat
ions to utilities and
regulators about possible safety margins of existing reactors and optimized accident
management. The
VALCO

project is specifically aimed at the validation (against experimental
data) of coupled neutronics/thermal
-
hydraulics codes for
VVERs (homepage http://www.fz
-
rossendorf.de/FWS/VALCO/). The work is based on results from an EU Phare project
(SRR1/95) which analysed transients initiated by perturbations in the VVERs secondary circuit.


The objective of the
RMPS
project is to propose
a specific methodology to assess the thermal
-
hydraulic reliability of passive systems. The main activities are the identification and
quantification of the sources of uncertainties, the propagation of the uncertainties through the
thermal
-
hydraulics models

and the introduction of passive systems unreliability in the accident
sequence analysis.


The objective of the
ITEM
network, co
-
ordinated by EdF, is to ensure a rapid and co
-
ordinated
implementation and dissemination of the developments of tools for comp
uter simulation of
radiation effects in materials (homepage

:
http://item.edf.fr
). These tools known as “Virtual Test
Reactors” (VTR) are becoming increasingly important in view of the expected insufficient
availability of

test reactors and hot cell facilities in the future.


3.2

High Burn
-
up and MOX Fuel

The project
MICROMOX

includes experimental tests aimed at understanding to what extent
the as
-
fabricated microstructure of the MOX fuel influences the gas release in norma
l and
abnormal conditions at high burn
-
up. Different numerical codes simulating the thermo
-
mechanical behaviour of the fuel are benchmarked.

This project is complemented by
OMICO
,
which has as a main objective to study and model the influence of microstruc
ture and matrix
composition on oxide fuel behaviour under normal LWR operating conditions.


The main objective of the
VALMOX

project is the validation of MOX fuel calculations at high
burn
-
up based on available experimental data from LWRs using the JEF nu
clear data base as
well as state
-
of
-
the
-
art neutronics codes. First indications on safety limits of the MOX fuels in
terms of helium production are also expected.


Two projects are addressing specific issues of the cladding materials of LWRs fuel assemblie
s.
One of them,
SIRENA
, has as a main objective the development of predictive tools to assess the
first barrier integrity of LWR fuel pin cladding manufactured with Zr
-
Nb alloys. The other one,
EXTRA
, consists mainly in an extension of the applicability of

the TRANSURANUS fuel code
(developed at JRC/ITU Karlsruhe) through implementation of newly developed models to
simulate high temperature behaviour of Nb contained in cladding materials.






17


CONCLUSION


Multidisciplinary and comprehensive research related

to plant modernisation (PLEM cluster),
severe accident management (SAM cluster) and innovative safety concepts (EVOL cluster) is in
progress in the current 73 multi
-
partner projects of FP
-
5 devoted to “Operational safety of
Existing Installations”. The st
ructure of the 3 clusters is described in Tables 1, 2 and 3. Co
-
operation amongst a variety of public and private organisations is enforced, thus allowing the
optimal utilisation of the available resources and the enhancement of the nuclear research fabric

within the Community.


The results of 41 out of the 73 projects were presented at FISA
-
2001 (EC Luxembourg,
November 12
-
14, 2001). A total of approximately 300 experts from 22 countries were
participating. The key players in the nuclear arena were presen
t, i.e. the regulatory authorities
(and/or their technical safety organisations), manufacturing industry, electric power utilities,
universities, research organisations and governemental organisations concerned (see also Tables
4, 5 and 6).


Research on pl
ant modernisation is contributing to improvements and a better harmonisation of
safety practices throughout the EU Member States and the candidate CEECs. It has also been
shown that severe accident research results in a better evaluation of these very unli
kely events:
the uncertainties are better quantified and subsequently the risks related to corium behaviour,
hydrogen explosions and/or radiological releases are reduced. Some interesting evolutionary
safety concepts and innovative numerical tools have bee
n discussed. Further exploitation of the
results of this innovative research will help the nuclear industry achieve the scope of reducing
even further the risks of both design basis and severe accidents.


Besides technological requirements, socio
-
economic

aspects are becoming increasingly
important due to the level of public and political acceptance and to the economic pressure of
deregulated electricity markets. It has been shown that research conducted in the Euratom
framework could contribute to meet th
ese requirements, thereby maintaining nuclear power as a
competitive and sustainable option for the energy policy of the European Union. More
information on the framework programme activities can be found on the Community research
homepage
http://www.cordis.lu/fp5
-
euratom/home.html

as well as on the homepage of the Unit
«

Nuclear Fission and Radiation Protection

»

http://europa.eu.
int/comm/research/energy/fi/fi_en.html



indirect actions

») and on the above mentioned JRC homepage («

direct

actions

»)
.


In the future, Community research in nuclear fission will continue to contribute to offer a
response to some of the main concerns
raised by the key stakeholders and by the European
citizens at large in line with the European Research Area concept (ERA). The new challenge to
Euratom research will be to reorganise itself using the implementation instruments proposed for
the next 6
th

fr
amework programme (2002
-
2006), in particular, networks of excellence and
integrated projects, bringing together private and public resources with the aim to achieve real
integration of European research.


"Helping exploit the full potential of nuclear ene
rgy in a sustainable manner", which is the main
aim of the current Euratom FP
-
5, is quite a challenge for now and will remain so in the future.
To meet this challenge in the future FP
-
6, the main actors sharing this aim are invited to propose
networks of e
xcellence and integrated projects. The aim of these new instruments is to solve
problems of common interest along the lines of well defined integration strategies, involving the
key actors. Acting in this way will contribute to make a reality of the Europe
an Area for nuclear
fission Research (ERA). F
urther information on the objectives and individual instruments for
implementing the FP6 thematic areas is available from the website of the European
Commisssion Directorate General “Research” http://europa.eu.i
nt/comm/research/fp6/networks
-
ip.html.


18



19

Plant Life Extension and Management Research under FP
-
5 (PLEM)




































Cluster PLEM

Integrity of equipment
and structures

On
-
line monitoring
and maintenance

Organisation and
management of safety

RETROSPEC


FRAME
PISA


FEUNMARR
-

RENION

CASTOC
-
PRIS
-
INTERWELD
FLOMIXR
-

WAHALOADS

LIRES

GRETE
-

SPIQNAR

REDOS

BESECBS
-

CEMSIS

NURBIM

VERSAFE
-

VERLIFE

IMPAMVVER
-
EUROSAFE

ADIMEW
-

VOCALIST

SMILE
-

THERFAT


* JSRI *


Corrosion
Thermalhydraulics

Safety margins

Welds


Risk assessment

Virtual reality

Digital Instrumentation


EU/CEEC
-
Harmonisation
of practices
-

VVER

safety

Embrittlement

Materials Research Reactors

ATHENA

Knowledge management

LEARNSAFE

ENPOWER


SPI
-

VRI
MOR

MAECENAS
-

CONMOD


Concrete ageing


Table 1 : Work programme and main issues of PLEM


21



Severe Accident Management Research under FP
-
5 (SAM cluster)














PHEBUS
-
FP Programme

Cluster SAM

SAM Measures


Assessment

of SA Risk

ICHEMM

THINCAT

PARSOAR


OPTSAM

EVITA


PHEBEN 2


EUROCORE

ARVI

COLOSS




ENTHALPY

PLINIUS
-

LACOMERA

LPP
-

THENPHEBISP

ASTERISM II

HYCOM

LISSAC

Reactor Pressure
Vessel





Code Development





By
-
Pass
Sequences




Hydrogen /
Containment







Source Term





Corium




Table 2


Work programme and main issues of SAM

SAMOS

E
URSAFE

SGTR

ECOSTAR

SCACEX


22


Evolutionary Concepts Research under FP
-
5 (EVOL cluster)



























Table 3: Work programme and main issues of EVOL


Evolutionary Safety
Concepts

Hi
gh Burn
-
up

and Mox


Analytical Tools

(codes, methodologies)

Operational Practices and
Design Improvement

Databases and

Education and Training


ASTAR
-

ECORA
-

RMPS

VALCO


CRISSUES
TEMPEST
-

ITEM

CERTA

EUROFASTNET

NACUSP

DEEPSSI

FABIS


MICROMOX

OMICO

Cluster EVOL


ENEN



EXTRA
-

SIRENA


VALMOX


23


Table 4: PLEM Cluster / List of projects, co
-
ordinators and dates




EMBRITTLEMENT
-

MATERIALS RESEARCH R
EACTORS

Acronym


Project Title


Co
-
ordi
nator
(Organisation)


Start
(duration )

PISA

(FIKS
-
CT2000
-
00080)


Phosphorus influence on steel aging

AEA Technology

01/12/00
36 months

GRETE

(FIKS
-
CT2000
-
00086)

Evaluation of non destructive testing techniques for
monitoring of material degradation

E
DF

01/10/00
36 months

RETROSPEC

(FKS
-
CT2000
-
00091)


Retrospective dosimetry focussed on the reaction
93nb(n,n')

NRG

01/10/00
18 months

FEUNMARR

(FIR1
-
CT2001
-
20122)


Future EU needs in materials research reactors

CEA/DTAP/SPI

01/11/01
12 months

FRAM
E

(FIKS
-
CT2000
-
00101)

Fracture mechanics based embrittlement trend
curves for the characterisation of nuclear pressure
vessel materials


VTT

01/09/00
36 months

REDOS

(FIKS
-
CT2001
-
00120)

Reactor Dosimetry: Accurate Determination and
Benchmarking of Radia
tion Field Parameters,
Relevant for RPV Monitoring

Tecnatom S.A.

01/11/01
36 months

RENION

(FIR1
-
CT2002
-
40157)


Reactor neutronic investigation on LR
-


reactor

NRI
-
UJV

01/09/02
24 months

SPIQNAR

(FIKS
-
CT2000
-
00065)

Signal processing and improved qua
lification for
non
-
destructive testing of aging reactors

Mitsui Babcock

01/10/03
36 months

CRACHS

(FIKS
-
CT2001
-
06004)

Consensus on Reconstitution Techniques and
Fracture Toughness Analysis of Charpy
-

Type
specimens (workshop)

SCK
-
CEN

5
-
7/0
9/01

MASC

(FIKS
-
CT2001
-
80126)

Use and application of the master curve method for
determining fracture toughness (workshop)


VTT

12
-
14/06/02

MARIE CURIE
FELLOWSHIP

(MCFI
-
2001
-
01088)


Positron annihilation study of reactor pressure
vessel surveillance sp
ecimens

Framatome ANP

01/02/02
12 months





CORROSION
-

THERMALHYDRAULICS


LIRES

(FIKS
-
CT2000
-
00012)


Development of light water reactor reference
electrodes

SCK
-
CEN

01/10/00
48 months

CASTOC

(FIKS
-
CT2000
-
00048)

Crack growth
behaviour of low alloy steel for
pressure boundary components under transient
light water reactor operating conditions

MPA

01/10/00
36 months

PRIS

(FIKS
-
CT2000
-
00084)

Properties of irradiated stainless steels for
predicting lifetime of nucl
ear power plant
components

ABB Atom Ab

01/10/00
36 months

INTERWELD

(FIKS
-
CT2000
-
00103)

Irradiation effects on the evolution of the
microstructure properties and residual stresses in
the heat affected zone of stainless steel welds

NRG

01/09/0
0
42 months

WAHALOADS

(FIKS
-
CT2000
-
00106)

Two
-
phase flow water hammer transients and
induced loads on materials and structures of
nuclear power plants

Université
Catholique de
Louvain

01/10/00
36 months


24

FLOMIX
-
R

(FIKS
-
CT2001
-
20197)

Fluid mixing a
nd flow distribution in the reactor
circuit

FZR

01/01/02
24 months

MARIE CURIE
FELLOWSHIP

(MCFI
-
2001
-
01139)


The influence of the corrosion potential on the
stress corrosion crack propagation of stainless steel
in PWR primary coolant

SCK
-
CEN

01/02/02

12 months



25


SAFETY MARGINS
-

WELDS

Acronym


Project Title


Co
-
ordinator
(Organisation)


Start
(duration )

ADIMEW

(FIKS
-
CT2000
-
00047)


Assessment of aged piping dissimilar metal weld
integrity

EDF

01/11/00
36 months

VOCALIS
T

(FIKS
-
CT2000
-
00090)


Validation of constraint
-
based assessment
methodology in structural integrity

SERCO Assurance

01/10/00
36 months

IPC

(FIKR
-
CT2000
-
80119)

integrity of pressurised components of nuclear
power plants (workshop)

GRS

1
7
-
21/09/01

SMILE

(FIKS
-
CT2001
-
00131)


Structural margin improvements In aged
-
embrittled
RPV with load history effects

EDF

01/11/00
36 months

ENPOWER

(FIKS
-
CT2001
-
00167)


Management of nuclear plant operation by
optimising weld repairs

Institut de Soudur
e

01/12/01
36 months

THERFAT

(FIKS
-
CT2001
-
00158)


Thermal fatigue evaluation of piping system "tee"
-

connections

E.ON Kernkraft
GmbH

01/12/01
36 months


RISK ASSESSMENT
-

VIRTUAL REALITY

NURBIM

(FIKS
-
CT2001
-
00172)

Nuclear risk
-
based inspection method
ology for
passive components

GRS

01/11/01
32 months

SPI

(FIKS
-
CT2001
-
20145)

Evaluation of alternative approaches for
assessment of safety performance indicators

for nuclear power plants

GRS

01/10/01
18 months

VRIMOR

(FIKS
-
CT2000
-
00114)


Virtual rea
lity for inspection, maintenance,
operation, and repair of nuclear power plant

NNC Limited

01/02/01
24 months


CONCRETE
AGEING

MAECENAS

(FIKS
-
CT2001
-
00186)


Modeling of aging in concrete nuclear power plant
structures

University of
Sheffiel
d

01/11/01
36 months

CONMOD

(FIKS
-
CT2001
-
00204)

Concrete containment management using the finite
element technique combined with in
-
situ non
-
destructive testing of conformity with respect to
design and construction quality

Force Institute

01/01/
02
36 months



26


DIGITAL INSTRUMENTAT
ION

Acronym


Project Title


Co
-
ordinator
(Organisation)


Start
(duration )

BE
-
SECBS

(FIKS
-
CT2000
-
00054)

Benchmark exercise on safety evaluation of
computer
-
based systems

EC
-
JRC

01/01/01
30 months

CEMSIS

(FIKS
-
CT2000
-
00109)

Cost effective modernisation of systems important
to safety

British Energy

01/01/01
36 months





EU/CEEC
-

HARMONISATION OF PRA
CTICES
-

VVER SAFETY

ATHENA

(FIR1
-
CT2001
-
20170)


AMES thematic network on aging

Tractebel S.A.

01/11/01
36 m
onths

VERSAFE

(FIKS
-
CT2000
-
20044)

Concerted utility review of VVer
-
440 safety research
needs

Fortum

01/09/00
24 months

VERLIFE

(FIKS
-
CT2001
-
20198)


Unified procedure for lifetime assessment of
components and piping in VVER NPPS

NRI
-

UJV

01/10/01
24 m
onths

IMPAM
-
VVER

(FIKS
-
CT
-
2001
-
00196)

Improved accident management of VVER nuclear
power plants

VTT

01/11/01
32 months

EUROSAFE

International approach towards convergence of
technical nuclear and radiological practices in
Europe

IRSN

01/11/02 (?)

12 mon
ths

PSARID

(FIKS
-
CT2000
-
80118)

Probabilistic safety assessment and risk
-
informed
decision making (workshop)

GRS

5
-
9/03/01

SMIRT
-
17

(FIKS
-
CT2001
-
06005)

17th International conference on structural
mechanics in reactor technology (workshop)

Brno University

of
Technology

18
-
22/08/
2003





KNOWLEDGE MANAGEMENT

JSRI

(FIR1
-
CT2000
-
20089)

Joint safety research index

GRS

01/01/01
30 months

LEARNSAFE

(FIKS
-
CT2001
-
00162)

Learning organisations for nuclear safety

VTT

01/11/01
30 months


27


Table 5 : SAM Cl
uster: list of projects, co
-
ordinators and dates




CORIUM

Acronym


Project Title


Co
-
ordinator
(Organisation)


Start
(duration )

ENTHALPY

(FIKS
-
CT1999
-
00001)


European nuclear thermodynamic database (for in
-

and ex
-

ves
sel applications)

IRSN

01/02/00
36 months

COLOSS


(FIKS
-
CT1999
-
00002)

Core loss during a severe accident

IRSN

01/02/00
36 months

ECOSTAR

(FIKS
-
CT1999
-
00003)

Ex
-
vessel core melt stabilisation research

FZK

01/02/00
48 months

EUROCORE

(FIKS
-
CT1999
-
20003
)


European group for analysis of corium recovery
concepts

CEA/DRN/DTP

01/03/00
24 months

PLINIUS

(FIR1
-
CT2001
-
40152)


Platform for improvements in nuclear industry and
utility safety

CEA/DTP

01/12/01
36 months

LACOMERA

(FIR1
-
CT2002
-
40158)


Large scale E
xperiments on core degradation, melt
retention and coolability

FZR

01/09/02
36 months

CORIUM COURSE

(FIKS
-
CT2001
-
00139)


European special training course on corium
(workshop)

CEA/DTP

27
-
31/01/
2003


REACTOR PRESSURE VESSEL

ARVI

(FIKS
-
CT1999
-
00011)

As
sessment of reactor vessel integrity

Kungl Tekniska
Högskolan/RIT

01/02/00
36 months

LISSAC

(FIKS
-
CT1999
-
00012)

Limit strains for severe accident conditions

FZK

01/02/00
36 months





SOURCE TERM

LPP

(FIKS
-
CT1999
-
00005)


Late p
hase source term phenomena

AEA Technology

01/02/00
30 months

ICHEMM

(FIKS
-
CT1999
-
00008)


Iodine chemistry and mitigation methods

AEA Technology

01/02/00
36 months

ASTERISM II

(FIKS
-
CT1999
-
20001)


Archive models for source term informat
ion and
system models

AEA Technology

1/02/00
18 months

THENPHEBISP

(FIKS
-
CT2001
-
20151)


Thematic network for a Phebus FPT
-
1 international
standard problem

IRSN


01/12/01
24 months







28


HYDROGEN / CONTAINMENT

Acronym


Project Title


Co
-
ordinator
(Organisation)


Start
(duration )

HYCOM

(FIKS
-
CT1999
-
00004)


Integral large scale experiments on hydrogen
combustion for severe accident code validation

FZK

01/02/00
36 months

THINCAT

(FIKS
-
CT1999
-
00006)


Hydrogen removal from LWR

containments by
catalytic coated thermal insulation elements

FZJ

01/01/00
29.5 months

PARSOAR

(FIKS
-
CT1999
-
20002)


Hydrogen hazard
-

passive autocatalytic
recombiners state
-
of
-
the
-
art

Technicatome

01/02/00
18 months

SCACEX

(FIR1
-
CT2001
-
20127)

Sc
aling of containment experiments

Becker Technologies

01/01/02
12 months


BY
-
PASS SEQUENCES

SGTR

(FIKS
-

CT1999
-
00007)


Steam generator tube rupture scenarios

VTT

01/01/00
36 months

OPTSAM

(FIKS
-
CT1999
-
00013)

Optimisation of severe accident ma
nagement
strategies for the control of radiological releases

NNC Ltd

01/06/00
24 months


CODE DEVELOPMENT

PHEBEN 2

(FIKS
-
CT1999
-
00009)


Validating severe accident codes against PHEBUS
FP for plant applications

JRC

01/03/00
48 months

EVITA

(FIKS
-
CT1
999
-
00010)


European validation of the integral code ASTEC

VTT

01/02/00
36 months

SAMOS

(FIKS
-
CT2001
-
20189)


A perspective on computerized severe accident
management operator support

SCK
-
CEN

01/12/01
18 months

MARIE CURIE
FELLOWSHIP

(FIKS
-
CT
1999
-
50001)


Advanced 2D two phase flow simulation tool for
application to reactor safety

Von Karman Institute
for Fluid Dynamics

01/03/00
12 months






NETWORKING




EURSAFE

(FIKS
-
CT2001
-
20147)


European expert network for the reduction of
uncert
ainties in severe accident safety issues

IRSN

01/12/01
24 months

PHEBUS
-
FP

3428
-
88
-
07
-
TP ISP F

PHEBUS Fission Product international programme

IRSN

07/1988


29


Table 6: Evol Cluster: list of projects, co
-
ordinators and dates




ANALYTICAL TOOLS (codes,

methodologies)

Acronym


Project Title


Co
-
ordinator
(Organisation)


Start
duration

ASTAR

(FIKS
-
CT2000
-
00050)

Advanced three
-
dimensional two
-
phase flow
simulation tool for application to reactor safety

CEA
-

DMT

01/09/00
36 months

RMPS

(FIKS
-
CT2000
-
00073)

Reliability methods for passive safety functions

CEA/DRN/DER

01/02/01
36 months

TEMPEST

(FIKS
-
CT2000
-
00095)

Testing and enhanced modeling of passive
evolutionary systems technology (for containment
cooling)

NRG

01/12/00
36 months

ECORA

(FIKS
-
C
T2001
-
00154)

Evaluation of computational fluid dynamic methods
for reactor safety analyses

GRS

01/10/01
36 months

VALCO

(FIKS
-
CT2001
-
00166)

Validation of coupled neutronics/thermal
-
hydraulics
codes for VVER reactors

FZR

01/01/02
24 months

CRISSUE
-
S

(
FIKS
-
CT2001
-
00185)


Revisiting critical issues in nuclear reactor design
safety by using 3
-
D neutronics / thermalhydraulics
models: state
-
of
-
the
-
art

University of Pisa

01/01/02
24 months

ITEM

(FIR1
-
CT2001
-
20136)

Improvement of techniques for multiscale
modeling of
irradiated materials

EDF

01/11/01
48 months


OPERATIONAL PRACTICES AND DESIGN IMPROVEMENT

NACUSP

(FIKS
-
CT2000
-
00041)

Natural circulation and stability performance of BWRs


NRG

01/12/00
48 months

DEEPSSI

(FIKS
-
CT2000
-
00113)

Desi
gn and development of a steam generator
emergency feedwater passive system for existing
and future PWR's using advanced steam Injectors


CEA/DRN/DER

01/12/00
36 months

FABIS

(FIKS
-
CT2001
-
00195)

Fast
-
acting boron injection system

VTT

01/09/01
24 months







30


DATABASES AND EDUCATION AND TRAINING

Acronym


Project Title


Co
-
ordinator
(Organisation)


Start
duration

CERTA

(FIR1
-
CT2000
-
20052)

European network for the consolidation of the integral
system experimental data bases for reactor
thermalhydrau
lic safety analysis

DIMNP

01/10/00
24 months

EUROFASTNET

(FIKS
-
CT2000
-
20100)

European group for future advances in sciences and
technology for nuclear engineering
thermalhydraulics

CEA

01/09/00
18 months

ENEN

(FIKI
-
CT2001
-
80127)

European nuclear eng
ineering network

SCK
-
CEN

01/01/02
24 months


NUCLEAR FUEL (High Burn
-
up and Mox)

OMICO

(FIKS
-
CT2001
-
00141)

Oxide fuels: microstructure and composition
variations


SCK
-
CEN

01/10/01
36 months

MICROMOX

(FIKS
-
CT2000
-
00030)

The influence of microstruct
ure of MOX fuel on its
irradiation behaviour under transient conditions


Belgonucléaire
S.A.

01/10/00
48 months

SIRENA

(FIKS
-
CT2001
-
00137)

Simulation of radiation effects in Zr
-
Nb alloys:
application to stress corrosion cracking behaviour in
iodine
-
ric
h environment


EDF

01/01/02
36 months

EXTRA

(FIKS
-
CT2001
-
00173)

Extension of TRANSURANUS code applicability with
Nb containing cladding models


KFKI / AEKI

01/12/01
24 months

VALMOX

(FIKS
-
CT2001
-
00191)

Validation of high
-
burnup MOX fuel calculations

Belgonucléaire
S.A.

01/10/01
30 months