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					                    NUCLEAR FISSION AND RADIATION PROTECTION
                     PROJECTS SELECTED FOR FUNDING 1999-2002
                                            ANNEX I
                OPERATIONAL SAFETY OF EXISTING INSTALLATIONS -
                     SUMMARIES FOR SELECTED PROJECTS
                                      Table of contents

PLANT LIFE EXTENSION AND MANAGEMENT
General
JSRI - Joint safety research index                                                           4
Integrity of equipment and structures
FRAME - Fracture mechanics based embrittlement trend curves for the characterisation
of nuclear pressure vessel materials                                                         6
RETROSPEC - Retrospective dosimetry focussed on the reaction 93nb(n,n')                      8
PISA - Phosphorus influence on steel ageing                                                 10
RENION - Reactor neutronic investigations on LR-0 reactor                                   12
INTERWELD - Irradiation effects on the evolution of the microstructure, properties and
residual stresses in the heat affected zone of stainless steel welds                        14
PRIS - Properties of irradiated stainless steels for predicting lifetime of nuclear power
plant components                                                                            16
CASTOC - Crack growth behaviour of low alloy steel for pressure boundary components
under transient light water reactor (LWR) operating conditions                              18
ADIMEW - Assessment of aged piping dissimilar metal weld integrity                          20
VOCALIST - Validation of constraint-based assessment methodology in structural
integrity                                                                                   22
SMILE - Structural margin improvements in aged-embrittled RPV with load history
effects                                                                                     25
THERFAT - Thermal fatigue evaluation of piping system "Tee"- connections                    28
WAHALOADS - Two-phase flow water hammer transients and induced loads on
materials and structures of nuclear power plants                                            30
FLOMIX-R - Fluid mixing and flow distribution in the reactor circuit                        33
FEUNMARR - Future EU needs in materials research reactors                                   35
MAECENAS - Modelling of ageing in concrete nuclear power plant structures                   37
CONMOD - Concrete containment management using the finite element technique
combined with in-situ non-destructive testing of conformity with respect to design and
construction quality                                                                        39
On-line monitoring and maintenance
GRETE - Evaluation of non destructive testing techniques for monitoring of material
degradation                                                                                 41
LIRES - Development of Light Water Reactor (LWR) reference electrodes                       43
SPIQNAR - Signal processing and improved qualification for non-destructive testing of
ageing reactors                                                                             45



                                                1
REDOS - Reactor Dosimetry: accurate determination and benchmarking of radiation
field parameters, relevant for reactor pressure vessel monitoring                         47
VRIMOR - Virtual reality for inspection, maintenance, operation, and repair of nuclear
power plant                                                                               49
NURBIM - Nuclear risk-based inspection methodology                                        51
ENPOWER - Management of nuclear plant operation by optimising weld repairs                53
Organisation and management of safety
BE-SECBS - Benchmark exercise on safety evaluation of computer-based systems              55
CEMSIS - Cost effective modernisation of systems important to safety                      57
LearnSafe - Learning organisations for nuclear safety                                     59
SPI - Evaluation of alternative approaches for assessment of safety performance
indicators for nuclear power plants                                                       61
Safety of VVER reactors
IMPAM-VVER - Improved accident management of VVER nuclear power plants                    63
VERLIFE - Unified procedure for lifetime assessment of components and piping in VVER
NPPS                                                                                      65
ATHENA - AMES thematic network on ageing                                                  67

SEVERE ACCIDENT MANAGEMENT
Assessment of severe accident risks
COLOSS - Core loss during a severe accident                                               69
LISSAC - Limit strains for severe accident conditions                                     71
ARVI - Assessment of reactor vessel integrity                                             73
ENTHALPY - European nuclear thermodynamic database (for in- and ex-vessel
applications)                                                                             75
ECOSTAR - Ex-vessel core melt stabilisation research                                      77
HYCOM - Integral large scale experiments on hydrogen combustion for severe accident
code validation                                                                           79
EVITA - European validation of the integral code ASTEC                                    81
LPP - Late phase source term phenomena                                                    83
PHEBEN 2 - Validation of severe accident codes against Phebus FP for plant applications   85
ASTERISM II - Archive models for source term information and system models                87
EURSAFE - European expert network for the reduction of uncertainties in severe
accident safety issues                                                                    89
THENPHEBISP - Thematic network for a Phebus FPT-1 international standard problem          91
SCACEX - Scaling of containment experiments                                               93
PLINIUS - Platform for improvements in nuclear industry and utility safety                95
LACOMERA - Large scale experiments on core degradation, melt retention and
coolability                                                                               97




                                                2
Severe accident management measures
EUROCORE - European group for analysis of corium recovery concepts                           99
SGTR - Steam generator tube rupture scenarios                                               101
ICHEMM - Iodine chemistry and mitigation methods                                            103
THINCAT - Hydrogen removal from LWR containments by catalytic coated thermal
insulation elements                                                                         105
PARSOAR - Hydrogen hazard - passive autocatalytic recombiners state-of-the-art              107
OPTSAM - Optimisation of severe accident management strategies for the control of
radiological releases                                                                       109
SAMOS - A perspective on computerized severe accident management operator support           111
VERSAFE - Concerted utility review of VVER-440 safety research needs                        113

EVOLUTIONARY CONCEPTS
Evolutionary safety concepts
ASTAR - Advanced three-dimensional two-phase flow simulation tool for application to
reactor safety                                                                              115
TEMPEST - Testing and enhanced modelling of passive evolutionary systems technology
(for containment cooling)                                                                   117
ECORA - Evaluation of computational fluid dynamic methods for reactor safety analyses       119
EUROFASTNET - European group for future advances in sciences and technology for
nuclear engineering thermalhydraulics                                                       121
CRISSUE-S - Revisiting critical issues in nuclear reactor design / safety by using 3-D
neutronics / thermalhydraulics models: state-of-the-art                                     123
VALCO - Validation of coupled neutronics/thermal hydraulics codes for VVER reactors         126
RMPS - Reliability methods for passive safety functions                                     128
NACUSP - Natural circulation and stability performance of BWRs                              130
DEEPSSI - Design and development of a steam generator emergency feedwater passive
system for existing and future PWR's using advanced steam injectors                         132
FABIS - Fast-acting boron injection system                                                  134
CERTA - European network for the consolidation of the integral system experimental
data bases for reactor thermal-hydraulic safety analysis                                    136
ITEM - Improvement of techniques for multiscale modelling of irradiated materials           138
High burn-up and MOX fuel
MICROMOX - The influence of microstructure of MOX fuel on its irradiation behaviour
under transient conditions                                                          141
OMICO - Oxide fuels: microstructure and composition variations                              143
VALMOX - Validation of high burnup mox fuels calculations                                   145
SIRENA - Simulation of radiation effects in Zr-Nb alloys: application to stress corrosion
cracking behaviour in iodine-rich environment                                               147
EXTRA - Extension of transuranus code applicability with Nb containing cladding models 149




                                               3
Nuclear Energy Programme                                      Plant life extension and
                                                              management
Operational safety of existing installations - RI             General

Title:
JOINT SAFETY RESEARCH INDEX

Acronym                 JSRI

Proposal number FIS5-1999-00302                       Contract number FIR1-CT2000-20089

Type of action          Concerted action              Duration              30 months

Starting date           1 January 2001                EC project officer G. Van Goethem

Total budget*           307.784 €                     EC contribution* 299.909 €


Co-ordinator
         Organisation          Gesellschaft für Anlagen - und Reaktorsicherheit (GRS) GmbH
                               Research Management Division
         Address               Schwertnergasse 1
                               D-50667 Köln
         Contact person        Dr. Axel Breest
                               Tel:      (49-221) 2068667
                               Fax:      (49-221) 2068629
                               Email     bre@grs.de

Partnership

          Country                   Organisations

            INT                 European Commission - JRC/ISIS
            E                   Centro de Investigaciones Energéticas, Medioambientales y
                                Tecnológicas (CIEMAT)
            RO                  Center of Technology and Engineering for Nuclear Projects
                                (CITON)
            NL                  Nuclear Research and Consultancy Group (NRG)
            I                   Ente per le Nuove Tecnologie, l'Energia e l'Ambiente (ENEA)
            D                   Forschungszentrum Karlsruhe GmbH (FZK)
            UK                  Health and Safety Executive (HSE)
            F                   Institut de Radioprotection et de Sureté Nucléaire (IRSN)
            CZ                  Nuclear Research Institute Rež plc (NRI)
            HU                  Nuclear Services Ltd
            CH                  Paul Scherrer Institute (PSI)
*
  Total eligible costs and EC contribution reduced respectively to 299.909 € and 292.034 € following
the changes in the consirtium composition.


                                                 4
          DK                  RISOE National Laboratory
          B                   Belgian Nuclear Research Centre (SCK-CEN)
          S                   Swedish Nuclear Power Inspectorate (SKI)
          FIN                 Technical Research Centre of Finland (VTT)

Project Summary
Reactor safety research is in principle the responsibility of the governments of States with
nuclear energy programmes to guarantee rigorous safety standards within their territories.
However, as the consequences of hypothetical nuclear accidents are not limited by the
boundaries of the countries on whose territory such an accident might occur, besides nuclear
power plant operating countries also countries not applying nuclear energy are performing
safety research with respect to nuclear installations.
The safety issues to be addressed in the various countries are similar as are the reactor
designs used, especially in EU Member States, and international co-operation is practised in
various fields of investigation to bundle research capacities, exchange information, and to
avoid duplication of work. To support such international information exchange and co-
operation an information tool is intended to be provided by this project which facilitates an
overview on light water reactor safety research currently performed in Member Countries
and countries associated to the European Union. It is the goal to prepare an Index to be
presented in the internet containing brief reports on current research projects in participating
countries. These reports describe the project objectives, work scope, approach and status.
Finished projects shall be dropped from the Index because there are numerous sources
providing detailed information on these.
The Index shall be updated every year so that the information provided is closely linked to
the research just performed in EU member and applicant countries. The timely information
on research work under way, recent results and activities planned for the immediate future
qualifies this database to be a strategic tool for further development of reactor safety research
programmes on the national as well as the international level.
The project is based on the status reached during the preceding Joint Safety Research Index
(JSRI) project performed under the 4th EU Framework Programme (1994-1998). This
activity resulted in preparing two releases of the JSRI database which were distributed on
CD-ROM. According to the experience gained in the proposed project the JSRI shall be
placed in the internet. To obtain full advantage of the possibilities offered by internet
databases, access to and retrieval of information from the JSRI shall be stepwise enhanced.
The 2000 issue of the JSRI shall be released in autumn 2001. This issue shall be based on the
standards agreed upon in the previous project and incorporate the input of new partners,
preferably from eastern European countries.
Further JSRI issues on research projects conducted in 2001 and 2002 are planned to be
released in autumn 2002 and 2003 respectively. These releases shall reflect the feedback
from participants based on experience made with previous issues of the Index.
Besides continuous communication between participants and co-ordinator, detailed
information exchange and discussions on further improvements of the JSRI shall be achieved
by meetings (one per planned JSRI release in 2001, 2002 and 2003).
In the final meeting to be held in late 2003 the experiences with the Joint Safety Research
Index shall be summarised and recommendations shall be given for future continuation.



                                               5
Nuclear Energy Programme                                                Plant life extension and
                                                                        management
Operational safety of existing installations                            Integrity of equipment and
                                                                        structures

Title:
FRACTURE MECHANICS BASED EMBRITTLEMENT TREND CURVES FOR THE
CHARACTERISATION OF NUCLEAR PRESSURE VESSEL MATERIALS

Acronym                  FRAME

Proposal number FIS5-1999-00325                                Contract number FIKS-CT2000-00101

Type of action           Shared cost                           Duration                     36 months

Starting date            1 September 2000                      EC project officer P. Manolatos

Total budget             858.905 € *                           EC contribution              429.453 € *


Co-ordinator
         Organisation              Technical Research Centre of Finland (VTT)
         Address                   Kemistintie 3
                                   FIN-02044 Espoo
         Contact person            Mr. Matti Valo
                                   Tel:      + 358-9-4566383
                                   Fax:      + 358-9-456 6479
                                   Email     matti.valo@vtt.fi

Partnership

          Country                       Organisations

        CZ              Nuclear Research Institute Rež plc (NRI)
        INT             European Commission - JRC/IE
        FIN             Fortum Nuclear Services Ltd
        B               Belgian Nuclear Research Centre (SCK-CEN)
        HU              AEKI (*)
_______________________________
* Budget expanded following inclusion of new partners from Newly Associated States (NAS).


Project Summary
Lifetime of a nuclear plant is ultimately limited by ageing of non-replaceable components
like the pressure vessel. Cleavage initiation fracture toughness is the property, which is
needed in the structural safety analyses of the vessel. However, this property is not
measured directly for the irradiated (neither for the annealed or re-irradiated) material
condition, instead a correlative embrittlement estimation based on the Charpy-V test is used.
It is difficult to quantify the uncertainties inherent in the current estimation and hence the


                                                         6
assumed uncertainties are addressed by the use of a conservative fracture toughness
reference curve and by added margins. Charpy-V impact toughness is in many respects a
clearly different material property than fracture toughness. Hence the current understanding
of embrittlement may be a biased one.

In the FRAME project fracture toughness based embrittlement models will be created and
they will be critically compared with the published Charpy-V based models. Model alloys
and pressure vessel steels will be included in the test matrix. The model alloys allow a
relatively large variation of the critical impurity elements, i.e. Cu, P and Ni, to be achieved
but the pressure vessel steels stand for real steels. Fracture toughness based trend curves do
not exist nowadays, because the required databases are non-existing or they are insufficient
in size. Trend curve development is in essence mathematical fitting of candidate functions to
measured irradiation shift data. It is clear that the FRAME project can identify only the main
response of impurity elements and their synergism to embrittlement. However, by FRAME
a physically correct base for embrittlement description will be established and possibly non-
conservatism in the current procedure will be identified.

Approximately twelve different materials can be included in the test matrix. It is considered
very essential that the materials experience the same irradiation condition. The relatively
large scatter of Charpy-V based trend curves is assumed to be due to material
inhomogeneities and variability of environmental parameters in surveillance databases.

VTT (FIN), SCK-CEN (BE), NRI-REZ (CZ), JRC Petten and Fortum Ltd (FIN) are the
project partners. The first three partners perform the experimental work in the project.
Fortum Ltd is a utility, which has paid much attention to vessel embrittlement and its
mitigation in his plants including vessel anneal. They will apply the created data in a
complementary analyses of their plants. JRC will perform material irradiation for the
project. Due to the relative high cost of fracture toughness testing in hot cells three testing
partners are required for sharing the own portion of the project costs and for shortening the
time required for testing.




                                              7
Nuclear Energy Programme                                    Plant life extension and
                                                            management
Operational safety of existing installations                Integrity of equipment and
                                                            structures

Title:
RETROSPECTIVE DOSIMETRY FOCUSSED ON THE REACTION 93NB(N,N')

Acronym                 RETROSPEC

Proposal number FIS5-1999-00305                       Contract number FIKS-CT2000-00091

Type of action          Shared cost                   Duration          25 months

Starting date           1 October 2000                EC project officer P. Manolatos

Total budget            318.982 €                     EC contribution   159.491 €


Co-ordinator
         Organisation          Nuclear Research and Consultancy Group (NRG)
                               Materials, Monitoring & Inspection
         Address               Westerduinweg 3
                               NL-1755 ZG Petten
         Contact person        Dr. Willem Voorbraak
                               Tel:     (31-224) 564295
                               Fax:     (31-224) 564457
                               Email    voorbraak@nrg-nl.com

Partnership
          Country                     Organisations

            B                   Belgian Nuclear Research Centre (SCK-CEN)
            FIN                 Technical Research Centre of Finland (VTT)

Project Summary
Accurate data on the neutron fluence combined with information from data bases with
material properties for the various structural materials will give advance information on the
condition of the various components of a Nuclear Power Plant (NPP), including the Reactor
Pressure Vessel (RPV). Finally this information will determine the End-Of-Live (EOL) of
that RPV. In Pressurised Water Reactors (PWR) this is the decisive factor determining the
life span of the whole NPP.

Most of the West European NPP’s have a well-defined RPV surveillance program in which
the neutron fluence is monitored also. Such a program informs about the degradation of the




                                                 8
RPV materials at EOL. This is not always the case for some of the East European reactor
types.

This lack of real monitors requires an alternative approach. Small amounts (about 1 to 10
mg) of the structural materials itself will be used instead. The material has to be made
available in the form of small chips or scrapings obtained by nibbling, scraping or drilling
using robotic tools. The procedure will be focussed on the reaction 93Nb(n,n’)93Nbm. This
reaction is very attractive because of its long half-life and therefore less sensitive for the
irradiation history. A procedure will be developed which can be applied by most
laboratories, which have a physical as well as a chemical laboratory. A systematic approach
is followed taking into account different materials and a pessimistic relation between cooling
time and half-life of cobalt. This approach will be reported and can be considered as a Code
of Practice, which can be applied by average-level laboratory instrumentation. The method
will help to improve the accuracy in experimental neutron dose estimation and contribute to
a more exact and firmly based determination of the End-Of-Life.




                                              9
Nuclear Energy Programme                                    Plant life extension and
                                                            management
Operational safety of existing installations                Integrity of equipment and
                                                            structures

Title:
PHOSPHORUS INFLUENCE ON STEEL AGEING

Acronym                 PISA

Proposal number FIS5-1999-00269                       Contract number FIKS-CT2000-00080

Type of action          Shared cost                   Duration          36 months

Starting date           1 December 2000               EC project officer P. Manolatos

Total budget            1.618.840 €                   EC contribution   794.912 €


Co-ordinator
         Organisation          AEA Technology Plc
                               Nuclear Science
         Address               220 Harwell
                               UK-OX11 0RA Didcot, Oxfordshire
         Contact person        Dr. Colin English
                               Tel:      (44-1235) 434342
                               Fax:      (44-1235) 435941
                               Email     colin.english@aeat.co.uk

Partnership
          Country                     Organisations

            INT                 European Commission - JRC/IE
            F                   Electricité de France (EDF)
            D                   Framatome ANP GmbH
            UK                  British Nuclear Fuel plc (BNFL)
            HU                  KFKI Atomic Energy Research Institute (AEKI)
            FIN                 Technical Research Centre of Finland (VTT)
            CZ                  Nuclear Research Institute Rež plc (NRI)
            E                   Tecnatom S.A.
            UK                  The University of Liverpool
            D                   Staatliche Materialprufungsanstalt (MPA Stuttgart)

Project Summary
The integrity of the pressure vessel is vital to the safe operation of a nuclear reactor. It is
therefore necessary to monitor or predict the changes in the pressure vessel material during



                                                 10
operation. Exposure to irradiation (or elevated temperatures) causes the segregation of
phosphorus to internal grain boundaries in RPV steels. This, in turn, encourages brittle
intergranular failure of the material. The objectives of PISA are to improve predictability of
a failure mechanism that can affect all types of reactor plant operating in Europe, and in
particular to improve the predictability of mechanical property changes in long service steels
for plant applications.

The approach employed to achieve this objective is to improve predictability through
developing improved physical understanding of both the segregation process and any
resultant change in mechanical properties. The necessary understanding will be developed
through experimental investigations of irradiated steels and model alloys, with associated
modelling studies. In addition, a critical aspect of the experimental measurements is the
methodology to the determination of the level of segregants on the grain boundaries,
particularly P and C, and here further technique development is required.

The integrity of the 'Reactor Pressure Vessel (RPV) is vital to the safe operation of a nuclear
reactor. It is therefore necessary to monitor or predict the changes in the pressure vessel
material during the operation. Exposure to irradiation (or elevated temperatures) causes the
segregation of phosphorus to internal grain boundaries in RPV steels. This, in turn,
encourages brittle intergranular failure of the material.

There is a need to develop deep understanding of two aspects of this ageing mechanism.
First, it is necessary to improve the experimental database on the segregation occurring in
representative steels and to establish the exact dependence of the segregation under
irradiation on flux, fluence and irradiation temperature, as well as metallurgical variables
such as phosphorus level, or internal state of the grain boundary. This data would also serve
the purpose of providing critical data to validate models of irradiation induced segregation.
Second it is important to investigate the conditions under which inter granular failure
becomes the dominant failure mechanism, and the consequential effects of the mechanical
properties. More specifically, this requires determining the effect of the coverage" of
phosphorus on the grain boundary on the failure mechanism, and once inter granular failure
occurs the effect of increased levels of phosphorus on the fracture toughness or impact
properties of the material. Such understanding is not available at present, and is required to
make predictions of the service conditions where this ageing mechanism is likely to be
important, particularly when life extension is considered.

The range of the RPV steels to be considered includes the MnMoNi steels employed in
European PWRs; the mild steels used in UK Magnox (steel) RPVs; and the steels employed
in VVER 440's. Intergranular fracture and/or P segregation is considered to be important in
plant applications involving all three reactor types.




                                              11
Nuclear Energy Programme                                      Plant life extension and
                                                              management
Operational safety of existing installations - RI             Integrity of equipment and
                                                              structures

Title:
REACTOR NEUTRONIC INVESTIGATIONS ON LR-0 REACTOR

Acronym                 RENION

Proposal number FIS5-2002-00052                         Contract number FIR1-CT2002-40157

Type of action          Research Infrastructures        Duration           24 months

Starting date           1 February 2003                 EC project officer A. Zurita

Total budget            137.444 €                       EC contribution    137.444 €


Co-ordinator
         Organisation          Ústav Jaderného Výzkumu Rež A.S.
                               Nuclear Power and Safety Divison
         Address
                               CZ-25068 Řež
         Contact person        Mr. Ivo Vasa
                               Tel:     (420 -2) 209 410 20
                               Fax:     (420-2) 209 410 29
                               Email    rub@ujv.cz; vas@nri.cz


Project Summary
The RENION project is designed to enable access to the experimental facility (LR-0 reactor)
for specialists coming from a number of European Community and Associated countries
(especially support will be given to young generation of experimental and theoretical
physicists). The experimental reactor LR-0 should enable different users, selected by a
selection panel, to realise the experimental projects related to VVER and PWR reactor
physics that can be utilised to extend their experimental databases and to validate the
computer codes. It is also expected that the obtained results may be applicable for PWRs, for
instance – for the relevant computer codes validation.

As a result, the project should contribute to the implementation of the EU policy in
supporting competitiveness of the nuclear option with regard to the existing VVER
installations (VVER-440 and VVER-1000) in the associated countries. Sharing of users
experience has in the past proved to be extremely beneficial in addressing reactor physics
experimental investigations (PWR and VVER), so continuation of such international co-
operation should allow to maintain and to extend the EU competence in the area of reactor
experimental research.


                                                   12
The project will start up with user selection panel meeting together with users group kick-off
meeting. There will be specified precise experimental programme for work-packages No. 1
and 2. The starting experimental activity are preliminary suppose to concentrate on the
measurement in a mock-up in LR-0 reactor. NRI will give all necessary information with
respect to RENION project (contact person, propose and approved working-package plan,
the description of the infrastructure including experimental equipment and procedures, etc.)
on NRI web side (www.nri.cz - alias name www.ujv.cz). The users, which are listed bellow,
have yet signed Letter of Interest to take part in the project.

Specialists from VVER operating countries expressed their interest in research at LR-0
reactor during WG meetings and AER conferences. It is expected that other potential users
from the Nuclear Power Plants (VVER, PWR) or Academic and Research Institutes could
take part in the project.

The design of LR-0 reactor permits easy rearrangement of the reactor core as well as
modifications of operational modes according to requirements of the particular experiment.
Within the reactor vessel, the core is supported by a supporting plate, the standard one
accommodates VVER-1000 fuel assemblies in the triangular lattice with the pitch of 236
mm. There are at our disposal VVER-1000 type elements for 68 fuel assemblies of different
enrichment (1.6%, 2 %, 3%, 3.3 %, 3.6% and 4.4 % U235). For experiments with VVER-440
type assemblies the special supporting plate allows arranging the necessary 147-mm pitch.
The same special supporting plate is used in other experiments that require different pitches
in the triangular lattice. The plate has radial grooves that provide radial fitting of the
assembly heel’s sliding nests. Angles between the grooves are chosen so that at each fixed
pitch the assembly lattice is triangular and symmetrical with respect to the plate centre. The
nests can be changed to accommodate either VVER-1000 or VVER-440 type fuel
assemblies, and with a small modification – also the PWR type ones.




                                             13
Nuclear Energy Programme                                    Plant life extension and
                                                            management
Operational safety of existing installations                Integrity of equipment and
                                                            structures

Title:
IRRADIATION EFFECTS ON THE EVOLUTION OF THE MICROSTRUCTURE,
PROPERTIES AND RESIDUAL STRESSES IN THE HEAT AFFECTED ZONE OF
STAINLESS STEEL WELDS

Acronym                 INTERWELD

Proposal number FIS5-1999-00332                       Contract number FIKS-CT2000-00103

Type of action          Shared cost                   Duration          42 months

Starting date           1 September 2000              EC project officer P. Manolatos

Total budget            1.739.960 €                   EC contribution   660.631 €


Co-ordinator
         Organisation          Nuclear Research and Consultancy Group (NRG)
                               Materials, Monitoring & Inspection
         Address               Westerduinweg 3
                               NL-1755 ZG Petten
         Contact person        Mr. Bob Van der Schaaf
                               Tel:     (31-224) 564665
                               Fax:     (31-224) 568490
                               Email    vanderschaaf@nrg-nl.com

Partnership

          Country                     Organisations

            D                   Framatome ANP GmbH
            INT                 European Commission - JRC/IE
            CH                  Paul Scherrer Institute (PSI)
            B                   Belgian Nuclear Research Centre (SCK-CEN)
            E                   Centro de Investigaciones Energéticas, Medioambientales y
                                Tecnológicas (CIEMAT)

Project Summary
The overall objective of this research project is to help define the radiation induced material
changes that promote cracking in the heat affected zone of PWR and BWR core internal
components. In order to reach these overall objective the following objectives can be
distinguished within the project:



                                                 14
-  fabrication of an industrial LWR core relevant weldment with representative residual
   stresses, microstructure and properties,
- irradiation of welded coupons under relevant LWR internals neutron fluence conditions,
- determine the evolution of weld residual stresses under neutron irradiation conditions,
- determine the stress corrosion behaviour of the material under neutron irradiation
   conditions and
- determine the (micro)mechanical, microstructural and microchemical properties of the
   weld material under neutron irradiation conditions.
Finally, an assessment will be made of the correlation between weld residual stresses,
microstructure/microchemistry and the stress corrosion behaviour.

In practice, test welds of AISI 304 and AISI 347 stainless steel will be produced with weld
residual stresses, microstructure and properties representative for core shroud application.
These welds will be characterised for weld residual stress state prior to irradiation by
destructive (ring core) and non-destructive (neutron and X-ray diffraction) methods.
Coupons and test specimens of the test weld will be irradiated in two materials test reactors
to two relevant neutron dose levels. An in-service weld from a decommissioned reactor will
be used to compare the results from the test weld with real internal component material. The
irradiated materials will be distributed to the different partners to perform the post-irradiation
test and examination campaign. The weld residual stresses will be measured by neutron
diffraction on the low and high dose level test weld coupons and coupons from the
unirradiated and irradiated in-service material. The corrosion behaviour of the material will
be determined by CERT and EPR tests in BWR and inert environment. The
(micro)mechanical properties will be determined both on irradiated test specimens and
specimens taken from the irradiated coupons and in-service material. The microstructure and
microchemistry of the weld, heat affected zone and plate structure will be examined by
optical, SEM, TEM, confocal, EPMA, STEM-EDX, SIMS and AUGER techniques. Finally,
the results from the weld residual stress measurements, the corrosion behaviour, the
(micro)mechanical properties and the microstructural and microchemical features will be
synthesised in order to assess the correlation between the weld residual stresses and
microstructure/microchemistry after neutron irradiation and the specific stress corrosion
resistance of the core shroud weldment and to deduce indications on the mechanism of the
cracking process and the controlling parameters, in particular the importance of the weld
residual stresses vs. local microstructure/chemistry.




                                               15
Nuclear Energy Programme                                     Plant life extension and
                                                             management
Operational safety of existing installations                 Integrity of equipment and
                                                             structures

Title:
PROPERTIES OF IRRADIATED STAINLESS STEELS FOR PREDICTING LIFETIME OF
NUCLEAR POWER PLANT COMPONENTS

Acronym                 PRIS

Proposal number FIS5-1999-00277                        Contract number FIKS-CT2000-00084

Type of action          Shared cost                    Duration          36 months

Starting date           1 October 2000                 EC project officer P. Manolatos

Total budget            1.143.531 €                    EC contribution   499.966 €


Co-ordinator
         Organisation          ABB Atom Ab
                               Nuclear Services
         Address               Gideonsbergsgatan, 2
                               S-72163 Västeras
         Contact person        Mr. Henrik Westermark
                               Tel:     (46-21)347000
                               Fax:     (46-21)348500
                               Email    henrik.westermark@se.westinghouse.com

Partnership

          Country                     Organisations

            E                   Centro de Investigaciones Energéticas, Medioambientales y
                                Tecnológicas (CIEMAT)
            F                   Framatome ANP
            B                   Belgian Nuclear Research Centre (SCK-CEN)
            D                   Framatome ANP GmbH
            S                   Studsvik Material AB
            FIN                 Technical Research Centre of Finland (VTT)

Project Summary
The objectives of the proposed project are to produce material data for irradiated austenitic
stainless steels of LWR internals as a function of fluence that can be used for structural
integrity and remaining lifetime assessments. The data will consist of validated initiation
fracture toughness, JIc, and fracture resistance curves (J-R), including tensile properties and



                                                  16
information on microstructural changes caused by irradiation. Materials from both BWR and
PWR internal components will be considered. The approach to achieve these objectives will
be to:
 validate a procedure for fracture resistance determination using sub-size specimens, i. e.
    measure the effect of specimen size and test type on fracture resistance using unirradiated
    reference materials with mechanical properties similar to irradiated stainless steels, and
    set criteria for specimen size and testing procedure in order to provide relevant fracture
    resistance data
 determine fracture resistance and tensile properties for irradiated austenitic stainless
    steels of LWR internals as a function of fluence
 determine microstructural and microchemical changes as a function of fluence (estimated
    fluence levels 0, 20 and 70 dpa)
The project is divided into seven work packages (WP). The first WP covers selection,
procurement and shipping of irradiated austenitic stainless steels from LWR RPV internal
components. Mechanical properties and microstructure of the irradiated materials will be
determined in WP2. This WP is divided into two parts, where the first is related to
characterisation of tensile properties and hardness as a function of fluence, as well as
microstructural studies (optical microscopy). Additionally, fracture properties of a unique
PWR component, thimble tube, with an estimated fluence ranging from 0 to 70 dpa will be
characterised using a pin-loading test technique. The second part of WP2 is connected to
investigations of the effects of fluence (estimated fluence levels 0, 20 and 70 dpa) on
microstructural and microchemical changes of the PWR thimble tube material.
WP3 covers selection and production of unirradiated reference materials for validation of the
fracture resistance testing with sub-size specimens. Work will commence with a literature
survey and a theoretical justification for the materials selected. Materials with mechanical
properties similar to irradiated materials will be produced for fracture resistance procedure
validation. Validation of the fracture resistance determination procedures using sub-size
specimens will be undertaken in WP4, using the reference materials from WP3. The outcome
of WP4 will be recommendations on specimen size and testing procedures for fracture
resistance testing of irradiated materials. With the irradiated materials defined (WP1), the
tensile properties determined (WP2), and validation of the fracture resistance testing
completed (WP4), validated fracture resistance data of irradiated materials will be
determined in WP 5.

All results from the project will be analysed and discussed in a detailed final report under
WP6. WP7 concerns co-ordination of the entire project. A Steering Committee will be
formed consisting of representatives for each partner of the project. The Steering Committee
will convene at least twice a year, and it will be the main tool for the project management.




                                              17
Nuclear Energy Programme                                    Plant life extension and
                                                            management
Operational safety of existing installations                Integrity of equipment and
                                                            structures

Title:
CRACK GROWTH BEHAVIOUR OF LOW ALLOY STEEL FOR PRESSURE BOUNDARY
COMPONENTS UNDER TRANSIENT LIGHT WATER REACTOR (LWR) OPERATING
CONDITIONS

Acronym                 CASTOC

Proposal number FIS5-1999-00198                       Contract number FIKS-CT2000-00048

Type of action          Shared cost                   Duration          36 months

Starting date           1 September 2000              EC project officer P. Manolatos

Total budget            1.431.759 €                   EC contribution   600.003 €


Co-ordinator
         Organisation          Staatliche Materialpruefungsanstalt (MPA Stuttgart)
                               Department for Environmental Effects
         Address               Pfaffenwaldring 32
                               D-70569 Stuttgart
         Contact person        Dr. Juergen Foehl
                               Tel:      (49-711) 6852564
                               Fax:      (49-711) 6852761
                               Email     foehl@mpa.uni-stuttgart.de

Partnership

          Country                     Organisations

            E                   Centro de Investigaciones Energéticas, Medioambientales y
                                Tecnológicas (CIEMAT)
            CZ                  Nuclear Research Institute Rež plc (NRI)
            CH                  Paul Scherrer Institute (PSI)
            D                   Framatome ANP GmbH
            FIN                 Technical Research Centre of Finland (VTT)

Project Summary
The life time of a nuclear power plant is decisively controlled by ageing processes. This
project addresses the ageing of primary pressure boundary components in particular the
reactor pressure vessel (RPV). It is conservatively postulated that the RPV contains flaws
which have penetrated the austenitic stainless steel cladding. Therefore, environmentally
assisted cracking (EAC) of the ferritic RPV material has to be considered as a major ageing


                                                 18
process. For plant life management (PLIM) data on crack growth rates under static and
cyclic loads must be available. With regard to the transferability of the laboratory test data to
practice the acting corrosion mechanisms have to be investigated and understood and have to
be verified in long-term experiments. The specific topic of this project is the investigation of
EAC in conjunction with transient conditions of water chemistry and loading.

The work programme is subdivided into 4 work packages WP 1 to WP 4. In this project data
on crack growth for low alloy ferritic steels of western type reactors and of Russian VVER
type reactors will be generated under water chemistry conditions applicable for both reactor
types. In each case two materials will be investigated, one with low and one with high
susceptibility to environmentally assisted cracking.

In WP 1 all participating institutions perform tests under nominally equal conditions with the
aim to demonstrate the variation in test results and to verify the general applicability of
laboratory test data. The work in WP 2 is focused on static loading conditions, in WP 3
predominantly on cyclic loading conditions in conjunction with long hold times. The static
tests will be carried out over long time periods to account for possible incubation phases, the
cyclic tests will be carried out with low frequency where corrosion processes are most
effective. In WP 4 all participating institutions evaluate the test results in a joint action with
regard to the applicability to practice. One of the major aspects is to give recommendation
for the implementation of the results into plant life management strategies and into Codes.




                                               19
Nuclear Energy Programme                                                Plant life extension and
                                                                        management
Operational safety of existing installations                            Integrity of equipment and
                                                                        structures

Title:
ASSESSMENT OF AGED PIPING DISSIMILAR METAL WELD INTEGRITY

Acronym                  ADIMEW

Proposal number FIS5-1999-00187                                Contract number FIKS-CT2000-00047

Type of action           Shared cost                           Duration                     36 months

Starting date            1 November 2000                       EC project officer P. Manolatos

Total budget             1.152.578 € *                         EC contribution              512.100 € *


Co-ordinator
         Organisation              Electricité de France (EDF)
                                   Engineering and Service Division - Septen
         Address                   12 -14 Avenue Dutrievoz
                                   F-69628 Villeurbanne Cedex
         Contact person            Mr. Claude Faidy
                                   Tel:     (33)-472827279
                                   Fax:     (33)-472827699
                                   Email    claude.faidy@edf.fr

Partnership
          Country                       Organisations

        F               Framatome S.A.
        F               Commissariat à l'Energie Atomique (CEA)
        UK              The Welding Institute
        UK              Serco Assurance
        FIN             Technical Research Centre of Finland (VTT)
        INT             European Commission - JRC/IE
        CH              Paul Scherrer Institute (PSI)
        HU              BZF (*)
_______________________________
* Budget expanded following inclusion of new partners from Newly Associated States (NAS).


Project Summary
ADIMEW aims to quantify the accuracy of structural integrity procedures used in the
European nuclear industry to ensure the safety of defect-containing dissimilar metal welds in
aged PWR Class 1 piping. Previous work on small-section bi-metallic welds will be


                                                         20
extended to industrial scale dissimilar metal welds at normal operating conditions. Various
forms of cracking have been observed in such welds between piping components in nuclear
power plants. Mixed mode loading, the variability of material properties, and residual
stresses across the weldment create problems for analysis methods to predict the cracking
behaviour and current engineering methods are considered overly conservative.

The focal point of the project will be a unique large-scale test to determine the actual
behaviour of cracks introduced into the surface of a ferritic-austenitic dissimilar metal weld
on pipes of an industrial scale. A cracked dissimilar metal weld forming a 16’’ diameter
piping assembly will be tested under conditions of four point bending at 300° C to determine
the load for crack initiation and subsequent tearing to collapse. Two welds between low
alloy A308/508 and austenitic 308/309 steel will be procured to a nuclear specification and
high quality control, and will contain a weld buttering layer at the ferritic interface. The
design of the test and the analysis of the results obtained will be supported by:

   a limited programme of materials testing will determine the tensile and fracture
    properties, while innovative multi-material testing techniques will be used to measure the
    property gradients at the weld interfaces.
   the residual stress field in the welds will be calculated numerically and determined
    experimentally by surface hole drilling and volumetric neutron diffraction measurements.
   the defect behaviour under the test loading regime will be analysed using established
    engineering methods and finite element analysis to establish the accuracy and
    conservatism of the different methods.

A detailed review of the program will provide recommendations on recommended flaw
assessment procedures for dissimilar metal welds containing cracks and on material property
testing standards for small multi-material specimens. It will also include a detailed synthesis
of the factors influencing the integrity of dissimilar welds. Transfer of the technology to
other parts of the European nuclear industry will be promoted through a formal link with the
Network for Evaluating Structural Components (NESC).




                                              21
Nuclear Energy Programme                                                Plant life extension and
                                                                        management
Operational safety of existing installations                            Integrity of equipment and
                                                                        structures

Title:
VALIDATION OF CONSTRAINT-BASED                                 ASSESSMENT             METHODOLOGY         IN
STRUCTURAL INTEGRITY

Acronym                  VOCALIST

Proposal number FIS5-1999-00303                                Contract number FIKS-CT2000-00090

Type of action           Shared cost                           Duration                     36 months

Starting date            1 October 2000                        EC project officer P. Manolatos

Total budget             1.747.338 € *                         EC contribution              746.169 € *


Co-ordinator
         Organisation              Serco Assurance
                                   Engineering Integrity Group
         Address                   Risley
                                   UK-WA3 6AT Warrington
         Contact person            Mr. David Lidbury
                                   Tel:     (44-1925) 252767
                                   Fax:     (44-1925) 252285
                                   Email    david.lidbury@sercoassurance.com

Partnership

          Country                       Organisations

        UK              British Nuclear Fuels plc (BNFL)
        F               Commissariat à l'Energie Atomique (CEA)
        F               Electricité de France (EDF)
        F               Framatome ANP
        D               Staatliche Materialprufungsanstalt (MPA Stuttgart)
        D               Framatome ANP GmbH
        FIN             Technical Research Centre of Finland (VTT)
        INT             European Commission - JRC/IE
        US              Oak Ridge National Laboratory
        D               E.ON Kernkraft GmbH
        CZ              Nuclear Research Institute Řež plc (NRI) (*)
___________________________
* Budget expanded following inclusion of new partners from Newly Associated States (NAS).




                                                         22
Project Summary
The pattern of crack-tip stresses and strains causing plastic flow and fracture in components
is different to that in test specimens. This gives rise to the so-called constraint effect. Crack-
tip constraint in components is generally lower than in test specimens. Effective toughness
is correspondingly higher. The fracture toughness measured on test specimens is thus likely
to underestimate that exhibited by cracks in components. The purpose of project
VOCALIST (Validation of Constraint-Based Methodology in Structural Integrity) is to
develop validated procedures for assessing crack-tip constraint in ageing nuclear pressure
boundary components. This is with the objective of achieving (i) an improved defect
assessment methodology for predicting safety margins; (ii) improved lifetime extension
arguments.
The project consists of the following six work packages (WP):
WP1: co-ordination and project management.
WP2: compilation of Handbook. In this initial phase Issue 1 of a Handbook will be
produced detailing the application of constraint-based fracture mechanics procedures based
on current best practice. Gaps in knowledge and understanding will be identified which
currently limit the assessment of defects in ageing components.
WP3: adoption of Benchmark tests and performance of structural Features tests. First,
existing large-scale fracture experiments will be identified in relation to the issues raised in
WP2. Particular reference will be made to the NESC (Network for Evaluation of Structural
Components) series of tests. Archive materials will be physically located and basic material
properties data compiled. Second, innovative fracture experiments (structural Features tests)
will be designed, procured and executed using the relevant archive materials to simulate, in
reduced scale tests, the constraint conditions applicable to defects in the corresponding large-
scale Benchmark experiments.
WP4: analysis. This work package will interact strongly with WP3. The initial analyses
will be concerned with calibrating fracture models using basic properties of the archive
materials. The calibrated models will then be used to design the structural Features tests and
predict their outcome. Data from the Features tests together with the results from further
analyses will be used to produce improved predictions of the original Benchmark
experiments. Comparisons between the analytical predictions and experimental results
during the various phases of this process will be used to verify and validate the constraint-
based procedures.
WP5: synthesis and update of best practice. The improved methodology assessing defects in
aged components will be based on an overall synthesis of the results obtained in WP2 to
WP4. This methodology will be detailed in Issue 2 of the Handbook of best practice
originally produced as part of WP2.
WP6: programme evaluation, including conclusions and recommendations.
The overall success of the project will be measured by the extent to which it has provided
Europe’s nuclear plant operators and their regulators with a practical methodology for
making/considering:
 Improved assessments of safety margins for aged pressure boundary components under
    normal and abnormal loadings
 Improved lifetime extension arguments for aged pressure boundary components
    consistent with maintaining current safety standards




                                               23
This measurement will be achieved both by an objective evaluation of results and
achievements in identified reports to DG XII of the EC throughout the lifetime of the project,
and independently through an ongoing process of peer review by virtue of the association
between VOCALIST and NESC.




                                             24
Nuclear Energy Programme                                    Plant life extension and
                                                            management
Operational safety of existing installations                Integrity of equipment and
                                                            structures

Title:
STRUCTURAL MARGIN IMPROVEMENTS IN AGED-EMBRITTLED RPV WITH LOAD
HISTORY EFFECTS

Acronym                 SMILE

Proposal number FIS5-2001-00023                       Contract number FIKS-CT2001-00131

Type of action          Shared cost                   Duration           36 months

Starting date           1 January 2002                EC project officer P. Manolatos

Total budget            1.725.053 €                   EC contribution    787.526 €


Co-ordinator
         Organisation          Electricité de France (EDF)
                               Division Production Nucléaire
         Address               Site Cap Ampère, 1 Place Pleyel
                               F-93282 Saint-Denis
         Contact person        Mr. Georges Bezdikian
                               Tel:     (33-1)43693848
                               Fax:     (33-1)43693482
                               Email    georges.bezdikian@edf.fr

Partnership

          Country                     Organisations

            UK                  Serco Assurance
            F                   Commissariat à l'Energie Atomique (CEA)
            F                   Ministere de l'Economie, des Finances et de l'Industrie
            D                   Staatliche Materialprufungsanstalt (MPA Stuttgart)
            UK                  British Energy Generation Ltd
            F                   Framatome ANP
            D                   Framatome ANP GmbH
            INT                 European Commission - JRC/IE
            D                   E. ON Kernkraft GmbH
            US                  Oak Ridge National Laboratory




                                                 25
Project Summary
The Reactor Pressure Vessel (RPV) is an essential component liable to limit the lifetime
duration of PWR power plants. The assessment, at an European level, of defects in RPV
subjected to PTS transients does not take into account the beneficial effect of load history /
warm pre-stressing (WPS). The aim of the SMILE project is to better understand this effect
in a RPV structural integrity assessment, and to define and to establish some
recommendations for a pre-codification in European Codes and Standards.
Within the framework of this project, all elements necessary to propose a method to take into
account this effect will be gathered or obtained. This will be done through experimental
works, leading to a deep understanding of metallurgical and mechanical phenomena, and
through numerical works and development of models. The results obtained will permit a
much more precise prediction of a possible fracture in a RPV submitted to a PTS transient.
Finally, this project aims to harmonise the different approaches in European Codes &
Standards regarding the inclusion of the WPS effect in the RPV integrity assessments. Some
guidelines will be prepared with this purpose.

The SMILE project is organised in 6 work-packages, in harmony with the NESC and
VOCALIST projects :
WP1 : Co-ordination and Management
WP2 : Calibration tests
It aims experimentally, by tests on small specimens, to check the beneficial effect of WPS in
like-reactor conditions.
WP3 : Assessment of models
Validation and comparison of available theoretical models. Some benchmarks are performed
on two chosen tests. The calibration tests are interpreted by the partners. Consistence of
predictions between partners and a good correlation with experimental results are the
necessary conditions for the validation of the models.
WP4 : Validation tests
A simulation of PTS transient will be performed on a large scale vessel-like specimen. This
should lead to an experimental demonstration of the WPS effect in real conditions.
WP5 : Cases studies
Two additional numerical applications are performed on real PTS transients taking into
account a subclad flaw and a through-clad surface crack.
WP6 : Programme evaluation, synthesis and final recommendations
Some guidelines and recommendations are proposed for a pre-integration in European Codes
& Standards.

More precisely, the objectives of the SMILE project are the following :
- Good understanding of fundamental mechanisms : all elements necessary to an in-depth
understanding of the origin of the beneficial effect of Warm Pre-Stressing (WPS) will be
obtained and validated
- Validation test : simulation of Pressurised Thermal Shock (PTS) conditions using a model
vessel with a circumferential shallow crack submitted to combined thermomechanical
loading. The objective of this test is to produce a pronounced preloading in the upper shelf
region of fracture toughness before the cleavage initiation at lower temperature. This
experiment must demonstrate the warm pre-stress effect under conditions very close to
realistic PTS loading scenarios
- Assessment of models : the theoretical and numerical tools needed to interpret the project
experimental works will be evaluated. It includes a critical review of already existing


                                             26
models. Both global (e.g. Curry or Chell models) and local (e.g. Beremin model) approaches
will be investigated. Finally these models will be implemented into numerical finite elements
codes
- Demonstration of the capabilities of numerical studies : the models will be used to
interpret the calibration and validation tests. This should lead to the demonstration of their
capability to anticipate a fracture event (or the level of reloading needed to obtain fracture)
of a vessel submitted to thermal transient exhibiting a pre-loading in the upper shelf of the
transition curve. Some numerical benchmarks will be also performed in order to test the
numerical implementation of models and to compare the different numerical finite elements
tools on some applications
- Elaborate synthesis, recommendations and guidelines for Codes and Standards : a
synthesis of all theoretical, numerical and experimental elements will be prepared related to
the inclusion of WPS in a RPV structural integrity assessment. Some guidelines and
recommendations for a pre-codification will be proposed to be introduced in various
European and US Codes and Standards




                                              27
Nuclear Energy Programme                                                Plant life extension and
                                                                        management
Operational safety of existing installations                            Integrity of equipment and
                                                                        structures

Title:
THERMAL FATIGUE EVALUATION OF PIPING SYSTEM "TEE"- CONNECTIONS

Acronym                  THERFAT

Proposal number FIS5-2001-00043/58                             Contract number FIKS-CT2001-00158

Type of action           Shared cost                           Duration                     36 months

Starting date            1 December 2001                       EC project officer P. Manolatos

Total budget             1.679.925 € *                         EC contribution              839.963 € *


Co-ordinator
         Organisation              E.ON Kernkraft GmbH
                                   Dept. of mechanics, materials, non destructive testing (TTF)
         Address                   Tresckowstrasse 5
                                   D-30457 Hannover
         Contact person            Mr. Klaus-Jürgen Metzner
                                   Tel:     (49-511)4394009
                                   Fax:     (49-511)4394377
                                   Email    klaus-juergen.metzner@eon-energie.com
Partnership
          Country                       Organisations
             F          Electricité de France (EDF)
             F          Framatome ANP
             F          Commissariat à l'Energie Atomique (CEA)
             E          Tecnatom S.A.
             D          Fraunhofer-Gesellschaft zur Foerderung der Angewandten
                        Forschung e.V.
        FIN             Technical Research Centre of Finland (VTT)
        D               Framatome ANP GmbH
        D               Staatliche Materialprufungsanstalt (MPA Stuttgart)
        INT             European Commission - JRC/IE
        FIN             Fortum Nuclear Services Oy
        D               Siempelkamp Pruef- und Gutachter-Gesellschaft mbH
        UK              Cinar Ltd.
        SK              Nuclear Power Plant Research Institute (VUJE) Trnava Inc
        E               Endesa Generacion S.A.
        SI              Jozef Stefan Institute (*)
___________________________
* Budget expanded following inclusion of new partners from Newly Associated States (NAS).



                                                         28
Project Summary
Thermal fatigue is a recurring problem when LWR plants become older and life time
extension activities are initiated. In general, the common thermal fatigue issues are
understood and can be controlled by plant instrumentation systems. However, the Civaux 1
and other incidents indicate that certain piping system Tee's are susceptible to turbulent
temperature mixing effects that cannot be adequately monitored by common thermocouple
instrumentations, putting the reliability of integrity evaluation procedures in doubt.
THERFAT proposes to review field data and perform advanced thermohydraulic flow
simulations and stress and fracture analysis. Critical elements of the procedure will be
investigated by targeted verification tests. Proposals will be made for improved load thermal
fatigue assessment procedures, screening criteria and for establishing a European
Methodology on Thermal Fatigue.
The project is managed in a series of five coherent work-packages to integrate data
collection, state-of-the-art assessments, and experimental verification. In-service experience
of thermal fatigue in mixing Tees will be collated and analysed in WP1. In WP2, thermo-
hydraulic loads in mixing representative Tees will be investigated by experiments and
advanced thermo-hydraulic analyses, with special emphasis on the heat transfer between
fluid and the wall. The tests will be performed using Plexiglas mock-ups with fully
instrumented steel segments at selected locations. A virtual sensor system based on neuro-
fuzzy concepts will be trained and developed for the thermal fatigue problem. An assessment
will be made of the reliability of the analyses to describe rapidly fluctuating flow behaviour
and quantify thermal loads on the Tee. These loads will be used as input in WP3, Integrity
Evaluation, which contains three parts: (a) determination of stresses in Tees induced by the
turbulence loads using 3-D elasto-plastic finite element analyses, as well as simpler
engineering methods (b) the computed stresses are analysed to predict damage initiation
using fatigue curves or local strain criteria (c) fracture analysis of discrete cracks using
different levels of complexity for loads, crack geometry and growth criteria. In WP4,
existing thermal fatigue tests will be reviewed and the most relevant selected as benchmarks
to verify procedures in WP3. Carefully targeted thermal fatigue tests will be performed to
allow check issues identified in the preceding WP’s. These include small straight pipes to
investigate the effect of welds and variable amplitude loads, and larger pipe segments and
Tees under high frequency loads. The results will be used to assess the sensitivity of various
parameters, to quantify safety margins and support recommendations for improved
instrumentation. An overall evaluation will be done in WP5, defining a road-map for a
"European methodology on Thermal Fatigue" and identifying full-scale verification tests.
THERFAT is expected to generate the following main results and deliverables.
     Assessment of field data on thermal fatigue of Tees.
     Capabilities of thermo-hydraulic analysis to capture turbulent thermal loads, verified
        by experiments.
     Development of virtual sensor for load monitoring.
     Advanced analysis of TF damage and impact on operational screening criteria.
     Verification of TF damage quantification using existing and new test data.
     Final Report on evaluation procedures for initiation and propagation of thermal
        fatigue in Tees and strategy for a "European Methodology on Thermal Fatigue".




                                             29
Nuclear Energy Programme                                    Plant life extension and
                                                            management
Operational safety of existing installations                Integrity of equipment and
                                                            structures

Title:
TWO-PHASE FLOW WATER HAMMER TRANSIENTS AND INDUCED LOADS ON
MATERIALS AND STRUCTURES OF NUCLEAR POWER PLANTS

Acronym                 WAHALOADS

Proposal number FIS5-1999-00114/341                   Contract number FIKS-CT2000-00106

Type of action          Shared cost                   Duration          36 months

Starting date           1 October 2000                EC project officer G. Van Goethem

Total budget            2.089.758 €                   EC contribution   1.269.869 €


Co-ordinator
         Organisation          Université Catholique de Louvain (UCL)
                               Unité Thermodynamique
         Address               Place du Levant, 2
                               B-1348 Louvain-la-Neuve
         Contact person        Prof. Michel Giot
                               Tel:      (32-10) 472210
                               Fax:      (32-10) 452692
                               Email     giot@term.ucl.ac.be

Partnership

          Country                     Organisations

            F                   Commissariat à l'Energie Atomique (CEA)
            E                   Iberdrola S.A.
            SI                  Institute "Josef Stefan"
            B                   Tractebel S.A.
            F                   Electricité de France (EDF)
            E                   Empresarios Agrupados Internacional S.A.
            D                   Framatome ANP GmbH
            HU                  KFKI Atomic Energy Research Institute (AEKI)
            D                   Fraunhofer-Gesellschaft zur Förderung der angewandten
                                Forschung e.V./ UMSICHT
            D                   Forschungszentrum Rossendorf (FZR)




                                                 30
Project Summary
The project aims at the elaboration of improved and innovative tools and methods for
maintaining and improving the safety of existing reactor installations. The global objective is
to predict the loads on equipment and support structures, which are caused by water
hammers and shock waves. In particular, the following goals are set:
 review, evaluation and selection of existing experimental data
 supply of new experimental data on water hammer using innovative two-phase flow
   instrumentation and including the measurement of loads on supports
 supply of new experimental data on dynamic stresses in equipment walls
 quantification of scaling effects by evaluating tests in different scales
 development of new as well as improvement of existing condensation models to increase
   accuracy of thermal hydraulic modelling for water hammer calculations
 development of a new 1D two-phase flow code for water hammer and shock wave
   transients in piping networks
 validation of thermal hydraulic models including the new computer code for
   condensation-induced water hammers and shock waves in two-phase flows
 qualification of 1D and 3D computational tools for the analysis of the structural response
   including fluid-structure interaction and validation of complex response models.
The results will enable a better understanding and an improved modelling of water hammer
and shock waves with respect to the dynamic pressures, the resulting fluid forces, and finally
provide loads and stresses to be expected. The project will provide validated models and
tools for considering structural response due to transient fluid loadings.
The project consists of four work packages.
WP 1 deals with experiments. Reference data will be obtained at three different test facilities.
Tests to characterise water hammers, shock waves and the resulting loads in relevant piping
configurations with condensation effects will be performed at two of these test facilities
(Pilot Plant Pipework (PPP) and PMK-2). Together with additional data from the 1/1 scale
UPTF facility, the process of condensation controlled water hammer will be studied in three
different scales up to the plant scale. Additionally in the PPP facility, tests involving rapid
valve closures and break openings leading to pressure waves in single phase and two phase
flow will be performed. At the third test facility (Cold Water Hammer Test Facility),
pressure waves typical for water hammers will be generated and the resulting 3D stress fields
in a component wall of difficult geometry (bend) will be measured.
WP 2 deals with thermal hydraulic modelling. It is necessary to develop a new code (WAHA
code) to examine the influence of the numerical methods on the water hammer prediction.
This work will be based on a 6-equation, 1D, two-fluid model for transient non-
homogeneous, non-equilibrium two-phase flow. Development will be made in the following
areas: flow regime maps for fast transients, non-equilibrium condensation model (also for
introduction into existing codes), advanced numerical methods for hyperbolic conservation
laws in order to reduce the numerical diffusion effects, separate integration scheme for the
accurate integration of the stiff source terms, calculation of hydraulic forces on pipes
following the ANS-58.2-1988 standard.
In WP 3, the experimental reference data from WP 1 will be analysed mainly by the
industrial partners. This includes thermal hydraulics as well as structural response. The code
validation will follow the ANS-10.4-1987 Standard.




                                              31
In WP 4, because of the large number of partners involved, coordination of the project will
require a major effort, and is considered as a separate work package. It covers project
management, quality assurance (preparation of a QA manual, and QA audits), and
documentation.




                                            32
Nuclear Energy Programme                                                Plant life extension and
                                                                        management
Operational safety of existing installations                            Integrity of equipment and
                                                                        structures

Title:
FLUID MIXING AND FLOW DISTRIBUTION IN THE REACTOR CIRCUIT

Acronym                  FLOMIX-R

Proposal number FIS5-2001-00119                                Contract number FIKS-CT2001-00197

Type of action           Shared cost                           Duration                     36 months

Starting date            1 October 2001                        EC project officer G. Van Goethem

Total budget             1.256.915 € *                         EC contribution              703.788 € *


Co-ordinator
         Organisation              Forschungszentrum Rossendorf E.V. (FZR)
                                   Institute of Safety Research
         Address                   PO Box 511019
                                   D-01314 Dresden
         Contact person            Prof. Frank-Peter Weiss
                                   Tel:      (49-351)2603480
                                   Fax:      (49-351)2603440
                                   Email     F.P.Weiss@fz-rossendorf.de

Partnership
          Country                       Organisations

        S               Vattenfall Utveckling AB
        UK              Serco Assurance
        D               Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) GmbH
        FIN             Fortum Nuclear Services Ltd
        CH              Paul Scherrer Institut (PSI)
        SK              VUJE Trnava Inc. (*)
        CZ              Ustav Jaderného Vyzkumu Rež a.s.(*)
        HU              KFKI (*)
        HU              Paks Nuclear Power Plant (*)
__________________________________
* Budget expanded following inclusion of new partners from Newly Associated States (NAS).


Project Summary
The project aims at performing a well-defined set of mixing experiments that are supported
with CFD calculations. The experiments will help to improve the basic understanding of the


                                                         33
effect of scale and structures and to provide data for CFD code validation. Emphasis will be
put on covering slug mixing phenomena relevant for local boron dilution scenarios, and
mixing phenomena of interest for operational issues and thermal fatigue. The participants
will carry out a co-ordinated effort of experimental and computational work. The steady state
and transient mixing experiments and flow distribution measurements will be performed with
1:5 scaled facilities. Improved measurement techniques are capable of providing data on
turbulent mixing phenomena with enhanced resolution in time and space.

The investigations are aimed at studying the mixing with operating pumps and during the re-
start of flow circulation in the primary system being relevant for operational problems and
life time management as well as for steam line break and local boron dilution issues. The
experiments will be carried out to complete the results from earlier programmes by
employing improved measurement techniques being capable of providing detailed data on
turbulence intensities, chaotic swirl behaviour and transient velocity fields.

Calculations will be done by a number of participants to justify application ranges of various
turbulence models, to govern numerical diffusion, to account for grid and time step effects
and possible user effects and to interpret the obtained data. Performing benchmark
calculations for a set of selected experiments with two different CFD codes, the applicability
of various turbulence modelling techniques will be studied for various transient and steady
state flow. Different approaches to model the flow in and around geometrically complicated
internal structures (e.g. sieve plates) will be assessed. Weak points of the models and not yet
fully understood physical phenomena will be identified. The key phenomena will be
surveyed including applications to the plant life management purposes. The flow distribution
data available from the commissioning tests (Sizewell-B for PWRs and Loviisa for VVERs)
will be used together with the data from the 1:5 scaled facilities as a basis for the flow
distribution studies.

The expected results are:
 the experimental fluid mixing data sets from the 1:5 scaled experiments with enhanced
   resolution in time and space,
 conclusions on flow distribution and temperature fluctuations in NPPs under normal
   operation conditions being important for economical operation and the estimation of
   thermal fatigue,
 the recommendations for the CFD applications concerning applied turbulence modelling
   features, numerical diffusion, grid, time step and user effects.




                                              34
Nuclear Energy Programme                                  Plant life extension and
                                                          management
Operational safety of existing installations - RI         Integrity of equipment and
                                                          structures

Title:
FUTURE EU NEEDS IN MATERIALS RESEARCH REACTORS

Acronym                 FEUNMARR

Proposal number FIS5-2001-00007                     Contract number FIR1-CT2001-20122

Type of action          Thematic network            Duration           12 months

Starting date           1 November 2001             EC project officer A. Zurita

Total budget            119.950 €                   EC contribution    119.950 €


Co-ordinator
         Organisation         Commissariat à l'Energie Atomique (CEA)
                              Nuclear Energy Direction
         Address              C.N. Cadarache BT.224
                              F-13108 Saint-Paul-les-Durance
         Contact person       Mr. Daniel Parrat
                              Tel:     (33-4)42257572
                              Fax:     (33-4)42254777
                              Email    daniel.parrat@cea.fr

Partnership
          Country                   Organisations

            B                  Belgian Nuclear Research Centre (SCK-CEN)
            INT                European Commission - JRC/IE
            UK                 NAC International
            CZ                 Nuclear Research Institute Rež plc (NRI)
            D                  Framatome ANP GmbH
            F                  Technicatom
            UK                 Independent Consultant
            F                  OECD - Nuclear Energy Agency
            F                  Commissariat à l'Energie Atomique (CEA)

Project Summary
Most of European Material Test Reactors will be more than 40 years old by 2010. This
Thematic network is addressing an increasingly urgent problem to have shortly in Europe
many obsolete research reactors from the safety, economy or performance viewpoints and



                                               35
therefore to have a high probability that most of the MTRs will be closed down by the end of
the decade and therefore not to have available the necessary tools able to provide the
knowledge needed by the industry, safety, research or policy makers. The objectives of this
TN are to determine what will be the future European irradiation needs in MTRs. Objectives
will be achieved by gathering literature and contributions from experts, synthesise the
material and write a report giving qualitative and possibly quantitative assessment of
irradiation needs.

Contribution to programme objective will be made in the following chapters: Plant life
management and extension (PLEM), integrity of equipments and structures, Study of
abnormal and accidental events, Study of high burn-up UO2 and MOX, Study of
transmutation (transmutation of actinides and LLFP), Study of corrosion, erosion, hydriding,
or deposition of corrosion products on out of core heat transfer surfaces in order to limit or
decrease the irradiation dose and decrease the amount of wastes. Medical and neutron beams
will be also considered.

Two kinds of participants will be involved in this Thematic network:
   A group of specialists (around 45 people cost free) from the areas of research covered
      in this TN named group 1. They will be invited to two workshops.
   A group of consultants (9 consultants including 3 cost free experts) invited to give
      advice on the management of the TN and help to write documents and reports. They
      will be invited to workshops and consultant meetings

The TN will include various tasks:
T0: CONTRACT NOTIFICATION
T1: Consultant meeting 1 at T0. Subject: organise the workshop 1 (participants, areas of
expertise, format and content of presentations, subject of panel discussions, list key issues
and questions.
T2: Workshop 1 at T0+1month, 3 days, subject: Give papers in areas of expertise (45 papers)
panel discussion on the main issues and synthesis including subject of research, type of
experiment, meaningful parameters, non-destructive examinations, destructive examination
T3: Consultant meeting 2 at T0+3months, two full days, make a synthesis report of
workshop1 and conclude on: areas of research, type of experiments, parameters measured,
non-destructive examinations, destructive examination….)
T4: Workshop 2 and Consultant meeting 3 at T0+10 months, 3 days, Read and discuss the
synthesis paper provided by T3 and make the final corrections/modifications. Draft of the
final document.

The main milestones and expected results are as follows:
 Gather literature and contributions on irradiation needs in MTRs
 Synthesise the material gathered and propose areas where research should be pursued and
   in each area a list of detailed experiments to be carried out.
 Issue a document on qualitative and possibly quantitative irradiation needs




                                             36
Nuclear Energy Programme                                    Plant life extension and
                                                            management
Operational safety of existing installations                Integrity of equipment and
                                                            structures

Title:
MODELLING OF AGEING IN CONCRETE NUCLEAR POWER PLANT STRUCTURES

Acronym                 MAECENAS

Proposal number FIS5-2001-00100                       Contract number FIKS-CT2001-00186

Type of action          Shared cost                   Duration             36 months

Starting date           1 November 2001               EC project officer G. Van Goethem

Total budget            1.160.424 €                   EC contribution      1.099.257 €


Co-ordinator
         Organisation          University of Sheffield
                               Dept. of Civil and Structural Engineering
         Address               Sir Frederick Mappin Building, Mappin Street
                               UK-S1 3JD Sheffield
         Contact person        Dr. Roger Crouch
                               Tel:      (44-144)2225716
                               Fax:      (44-144)2225700
                               Email     r.crouch@sheffield.ac.uk

Partnership
          Country                     Organisations

            CZ                  Ceské Vysoké Uceni Techniké v Praze (Technical University)
            UK                  University of Glasgow
            UK                  Health and Safety Executive (HSE)
            F                   Ecole Centrale de Nantes
            UK                  British Energy Generation Ltd
            I                   University of Rome "La Sapienza"
            I                   International Centre for Mechanical Sciences

Project Summary
The MAECENAS project will create an advanced engineering analysis tool which will allow
the structural integrity for aged, reinforced, pre-stressed concrete NPP structures to be
assessed in a meaningful, scientifically rational manner.




                                                 37
To achieve this objective, the project will involve the design and undertaking of novel
laboratory tests to detect the behaviour of plain concrete under simulated ageing conditions.
Using data from these experiments, together with existing laboratory data, a generalised
thermo-mechanical constitutive model for concrete is to be constructed. This model will
simulate the time-dependent deformation response under arbitrary multi-axial stress states at
temperature excursions up to 600 Celsius. This formulation will be embedded within a fully
coupled, multi-phase hygro-thermo-mechanical theoretical framework which is able to
describe the interaction between temperature dependent moisture movement and material
damage in concrete. The resulting system of equations will be solved using an object-
oriented Finite Element code. The latter will be developed during the project taking
advantage of multi-processor computing environments to speed-up analysis run-time for
complex 3D problems.

Representative pre-stressed concrete containment vessels (PCCVs) and pre-stressed concrete
pressure vessels (PCPVs) will be identified as part of this project and the ageing processes
simulated using the newly developed FE code. Comparisons between predicted states and
measured states will be made before undertaking a series of safety-margin FE analyses.
These analyses will involve simulating pre-defined severe accidents to detect any change in
the vessel safety margin.

Finally, the results from these simulations (together with information on the statistical
variation of all key parameters) will be used to construct a simplified (reliability-based)
safety-cost analysis procedure to enable engineers to determine appropriate repair or
strengthening strategies as part of the overall structural management process.




                                             38
Nuclear Energy Programme                                      Plant life extension and
                                                              management
Operational safety of existing installations                  Integrity of equipment and
                                                              structures

Title:
CONCRETE CONTAINMENT MANAGEMENT USING THE FINITE ELEMENT
TECHNIQUE COMBINED WITH IN-SITU NON-DESTRUCTIVE TESTING OF
CONFORMITY WITH RESPECT TO DESIGN AND CONSTRUCTION QUALITY

Acronym                 CONMOD

Proposal number FIS5-2001-00125                       Contract number FIKS-CT2001-00204

Type of action          Shared cost                   Duration           36 months

Starting date           1 January 2002                EC project officer G. Van Goethem

Total budget            1.334.760 €                   EC contribution    454.000 €


Co-ordinator
         Organisation          Force Institute
                               Division for Materials and Chemical Analysis
         Address               Park Alle 345
                               DK-2605 Broendby
         Contact person        Mr. Oskar Klinghoffer
                               Tel:     (45-43)267255
                               Fax:     (45-43)267011
                               Email    osk@force.dk

Partnership

          Country                     Organisations

            S                   Scanscot Technology AB
            S                   Barsebaeck Kraft AB
            F                   Electricité de France (EDF)

Project Summary
Safety-related concrete structures such as concrete containments at nuclear power plants are
subject to ageing processes, which can reduce their safety as well as functional lifetime.
Serious problems have also been known to occur to these structures due to defects caused at
the construction stage. In order to establish the actual status of concrete containments it is
necessary to apply investigative techniques that are capable of providing information about
the internal structure and condition of the concrete and reinforcing.




                                                 39
By using the Finite Element (FE) method it is possible to study the behaviour of concrete
containments under various loading conditions and thus to identify critical sections. This
information can then be used to plan non-destructive testing in order to obtain a more
accurate description of the true nature of the structure at these points. Subsequently this new
information can be used as input to FE-models to allow more realistic behavioural
predictions.
The FE-model can then be modified with time by repeating the investigative process at
intervals. This process includes the use of Non Destructive Testing (NDT) to monitor
changes in the chemical and physical properties of the concrete and if applicable to quantify
deterioration processes. The condition assessment and ageing management of concrete
containments will thus be based on realistic evaluations.
This project proposes the application and exploitation of the mutual benefits of state-of-the-
art NDT technology to concrete containments and combination with the latest FE-modelling
techniques.
The NDT examinations will be carried out by Force Institute (Denmark), mainly at the
Barsebäck NPP (Sweden) but also at the EDF MAEVA model compartment. FE-modelling
will be done for both the containments by Scanscot Technology (Sweden) and EDF (France).
The objective of this project is to develop the application and understanding of NDT
techniques for conformity and condition assessment of concrete containments and to
integrate this with state-of-the-art and developed FE-modelling techniques and FE-analysis
of structural behaviour. This will enable optimisation of maintenance activities and will
ensure safer operation of nuclear plants throughout their planned, and where applicable
extended lifetimes.
Emphasis will be placed on establishing the actual conformity and condition of the structures
by identifying structure-specific features and possible critical defects as well as damage
mechanisms. The information obtained can be used in establishing whether the structures
have higher or lower safety margins compared with original assumptions. The same
principles can then be applied in evaluating the effectiveness of remedial actions such as
repair of critical defects.
Verification of the revised FE structural models by full-scale load testing will provide
increased reliability in safety analysis. Comparison of the FE-results with the original
pressure test at Barsebäck containment B1 (linear behaviour) and pressurisation to collapse
at EDF MAEVA model compartment (linear and non-linear behaviour) will be done. A
programme for a new pressure test at Barsebäck NPP 1 will be specified.
It is envisaged that development of the NDT-techniques, their adaptation and application to
this type of structure will provide a basis for standardisation of testing procedures and to
their wider use and availability in general. The improved knowledge of FE-modelling
techniques regarding reactor containments gained in this project can be used, not only for
ageing management, but also for other important structural analyses and investigations.




                                              40
Nuclear Energy Programme                                    Plant life extension and
                                                            management
Operational safety of existing installations                On-line monitoring and
                                                            maintenance

Title:
EVALUATION OF NON DESTRUCTIVE TESTING TECHNIQUES FOR MONITORING
OF MATERIAL DEGRADATION

Acronym                 GRETE

Proposal number FIS5-1999-00280                       Contract number FIKS-CT2000-00086

Type of action          Shared cost                   Duration          36 months

Starting date           1 October 2000                EC project officer P. Manolatos

Total budget            1.527.358 €                   EC contribution   669.975 €


Co-ordinator
         Organisation          Electricité de France (EDF) - R&D Division
                               Materials Studies Branch
         Address               Route de Sens, Ecuelles B.p. 1
                               F-77818 Moret-sur-Loing Cedex
         Contact person        Mr. Marc Delnondedieu
                               Tel:     (33-1) 60736315
                               Fax:     (33-1) 60736889
                               Email    marc.delnondedieu@edf.fr

Partnership

          Country                     Organisations

            NL                  Nuclear Research and Consultancy Group (NRG)
            FIN                 Technical Research Centre of Finland (VTT)
            E                   Tecnatom S.A.
            D                   Fraunhofer-Gesellschaft zur Förderung der angewandten
                                Forschung e.V. (FhG-IZFP)
            INT                 European Commission - JRC/IE
            UK                  Serco Assurance
            E                   Centro de Investigaciones Energéticas, Medioambientales y
                                Tecnológicas (CIEMAT)
            A                   Österreichisches Forschungszentrum Seibersdorf Ges.m.b.H.
                                (ARCS)
            HU                  KFKI Atomic Energy Research Institute (AEKI)
            D                   University of Hannover
            CH                  Paul Scherrer Institut (PSI)



                                                 41
          CZ                  Nuclear Research Institute Rež plc (NRI)
          D                   Siempelkamp Pruf- und Gutachter - Gesellschaft MBH
          F                   INSAVALOR
          RU                  All-Russian Institute for Nuclear Power Plants Operation
                              (VNIIAES)
          D                   Framatome ANP GmbH

Project Summary
The lifetime extension of ageing power plants for electricity production is an economical
way to reduce the electricity generating costs for the benefit of the customers. Extending the
lifetime of existing installations requires the development of innovative reliable techniques
for the inspection of critical components. Such techniques will detect changes in the
materials and will allow to plan the actions for failure prevention, e.g. change of operation
parameters, increased inspection intervals or replacement of components.

The main objective of this project is to assess the capability and the reliability of innovative
inspection techniques by means of a round robin exercise. Aged samples will be tested by the
partners using various techniques (ultrasonics, magnetics, thermoelectricity and dynamic
indent). The non-destructive techniques that will be tested are different from standard
inspection methods. The aim of standard techniques is to detect macroscopic defects like
cracks, including for certain applications sizing and imaging. The methods applied in this
project are sensitive to any microstructural change in the material leading to a degradation of
the mechanical properties of the component long before macroscopic cracks are initiated and
eventually grow. However, these indirect methods require a careful interpretation of the
signal measured in terms of microstructural evolutions due to ageing in the material.

Two ageing mechanisms were chosen: one is the neutron irradiation damage occurring in
reactor pressure vessels made of ferritic steels and the other is the thermal fatigue affecting
austenitic stainless steel pipings.
1. Irradiation damage : Samples irradiated in nuclear reactors will be provided by some of
   the partners as well as results/data already available on these materials. The evaluation of
   the non-destructive techniques will be performed in hot cells by different NDT teams. The
   results of the testing will be gathered and interpreted in terms of microstructural and
   mechanical changes.
2. Thermal fatigue damage : Samples will be tested in low cycle fatigue and in thermocyclic
   fatigue conditions. The microstructural changes related to fatigue damage (dislocation
   network and martensitic phase) will be observed using Transmission Electron
   Microscopy, Neutron Diffraction and advanced X-Ray Diffraction methods. The
   evaluation of the non-destructive techniques will be performed in each laboratory
   participating to the testing. The results of the testing will be gathered and interpreted in
   terms of microstructural and mechanical changes.




                                              42
Nuclear Energy Programme                                    Plant life extension and
                                                            management
Operational safety of existing installations                On-line monitoring and
                                                            maintenance

Title:
DEVELOPMENT OF LIGHT WATER REACTOR (LWR) REFERENCE ELECTRODES

Acronym                 LIRES

Proposal number FIS5-1999-00113                       Contract number FIKS-CT2000-00012

Type of action          Shared cost                   Duration          48 months

Starting date           1 October 2000                EC project officer P. Manolatos

Total budget            1.256.416 €                   EC contribution   649.305 €


Co-ordinator
         Organisation          Belgian Nuclear Research Centre (SCK-CEN)
                               Reactor Materials Research
         Address               Boeretang 200
                               B-2400 Mol
         Contact person        Dr. Rik-Wouter Bosch
                               Tel:     (32-14) 333428
                               Fax:     (32-14) 321336
                               Email    rbosch@sckcen.be

Partnership
          Country                     Organisations

            HU                  KFKI Atomic Energy Research Institute (AEKI)
            F                   Commissariat à l'Energie Atomique (CEA)
            E                   Centro de Investigaciones Energéticas, Medioambientales y
                                Tecnológicas (CIEMAT)
            B                   Katholieke Universiteit Leuven (KUL)
            CZ                  Nuclear Research Institute Rež plc (NRI)
            D                   Framatome ANP GmbH
            S                   Studsvik Material AB
            FIN                 Technical Research Centre of Finland (VTT)

Project Summary
The main objective of the LIRES project is to develop reference electrodes, that are robust
enough for use inside a Light Water Reactor. The developed reference electrodes must
survive in harsh LWR conditions, i.e. high temperature, high pressure and irradiation. The



                                                 43
development of such a reference electrode is important for monitoring the corrosion
performance of stainless steel core components, which accumulate extensive irradiation
damage over time and hence are susceptible to IASCC. The corrosion potential, measured
against a reference electrode, allows to distinguish between situations where IASCC is likely
to occur (high value of the corrosion potential) or not (low value of the corrosion potential).
Distinction is made between reference electrodes for a BWR and a PWR as water chemistry
and operating temperatures are different. Also much work has been done on BWR reference
electrodes, while little or no attention has been given to the development of a PWR reference
electrode. Therefore different trajectories for the BWR and PWR electrode will be followed.
Four main work-packages are foreseen:
(1) Two testing standards are to be written, based on an evaluation of existing high
temperature reference electrodes and testing methods to prove their reliability. The first
standard is to describe testing in a laboratory under high temperature and high pressure
conditions. The second standard is to describe a test procedure for testing in a Material Test
Reactor.
(2) Design and development of high temperature reference electrodes for PWR-conditions
(operating temperature up to 350°C) at laboratory scale. Four different designs are
investigated by four different laboratories, each originating from a typical category of
reference electrodes.
(3) Round robin test among the participating laboratories of the just developed reference
electrodes, using the test procedure developed under WP 1. Based on the Round Robin
results, the best reference electrode will be selected and used for the irradiation experiment.
(4) Testing of high temperature reference electrodes under appropriate irradiation conditions
in a Material Test Reactor. The PWR reference electrode is selected from the Round Robin.
The BWR reference electrode is selected based on existing knowledge in the consortium and
a recently finished international scientific program (WACOL) on high temperature (BWR)
reference electrodes. The irradiation experiments are hanging-on experiments, i.e. they are
combined with other irradiation experiments to reduce the costs.
Work package 1 will deliver two testing standards for High Temperature Reference
Electrodes (HTRE); one for use in the laboratory and one for use in a MTR. Work package 2
will deliver four prototype HTREs. Work package 3 will deliver a laboratory performance
appraisal for all four HTREs. Work package 4 will deliver a performance appraisal of
HTREs under BWR and PWR conditions. The final deliverable is one (in-core) HTRE for
use in a BWR and one for use in a PWR conditions.




                                              44
Nuclear Energy Programme                                    Plant life extension and
                                                            management
Operational safety of existing installations                On-line monitoring and
                                                            maintenance

Title:
SIGNAL PROCESSING AND IMPROVED QUALIFICATION FOR NON-DESTRUCTIVE
TESTING OF AGEING REACTORS

Acronym                 SPIQNAR

Proposal number FIS5-1999-00233/251                   Contract number FIKS-CT2000-00065

Type of action          Shared cost                   Duration          36 months

Starting date           1 October 2000                EC project officer P. Manolatos

Total budget            1.746.368 €                   EC contribution   999.995 €


Co-ordinator
         Organisation          Mitsui Babcock
                               Technology Centre
         Address               High Street
                               UK-PA4 8UW Renfrew
         Contact person        Mr. Neil Cameron
                               Tel:      +44 141 886 4141
                               Fax:      +44 141 885 3338
                               Email     ncameron@mitsuibabcock.com

Partnership

          Country                     Organisations

            UK                  British Energy Generation Ltd
            B                   AIB - VINCOTTE International
            S                   Uppsala University
            D                   Universität Gesamthochschule Kassel
            F                   CEA - Centre d'Etudes et de Recherches sur les Matériaux
                                (CEREM)
            INT                 European Commission - JRC/IE
            F                   Intercontrôle
            CZ                  Nuclear Reseach Institute Rež plc (NRI)
            E                   Tecnatom S.A.
            D                   E.ON Kernkraft GmbH
            S                   SQC Swedish NDT Qualification Centre
            UK                  Serco Assurance




                                                 45
Project Summary
Ultrasonic inspection plays an important role in assuring the safe and economic operation of
nuclear plant. The overall objectives of this project are:

   to improve the performance of ultrasonic inspection for the detection and sizing of cracks
    in important austenitic stainless steel components (which are among the most difficult to
    inspect)

   to improve confidence in the way in which ultrasonic inspection procedures are qualified
    (demonstration that performance matches requirements) by improving test piece trials

The first overall objective will be achieved by developing and assessing signal processing
methods designed to improve performance. The second objective will be achieved by
determining and comparing the ultrasonic responses of real and synthetic stress corrosion
and fatigue cracks, to provide guidance on the extent to which synthetic or “virtual” defects
can be used in test piece trials, instead of real defects.

The work involves measuring the ultrasonic response from real and synthetic defects, mainly
in austenitic specimens and testpieces. The synthetic defects will include “realistic” defects
intended to simulate the complex morphology of real defects, and also “artificial” defects
which are simple in shape but easier to insert and more reproducible. Comparison of the
responses will determine which aspects, if any, can be replicated using synthetic defects.
The feasibility of using “virtual defects” will be investigated, whereby measured signals
from real defects are injected into the ultrasonic equipment in such a way that the effect to
the inspector is identical to what would have occurred had a real defect been present. The
European Network on Inspection qualification (ENIQ) will be a play a major role in
dissemination of these results.

Ultrasonic data from the austenitic specimens and testpieces will also be provided to signal
processing specialists who will develop signal processing methods aimed at overcoming
current problems in detecting and sizing cracks in austenitic welds. There needs to be an
interface between the signal processing methods developed and the ultrasonic inspection
systems which will apply them. A software tool will therefore be produced to read ultrasonic
data files, display the resultant images in a common format and apply the signal processing
methods to the images. Final practical trials on defective specimens and testpieces will be
performed to compare performance with and without using the signal processing methods.




                                             46
Nuclear Energy Programme                                    Plant life extension and
                                                            management
Operational safety of existing installations                On-line monitoring and
                                                            maintenance

Title:
REACTOR DOSIMETRY: ACCURATE DETERMINATION AND BENCHMARKING OF
RADIATION FIELD PARAMETERS, RELEVANT FOR REACTOR PRESSURE VESSEL
MONITORING

Acronym                 REDOS

Proposal number FIS5-2001-00004                       Contract number FIKS-CT2001-00120

Type of action          Shared cost                   Duration           36 months

Starting date           1 November 2001               EC project officer S. Casalta

Total budget            916.919 €                     EC contribution    499.949 €


Co-ordinator
         Organisation          Tecnatom S.A.
                               Inspection Engineering Division
         Address               Avda. Montes de Oca, 1
                               E-28709 San Sebastian de los Reyes (Madrid)
         Contact person        Mr. Antonio Ballesteros
                               Tel:     (34-91)6598723
                               Fax:     (34-91)6598677
                               Email    aballesteros@tecnatom.es

Partnership

          Country                     Organisations

            D                   Forschungszentrum Rossendorf e.V. (FZR)
            CZ                  Nuclear Research Institute (NRI)
            HU                  KFKI Atomic Energy Research Institute (AEKI)
            BG                  Institute of Nuclear Research and Nuclear Energy (INRNE)
            INT                 European Commission - JRC/IE
            CZ                  Skoda JS a.s.
            D                   Framatome ANP GmbH

Project Summary
The radiation embrittlement of RPVs has become one of crucial consideration for safe
operation of ageing nuclear power plants. The qualification of measuring and calculational
methodology for the determination of neutron and gamma exposures in critical locations of
RPV will be done in the REDOS project via a corresponding benchmark, using data obtained


                                                 47
in the LR-O facility. This activity will be carried out with experimental measurements in a
VVER-1000 mock-up. The space-energy indices, dpa and other values will be derived from
the spectra or evaluated from direct measurements.

Particular objectives of the REDOS project are:

I. Improvement of the RPV monitoring
II. Improvement of the neutron-gamma calculation methodologies through the LR-0
     engineering benchmark
III. Accurate determination of radiation field parameters in the vicinity and over the
     thickness of the RPV.

The project will focus on VVER reactor type, but the results will be also of interest for
western PWRs.

The project is divided in four work packages as describe below:

Work-package 1: Review of available experimental data.
Work-package 2: Experimental programme in VVER-1000 Mock-up (engineering
                benchmark).
Work-package 3: Analytical area. Analysis of calculated and measured data, conclusions.
Work package 4: Radiation field parameters in the vicinity of and over the thickness of the
                reactor pressure vessel.

The project will start up with a review of existing data, namely with the VVER-440 and
VVER-1000 engineering benchmark experimental data and NPP data relevant for attenuation
coefficients through the vessel wall. The experimental activity (WP-2) will be concentrated
on gamma-ray spectra measurement and extended neutron spectra measurements in a mock-
up in the LR-0 reactor to create a 3D (three dimensional) benchmark in the vicinity of a RPV
simulator (VVER-1000 engineering benchmark). The analysis of the data obtained (WP-1,
WP-2) will be carried out jointly by the project partners (WP-3). The results of WP-4 should
provide accurate information on the neutron-gamma exposure parameters through the
thickness of the RPV, where important changes in neutron-gamma spectrum are present.




                                            48
Nuclear Energy Programme                                    Plant life extension and
                                                            management
Operational safety of existing installations                On-line monitoring and
                                                            maintenance

Title:
VIRTUAL REALITY FOR INSPECTION, MAINTENANCE, OPERATION, AND REPAIR
OF NUCLEAR POWER PLANT

Acronym                 VRIMOR

Proposal number FIS5-1999-00328                       Contract number FIKS-CT2000-00114

Type of action          Shared cost                   Duration           24 months

Starting date           1 February 2001               EC project officer S. Casalta

Total budget            1.198.947 €                   EC contribution    599.474 €


Co-ordinator
         Organisation          National Nuclear Corporation Limited (NNC)
                               Simulation Business Team
         Address               Booths Hall, Chelford Road
                               UK-WA16 8QZ Knutsford, Cheshire
         Contact person        Dr. David Lee
                               Tel:      (44-156)5843732
                               Fax:      (44-156)5843441
                               Email     david.lee@nnc.co.uk

Partnership

          Country                     Organisations

            E                   Tecnatom
            E                   Universidad Politécnica de Madrid (UPM)
            E                   Centro de Investigaciones Energéticas, Medioambientales y
                                Tecnológicas (CIEMAT)
            B                   Centre d'Etude de l'Energie Nucléaire (SCK-CEN)
            UK                  Z+F (UK) Ltd

Project Summary
The objective of VRIMOR is to develop a methodology and prove the viability of
minimising occupational exposure, reducing safety risks, and minimising costs associated
with manual maintenance and other activities on operational nuclear power plant (NPP). This
will be achieved through the development of computer simulation tools and interfaces
combined with the enhancement of laser and radiological scanning techniques to be applied



                                                 49
cost effectively to the planning, training and assessment of maintenance tasks. The aims are
to develop a suite of interchangeable technologies in response to nuclear plant operatives’
needs, to evaluate its performance in a practical application, and to provide recommendations
for the future development and adoption of the tools.

Sophisticated human computer simulation models will be used as the basis for developing
intuitive user interfaces that will allow the plant operatives to use these complex tools. Two
development streams are proposed; a graphical interface complemented with voice control;
and a hardware (joystick) interface complemented with stereo vision. These will be
developed on differing commercial software systems and will be used by operators to
evaluate optimum methods of access and working.

The method of working can only be considered optimal if due consideration has been given
to occupational exposure. It is therefore planned to develop automated methods for
calculation of human dose uptake based on the human simulation. This will again follow two
parallel tracks; one where the dose to key body parts is computed as the simulation
progresses (real-time), and the other where a trajectory is computed from the simulation and
doses calculated off-line.

The input requirements for the dose calculations will be in the form of a dose rate field
which will be generated in one of two ways; from the development of a radiological
scanning system and from the development of a computational tool that uses conventional
source and activity data derived for the plant.

The geometry of the plant area will be provided using laser-scanning technology which will
be developed to provide a 3D model incorporating the radiological data which can be read by
commercially available human simulation packages.

The developed technologies will be interface tested and station operators trained in their use
in order to assess their performance and benefits. The applications will be reported, human
factors assessed, technologies compared, and recommendations provided for future
development and uptake.




                                             50
Nuclear Energy Programme                                    Plant life extension and
                                                            management
Operational safety of existing installations                On-line monitoring and
                                                            maintenance

Title:
NUCLEAR RISK-BASED INSPECTION METHODOLOGY

Acronym                 NURBIM

Proposal number FIS5-2001-00082                       Contract number FIKS-CT2001-00172

Type of action          Shared cost                   Duration           32 months

Starting date           1 November 2001               EC project officer S. Casalta

Total budget            1.209.094 €                   EC contribution    604.546 €


Co-ordinator
         Organisation          Gesellschaft für Anlagen - und Reaktorsicherheit (GRS) GmbH
         Address               Schwertnergasse 1
                               D-50667 Köln
         Contact person        Dr. Helmut Schulz
                               Tel:     (49-221)2068603
                               Fax:     (49-221)2068888
                               Email    suh@grs.de

Partnership

          Country                     Organisations

            F                   Electricité de France (EDF)
            S                   OKG AB
            D                   E. ON Kernkraft GmbH
            S                   Det Norske Veritas AB
            CZ                  Nuclear Research Institute Rež plc (NRI)
            UK                  OJV Consultancy Ltd
            UK                  The Welding Institute (TWI) Ltd
            FIN                 Technical Research Centre of Finland (VTT)
            UK                  Mitsui Babcock Energy Ltd
            INT                 European Commission - JRC/IE
            E                   TECNATOM S.A.

Project Summary
Inspection and maintenance of nuclear power plants (NPPs) is a prerequisit for safe
operation but represents a significant burden for plant operators in Europe. If the European



                                                 51
nuclear industry is to remain competitive and maximise its contribution to the reduction of
global warming, then more focussed inspection and maintenance schedules are needed that
will reduce costs and outage times, while maintaining or increasing plant safety. The
conclusions of EURIS1 was that this could be best achieved through a ‘Risk Based
Management Philosophy’. The objective of the proposed project NURBIM (Nuclear Risk
Based Inspection Methodology) is to progress the recommendations of EURIS and
subsequent work of ENIQ TG42 to develop improved procedures to identify where the
highest likelihood of damage/failure is located in passive systems, structures and
components. To then provide quantitative measures of the associated risk. Within this
context, risk is defined in terms of a consequences and the probability of incurring those
consequences. Such a risk-based approach would, through the focusing of resources, lead to
increased safety, reliability and availability of the overall plant. The NURBIM project will
focus on the definition of best practice methodologies for performing risk-based analyses 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.
The particular focus of NURBIM corresponds to the following needs highlighted earlier in
2000 in a discussion document produced for DG RTD by the EURIS Concerted Action
Group:
 Development of structural reliability models (SRMs) to help in assessing the probability
    of failure of passive components subject to specific in-service degradation mechanisms.
 Interpretation of existing plant-specific probability safety assessments (PSAs) for
  assessing passive components.
 Providing a reference to be used in the development of a future european standard to
  risk-based inspection methodology.
To reach the goals of NURBIM the following steps are intended. A compilation of a data
base of actual and potential damage mechanisms. Establishing criteria to be met by SRM’s.
Selection of reference procedures to estimate risk of component failure. Assessment, review
and comparison of methods to estimate component failure frequencies. Establishing an
interface that is tailored to the needs of a PSA’s. An investigation of present procedures to
identify risk significant locations. Investigation of the relationship between the capability of
the inspection and the risk-based management. The developed methodology will be applied
in a practical case of primary components on the Oskarshamn BWR.
The consortium is formed of utilities operating PWR and BWR nuclear power plants
representing half of the nuclear generating capacity within Europe and technical support
departments and organisations with a strong background in structural integrity issues, risk
assessment, inspection of nuclear power plant components and evaluation of operating
experience. Therefore, the consortium represents all necessary facets for such a
multidisciplinary task as being proposed in NURBIM.
The final result of the project will be a handbook giving guidance for methods and
approaches to be used for risk-based inspection. It is the intention to make the handbook
public available.




1
  EURIS (European Network of Risk-informed In-service Inspection) a Euratom Research Framework Programme 1994-1998
“Nuclear Fission SafetY”
2
  ENIQ Report ERN 19742 EN



                                                        52
Nuclear Energy Programme                                    Plant life extension and
                                                            management
Operational safety of existing installations                On-line monitoring and
                                                            maintenance

Title:
MANAGEMENT OF NUCLEAR PLANT OPERATION BY OPTIMISING WELD REPAIRS

Acronym                 ENPOWER

Proposal number FIS5-2001-00071                       Contract number FIKS-CT2001-00167

Type of action          Shared cost                   Duration          36 months

Starting date           1 December 2001               EC project officer P. Manolatos

Total budget            1.713.251 €                   EC contribution   919.263 €


Co-ordinator
         Organisation          Institut de Soudure
         Address               90, rue des Vanesses
                               F-93420 Villepinte
         Contact person        Dr. Christian Boucher
                               Tel:       (33-1)49903633
                               Fax:       (33-1)49903628
                               Email      c.boucher@institutdesoudure.com

Partnership

          Country                     Organisations

            UK                  British Energy Generation Ltd
            UK                  Mitsui Babcock Energy Ltd
            D                   Framatome ANP GmbH
            INT                 European Commission - JRC/IE
            UK                  University of Bristol
            F                   Usinor Industeel SA

Project Summary
The project duration is 36 months. The project is divided into 8 work packages, 7 of which
adress the technical objectives described above. Three dimensional finite element weld
simulation approaches will be developed for simple geometries early in the project. Results
from this work will form a basis for weld procedure optimisation and new stress relief
treatments. The weld optimisation studies will examine the influences of groove geometry,
welding sequence, weld parameters and pre-heat. The optimised procedures will be
demonstrated for 3 nuclear applications. The basic principles of novel thermo mechanical



                                                 53
stress relief treatments will be examined using numerical models, followed by the
development of general procedures for nuclear components. An important aspect of the
project is to improve the understanding of how weld residual stresses and post weld
treatments influence the integrity of aged components. Advanced numerical modelling
methods will be used to study the interactions between residual stresses, post weld
treatments, operational loads, crack growth and fracture. The fracture results will be
interpreted and used to develop and underpin new advice in defect assessment procedures
and standards for dealing with residual stress. The numerical modelling will be supported by
a programme of experimental work entailing the manufacture of mock ups for weld repair
and alternative post weld treatments optimisation trials, material property tests, residual
stress/strain measurements using neutron diffraction, deep hole drilling, the ring core method
and laser strain scanning, and fracture mechanics tests. Ferritic and austenitic stainless steel
components and a low alloy ferritic plate clad with stainless steel will be examined in the
programme. The final 6 months of the project will focus on interpreting the technical results
and producing sets of guidelines for optimised weld repairs, alternative post weld treatments
and on the treatment of residual stresses in fracture assessment, with a view to their
incorporation Codes, Standards and Procedures.




                                              54
Nuclear Energy Programme                                     Plant life extension and
                                                             management
Operational safety of existing installations                 Organisation and
                                                             Management of safety

Title:
BENCHMARK EXERCISE ON SAFETY EVALUATION OF COMPUTER-BASED
SYSTEMS

Acronym                 BE-SECBS

Proposal number FIS5-1999-00216                       Contract number FIKS-CT2000-00054

Type of action          Shared cost                   Duration            30 months

Starting date           1 January 2001                EC project officer G. Van Goethem

Total budget            791.785 €                     EC contribution     395.892 €


Co-ordinator
         Organisation          European Commission
                               JRC/IE, Petten
         Address               Postbus 2
                               NL-1755 ZG Petten
         Contact person        Dr. Christian Kirchsteiger
                               Tel:       (31-224) 565118
                               Fax:       (31-224) 565641
                               Email      christian.kirchsteiger@jrc.nl

Partnership

          Country                     Organisations

            FIN                 Radiation and Nuclear Safety Authority (STUK)
            FIN                 Technical Research Centre of Finland (VTT)
            D                   Inst. für Sicherheitstechnologie (ISTec) GmbH
            F                   Institut de Radioprotection et de Sûreté Nucléaire (IRSN)
            D                   Framatome ANP GmbH

Project Summary
The evaluation of the reliability and safety of the computer-based system embedded in a
nuclear power plant represents more and more a crucial part of the overall assessment of a
nuclear installation. Improving such kind of activity is an important goal to achieve in order
to enhance safety and reliability of nuclear installations. However, although techniques for
hardware assessment are now rather consolidated, the assessment of software reliability and
safety in not a resolved issue.



                                                 55
The project objectives are thus mainly concerned with the development of an international
Benchmark Exercise for a comparative evaluation of existing methodologies in use in the
nuclear field among EU regulators and technical support organisations, tackling the problem
of assessing safety-critical computer-based systems, with particular attention to the software
part.

In this project, Framatome ANP GmbH, the industrial partner of the consortium, will provide
a reference case study. To this purpose, a hypothetical reactor protection system will be
taken into account. Framatome ANP GmbH possesses the competence and know-how to
provide background information of the reference reactor specifying typical Instrumentation
and Control (I&C) functions important to safety that may be considered. Framatome ANP
GmbH will thus provide the requirements and functional specification of a limited number of
safety functions that will be selected by the project partners. Moreover, the industrial partner
will perform the design and implementation of the selected safety functions and will employ
his proprietary tools for automatic code generation and documentation. The source code and
the documentation concerning all the software lifecycle phases will be made available to the
assessor partners, namely STUK, VTT, ISTec and IRSN, who will be involved in an
independent assessment activity by applying the methodology in use in their organisation.
Framatome ANP GmbH will also support the additional testing activity eventually required
by the assessors. It will be the role of JRC-IE to define a common glossary containing all the
terms and concepts used throughout the project, to design proper metrics to compare the
assessment methodologies proposed and applied by the assessor partners and to actually
perform the comparison between the proposed assessment methodologies.

The main expected results and milestones of the Benchmark Exercise are the following:
 The definition of a common glossary of terms concerned with the assessment of safety-
   critical computer-based systems in nuclear power plants;
 The development of a reference case study;
 The description of independent assessment techniques for safety-critical computer-based
   systems adopted and applied among the project partners;
 The application of the independent assessment techniques to the reference case study;
 The proposal of comparison criteria suitable for the comparison of the independent
   assessment techniques;
 The actual comparison of the assessment techniques to identify, in particular, their
   strengths and weaknesses in order to allow a possible improvement in the field.




                                              56
Nuclear Energy Programme                                    Plant life extension and
                                                            management
Operational safety of existing installations                Organisation and
                                                            Management of safety

Title:
COST EFFECTIVE MODERNISATION OF SYSTEMS IMPORTANT TO SAFETY

Acronym                 CEMSIS

Proposal number FIS5-1999-00355                       Contract number FIKS-CT2000-00109

Type of action          Shared cost                   Duration            36 months

Starting date           1 January 2001                EC project officer G. Van Goethem

Total budget            1.948.547 €                   EC contribution     775.000 €


Co-ordinator
         Organisation          British Energy (Generation) Ltd.
                               Plant Engineering Branch
         Address               Bernett Way
                               UK-GL4 3RS Barnwood, Gloucester
         Contact person        Mr. Paul Tooley
                               Tel:      (44-1552) 653503
                               Fax:      (44-1552) 654897
                               Email     p.tooley@ne.british-energy.com

Partnership
          Country                     Organisations

            B                   AIB Vinçotte Nuclear
            UK                  British Nuclear Fuels plc (BNFL)
            F                   Electricité de France (EDF)
            D                   Framatome ANP GmbH
            S                   Sycon Energikonsult AB
            S                   Lund University
            UK                  Adelard (Safety Consultancy)

Project Summary
There are many nuclear power installations within the EU which require maintenance and
modernisation. These installations contain many I&C systems that are regarded as “systems
important to safety” (SIS), i.e.:
      safety systems: systems in the highest safety class, e.g., a protection system
      safety-related systems: systems in lower safety classes, e.g., a control system.


                                                 57
In the past, SIS were specially developed for the nuclear industry in a particular country.
These systems would often be implemented using simple analogue, relay or discrete logic
technologies that were relatively easy to analyse and justify. In addition SIS tended to be
developed to comply with the requirements of a single national regulatory body. This
situation has changed dramatically, SIS are now becoming heavily reliant on computer-based
systems. The current control system market is subject to increasing globalisation and
competition within the EU. These issues pose considerable additional problems in the
justification and regulatory approval of SIS refurbishments for nuclear plants in Member
States.
The CEMSIS project seeks to:
        maximise safety
        minimise costs
by developing common approaches within the EU to the development and approval of SIS
refurbishments that use modern commercial technology. The specific technical objectives are
to:
       1. develop a safety justification framework for the refurbishment of SIS that is
           acceptable to different stakeholders (licensing bodies, utilities) within the Member
           States
       2. develop methods for establishing the safety requirements for control system
           refurbishment and develop an associated engineering process
       3. develop justification approaches for widely used modern technologies, i.e. - COTS
           (Commercial Off The Shelf) products and graphical specification (logic diagram)
           languages
       4. evaluate these developments on realistic examples taken from actual projects
       5. disseminate the results of our work to plant operators and regulators within the EU.
CEMSIS will take input from regulators on licensing issues and draw on existing experience
of nuclear regulators within the EU on acceptable approaches. This experience will be fed
into our justification framework. CEMSIS will also draw on the experience of a wide range
of “stakeholders” in the industry “operators, I&C suppliers, system integrators and software
specialists to identify acceptable and economic approaches to refurbishment. To focus the
effort the concepts will be applied to at least three industrial case studies (led by BNFL,
Sycon, and EDF). The examples are under review but possibilities include:
        Replacement of PDP11-based control software on nuclear fuel reprocessing plant
        I&C replacement on a French PWR
        Replacement of a safety monitoring system in a Swedish Nuclear plant (either the
           MOD modernisation project at Oskarshamm or the TWICE project at Ringhals 2).
The case studies will also help to refine the guidance produced, and the public guidance
handbooks will use a public domain refurbishment example to illustrate the application of
the guidance. Evaluation will also be supported by liaison via an Open Workshop and an
industrial interest group as the project progresses. We will also liase with projects in the
PLEM (Plant Life Extension and Management) cluster of the current FP-5 and specifically
with the BE-SECBS (Benchmark Exercise on Safety Evaluation of Computer Systems)
project.
The anticipated public domain deliverables will be ‘best practice’ guidance to assist the
utilities, regulators and manufacturers in achieving cost and safety advantages. The partners
will also disseminate to influential standards bodies.




                                              58
Nuclear Energy Programme                                    Plant life extension and
                                                            management
Operational safety of existing installations                Organisation and
                                                            Management of safety

Title:
LEARNING ORGANISATIONS FOR NUCLEAR SAFETY

Acronym                 LearnSafe

Proposal number FIS5-2001-00066                       Contract number FIKS-CT2001-00162

Type of action          Shared cost                   Duration           30 months

Starting date           1 November 2001               EC project officer S. Casalta

Total budget            1.175.258 €                   EC contribution    500.801 €


Co-ordinator
         Organisation          Technical Research Centre of Finland (VTT)
         Address               Tekniikantie 12
                               FIN-02044 Espoo
         Contact person        Prof. Björn Wahlström
                               Tel:      (358-9)4566400
                               Fax:      (358-9)4566752
                               Email     bjorn.wahlstrom@vtt.fi

Partnership

          Country                     Organisations

            D                   Technische Universitaet Berlin (TUB)
            UK                  Loughborough University (LBORO)
            E                   Centro de Investigaciones Energéticas, Medioambientales y
                                Tecnológicas (CIEMAT)
            S                   SwedPower AB
            E                   Asociacion Espanola de la Industria Electrica (UNESA)
            INT                 World Association of Nuclear Operators (WANO)
            FIN                 Teollisuuden Voima Oy (TVO)
            S                   Forsmarks Krattgrupp AB
            D                   E. ON Kernkraft GmbH
            D                   Kernkraftwerk Krümmel GmbH
            UK                  British Nuclear Fuels plc (BNFL)
            S                   OKG Aktiebolag
            S                   Ringhals AB




                                                 59
Project Summary
The main objective of the LearnSafe project is to create methods and tools for supporting
processes of organisational learning at the nuclear power plants (NPP). Organisational
learning has become increasingly important for the nuclear industry in its adaptation to
changes in the political and economic environment, changing regulatory requirements, a
changing work force, changing technology in the plants, and the changing organisation of
NPPs and power utilities. The danger during a rapid process of change is that minor
problems may trigger a chain of events leading to actual degrading of safety and/or
diminishing political and public trust in the safety standards of the particular NPP, utility or
corporation.
The focus of the project is senior managers at NPPs and power utilities who are responsible
for strategic choice and resource allocation. This focus was selected with the understanding
that their decisions, approaches and attitudes have an important influence both on safety and
economy of the NPPs. The LearnSafe project will develop methods and tools, which can be
used in the management of change, and in ensuring an efficient organisational learning.
Project results will include recommendations and inventories of good practices. The project
builds on and extends results of an earlier EU-project "Organisational factors; their definition
and influence on nuclear safety" (ORFA).

The project is set up in two major phases, which cover both theoretical considerations and
empirical investigations. The first phase places an emphasis on management of change and
the second on components of organisational learning. Both phases start with the creation of
data collection tools to be used in the empirical part of the work. The second theoretical and
empirical phase takes a major step towards developing methods and tools, which can be
applied by the NPPs themselves in creating maintaining efficient processes of organisational
learning.

One important feature of the project is a continuous interaction between the researchers and
managers at the NPPs in addressing issues connected to organisation and management,
which are important for safety and efficiency. Preliminary results of the project will be
presented and discussed in small workshops to be held at the NPPs during the project, to
ensure that relevant problems are addressed and solved in a practical way. It is assumed that
the participating NPPs will expand some of the LearnSafe tasks into small spin-off projects.

Five milestones are identified. The research model includes a framework of concepts and
phenomena to be considered in the project. Tools for describing organisations and data
collection instruments for the first empirical phase are also a part of the first milestone. The
second milestone marks the completion of the first major theoretical and empirical phase of
the project. The third milestone and the mid-project evaluation is based on the finalised
analysis of NPP approaches to change and the data collection methods and tools to be used in
the second phase of the project. A mid-project seminar for a larger audience for presenting
preliminary project results is also planned. The fourth milestone marks the completion of the
first major theoretical and empirical phase of the project. At that time a tentative set of
criteria for efficient learning organisations has been established and the preparation of the
final report has been started. The fifth milestone is connected to the completion of the
project. A final seminar will be used to collect comments to a draft final report. It is the
intention to place the completed final report in the public domain after due review by project
partners.



                                              60
Nuclear Energy Programme                                   Plant life extension and
                                                           management
Operational safety of existing installations               Organisation and
                                                           Management of safety

Title:
EVALUATION OF ALTERNATIVE APPROACHES FOR ASSESSMENT OF SAFETY
PERFORMANCE INDICATORS FOR NUCLEAR POWER PLANTS

Acronym                 SPI

Proposal number FIS5-2001-00041                     Contract number FIKS-CT2001-20145

Type of action          Concerted action            Duration            21 months

Starting date           1 October 2001              EC project officer S. Casalta

Total budget            337.850 €                   EC contribution     193.150 €


Co-ordinator
         Organisation         Gesellschaft für Anlagen - und Reaktorsicherheit (GRS) GmbH
         Address              Forschungsgelände
                              D-85748 Garching
         Contact person       Dr. Klaus Koberlein
                              Tel:      (49-89)32004445
                              Fax:      (49-89)32004306
                              Email     koe@grs.de

Partnership

          Country                   Organisations

            E                   Consejo de Seguridad Nuclear (CSN)
            CH                  ERI Consulting, Khatib, Attenhofer & Co.
            CH                  Swiss Federal Nuclear Safety Inspectorate (HSK)
            F                   Institut de Radioprotection et de Sureté Nucléaire (IRSN)
            UK                  Health & Safety Executive
            CH                  Nordostschweizerische Kraftwerke (NOK)
            S                   Swedish Nuclear Power Inspectorate (SKI)
            HU                  Institute for Electric Power Research Co.(VEIKI)
            S                   Swedpower AB
            SK                  Nuclear Regulatory Authority (UJD)

Project Summary
In the past several years the application of safety performance indicators (SPI) to nuclear
power plants became an important topic on national and international levels.



                                               61
The general objective of the proposed Concerted Action (CA) is to review and evaluate the
application of SPIs - 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, and it will exploit PSA techniques for the development
and use of meaningful alternative Safety Performance Indicators (SPIs). Since regulators and
operators will participate in the CA, it is expected that the CA will stimulate and enhance the
process of identifying best practices in the application of SPIs, commensurate with specific
needs of regulatory authorities and utilities and aiming towards risk-based safety
performance indicators, including the development of candidate methods and
recommendations for future developments, and preparation of relevant implementation
guidelines. The CA is expected to promote the transition to risk-informed performance-based
regulation in Europe.

The specific objectives of the proposed Concerted Action project are:
1. To review the existing approaches to collection and reporting of Safety Performance
   Indicators (SPIs);
2. To evaluate merits and limitations of current practice in various countries;
3. To identify best practices relative to the needs of the regulators and the utilities;
4. To formulate the relationship between safety inspection and performance monitoring, as
   manifested by the information obtained from SPIs;
5. To identify research needs as related to incorporation of the impact of organisational
   aspects on nuclear plant safety performance;
6. To develop a list of candidate methods and recommendations for future development and
   implementation guidelines for alternative SPIs.

Technical benefit will include (a) providing information on state of the art , and (b)
evaluating the potential for moving towards the development of a system of risk based safety
performance indicators (SPIs). Such indicators are intended to have a predictive capability in
order to enable an early indication of potential degradation in safety performance and,
importantly, provide a risk measure of how serious the situation may become. In addition,
the following safety related issues will also benefit from the CA:
 Potential improvement in operational safety for NPPs;
 Potential for the identification of the need for risk-guided inspection strategies;
 Identification of future research needs related to operational safety and risk management
    of NPPs, including influences of safety culture;
 Lower risk to European citizens and to the environment;
 Improve competitiveness by avoiding problems and costs of corrective measures;
 Improved safety balance within the fleet of the European NPPs.

The CA will be based on the broad project relevant experience of the CA partners, on
additional information about current practices collected from literature and by a specific
questionnaire and on the review of this material mainly during meetings and workshops.
Supplemented by some home work a list of SPI candidates will be compiled and the best
practice to use and implement them will be evaluated and finally disseminated to interested
parties in European countries during a public workshop at the end of the project.




                                              62
Nuclear Energy Programme                                                Plant life extension and
                                                                        management
Operational safety of existing installations                            VVER operational safety
                                                                        issues

Title:
IMPROVED ACCIDENT MANAGEMENT OF VVER NUCLEAR POWER PLANTS

Acronym                  IMPAM-VVER

Proposal number FIS5-2001-00117                                Contract number FIKS-CT2001-00196

Type of action           Shared cost                           Duration                     32 months

Starting date            1 November 2001                       EC project officer P. Manolatos

Total budget             1.169.928 € *                         EC contribution              699.942 € *


Co-ordinator
         Organisation              Technical Research Centre of Finland (VTT)
                                   Nuclear Energy
         Address                   Techniikantie 4C
                                   FIN-02044 Espoo
         Contact person            Mr. Heikki Holmström
                                   Tel:     (358-9)4565050
                                   Fax:     (358-9)4565000
                                   Email    heikki.holmstrom@vtt.fi

Partnership
          Country                       Organisations

        HU              KFKI Atomic Energy Research Institute (AEKI)
        D               Forschungszentrum Rossendorf e.V. (FZR)
        FIN             Fortum Nuclear Services Oy
        HU              Paks Nuclear Power Plant
        FIN             Lappeenranta University of Technology
        F               Commissariat à l'Energie Atomique (CEA)
        CZ              Nuclear Research Institute Řež plc (NRI) (*)
        SK              VUJE (*)
        SK              IVS (*)
        BG              IRNE (*)
_________________________
* Budget expanded following inclusion of new partners from Newly Associated States (NAS).




                                                         63
Project Summary
The objective of the project is to resolve a relevant safety issue identified in recent safety
studies carried out in the advanced VVER countries Hungary and Finland. The issue was
raised using analytical tools, but the resolution requires experimental investigation as well as
specific computer code validation. The results will be utilised in Hungary and Finland
directly, but they will undoubtedly affect the safety management practices in other "VVER
countries" as well. The information produced will effectively contribute to improved VVER
safety by providing important publicly available information for both utilities and safety
authorities.

In some VVER small break LOCA scenarios it has been found out that there may be
problems to depressurise the primary system in order to allow the emergency core coolant
injection from the low-pressure system. The main objective of this project is to investigate
which means and criteria for starting depressurisation measures, like feed and bleed, would
be most efficient. It will also assess whether the computer codes can adequately predict
important phenomena, like the effect of steam generator reverse heat transfer at low primary
inventories and at high temperature core processes.

The research activities will be divided in the following two work packages:

1. Experimental investigation using PMK-2 and PACTEL test facilities
2. Pre- and post-test analyses of the experiments using advanced codes

The emphasis is on experiments to find out whether the issues raised by earlier analytical
studies require consideration of changes in operating practices. Advanced computer codes
are used for both defining and analysing the experiments, and to assess their capabilities in
predicting the associated complex VVER related phenomena. Important European thermal
hydraulic system codes e.g. CATHARE, ATHLET, APROS will be used. The project will
utilise the two unique integral thermal hydraulic VVER440-facilities (the only ones in the
world) and benefit of their complementarity. The smaller Hungarian PMK facility is first
used to check the effects of all relevant initial parameters, and the larger multiloop Finnish
PACTEL facility, with higher operating costs, is used to investigate the most interesting
situations more realistically. By this kind of counterpart testing also the essential scaling
effects are to be addressed. The instrumentation of the facilities will be upgraded by special
advanced equipment from FZR, Germany. The German and French partners will also
provide valuable additional expertise with regard to computational tools. Industrial
participation by VVER utilities Fortum (formerly IVO) and Paks NPP ensures focusing on
production of practically useful results.




                                              64
Nuclear Energy Programme                                                Plant life extension and
                                                                        management
Operational safety of existing installations                            VVER operational safety
                                                                        issues

Title:
UNIFIED PROCEDURE FOR LIFETIME ASSESSMENT OF COMPONENTS AND PIPING
IN VVER NPPS

Acronym                  VERLIFE

Proposal number FIS5-2001-00120                                Contract number FIKS-CT2001-20198

Type of action           Thematic network                      Duration                     24 months

Starting date            1 October 2001                        EC project officer P. Manolatos

Total budget             220.898 € *                           EC contribution              220.898 € *


Co-ordinator
         Organisation              Ustav Jaderného Vyzkumu Rež A.S.
                                   Division of Integrity and Technical Engineering
         Address                   Rež 130
                                   CZ-25068 Rež
         Contact person            Dr. Milan Brumovsky
                                   Tel:      (420-2)20941110
                                   Fax:      (420-2)20940519
                                   Email     bru@ujv.cz

Partnership

          Country                       Organisations

        CZ              State Office for Nuclear Safety
        CZ              Dukovany Nuclear Power Plant Station
        CZ              SKODA
        FIN             Fortum Nuclear Services Ltd
        SK              Slovenske Elekrarne
        SK              Nuclear Power Plant Research Institute (VUJE) Trnava Inc
        HU              KFKI Atomic Energy Research Institute (AEKI)
        SK              Urad Jadroveho dozoru Slovenskej republiky
        BG              Institute of Metal Science (IMS) (*)
        CZ              Institute of Applied Mechanics Brno, Ltd (*)
        CZ              CEZ, a.s. Divize JE Jemelin (*)
_____________________________________
* Budget expanded following inclusion of new partners from Newly Associated States (NAS).




                                                         65
Project Summary
Lifetime assessment of individual components and piping in nuclear power plants (NPP) is a
mandatory part of every Periodic Safety Report as well as it is necessary for component/plant
life management and potential plant life extension. In the same time, such assessment is also
necessary for safe operation of components in NPPs. Today, no legal procedures or standard
guidelines exist for lifetime/integrity assessment of components and piping in operating
NPPs of VVER type. Former Soviet rules and standards were prepared and approved only
for design and manufacturing stage of NPPs. These rules/standards mostly are not applicable
for operating plants or they need some modifications and extensions to be usable also for
operating components. Approaches used in VVER Codes and standards are in some parts
different than they are applied in PWR ones, thus a comparison of lifetime assessment using
these two types of codes could be different and noncomparable.

Main goal of the project will be in a preparation, evaluation and mutual agreement of a
“Unified procedure for Lifetime Assessment of Components and Piping in VVER Type
Nuclear Power Plants”. This procedure should be based on former Soviet rules and codes, as
VVER components were designed and manufactured in accordance with requirements of
these codes and from prescribed materials. Then, critical analysis of possible application of
some approaches used in PWR type components will be performed and such approaches will
be incorporated into the prepared procedure as much as possible with the aim of a
harmonisation of VVER and PWR Codes and procedures.

Preparation of a Unified Procedure for VVERs operating in Finland, Czech Republic, Slovak
Republic and Hungary will increase the level of lifetime/integrity evaluation in these
countries and will help to elaborate a unified approach and fully comparable results between
individual plants and countries. Then, harmonisation with PWR codes allows to obtain
results that will be comparable, reliable and more sophisticated as similar approaches will be
used in both types of reactors.




                                             66
Nuclear Energy Programme                                  Plant life extension and
                                                          management
Operational safety of existing installations - RI         VVER operational safety
                                                          issues

Title:
AMES THEMATIC NETWORK ON AGEING

Acronym                 ATHENA

Proposal number FIS5-2001-00079                     Contract number FIR-CT2001-20170

Type of action          Thematic network            Duration          36 months

Starting date           1 November 2001             EC project officer P. Manolatos

Total budget            380.000 €                   EC contribution   380.000 €


Co-ordinator
         Organisation         Tractebel S.A.
                              Energy Engineering - Operation and Maintenance Dept.
         Address              Avenue Ariane, 4
                              B-1200 Brussels
         Contact person       Mr. Robert Gerard
                              Tel:     (32-2)7738363
                              Fax:     (32-2)7738900
                              Email    robert.gerard@tractebel.be

Partnership
          Country                   Organisations

            UK                 Magnox Electric plc
            UK                 LMD Consultancy
            INT                European Commission - JRC/IE
            CZ                 Nuclear Research Institute Rež (NRI)
            FIN                Technical Research Centre of Finland (VTT)

Project Summary
The AMES Thematic network on ageing, ATHENA, aims, within the « enlarged » Europe, at
reaching a consensus on important issues, identified by the AMES European Network
Steering Committee, that have an impact on the life management of nuclear power plants.
ATHENA creates a structure enhancing the collaboration between European-funded R&D
projects, national programs, and TACIS/PHARE programs. This will greatly increase the
return from the individual projects and maximise the European added value.




                                               67
The ATHENA Thematic network is organised in separate task groups carrying out different
work packages on important issues identified by the AMES European Network Steering
Committee. The membership of AMES Streering Committee, and its link with the SCORE
Committee (Safe and Competitive Operation of Reactors in Europe), ensures that the issues
covered by ATHENA are in line with the priorities of the European industry and Safety
Authorities and with the ageing management strategies of the member states.

The Work Packages of ATHENA are the following:
- Linking AMES strategy with Central and East Europe
- Master Curve implementation for fracture toughness assessment
- Annealing and re-embrittlement issues for nuclear power plant life management
- Radiation embrittlement understanding
- Ageing mechanisms: influence and synergism

This work is fully in line with the priorities defined in the chapter “operational safety of
existing installations” of the key action 2 (nuclear fission) of the Euratom program.
ATHENA brings together leading experts in each of these fields in order to integrate the
information coming from different programs on key ageing issues carried out in different
frameworks (EU-funded, national, TACIS-PHARE). ATHENA will establish the basis for a
common European position on the technical issues and ensure a wide dissemination of the
final results which will be presented in a final plenary AMES/ATHENA conference open to
a wide audience.




                                            68
Nuclear Energy Programme                                    Severe accident management
Operational safety of existing installations                Assessment of risks

Title:
CORE LOSS DURING A SEVERE ACCIDENT

Acronym                 COLOSS

Proposal number FIS5-1999-00013                       Contract number FIKS-CT1999-00002

Type of action          Shared cost                   Duration           36 months

Starting date           1 February 2000               EC project officer A. Zurita

Total budget            3.186.437 €                   EC contribution    1.600.000 €


Co-ordinator
         Organisation          Institut de Radioprotection et de Sûreté Nucléaire (IRSN/DRS)
         Address               CEA CADARACHE
                               F-13108 Saint-Paul-lez-Durance
         Contact person        Dr. Bernard Adroguer
                               Tel:      33 4 42 25 23 34
                               Fax:      33 4 42 25 29 29
                               Email     bernard.adroguer@irsn.fr

Partnership

          Country                     Organisations

            HU                  KFKI Atomic Energy Research Institute (AEKI)
            F                   Electricité de France (EDF)
            I                   Ente per le Nuove Tecnologie, l'Energia e l'Ambiente (ENEA)
            F                   Framatome ANP
            D                   Forschungszentrum Karlsruhe GmbH (FZK)
            INT                 European Commission - JRC/IE
            INT                 European Commission - JRC/ITU
            CH                  Paul Scherrer Institut (PSI)
            D                   Framatome ANP GmbH
            CZ                  SKODA-UJP Praha a.s.
            E                   Universidad Politécnica de Madrid (UPM)
            D                   Ruhr-Universität Bochum (RUB)
            D                   Universität Stuttgart - IKE
            FIN                 University Lappeenranta




                                                 69
Project Summary
The project complies with the "Nuclear fission programme" on Severe Accidents and
Management Measures. The objective of COLOSS is to improve the safety of European
reactors. Therefore additional research on core degradation is proposed combining
experiments, model developments and SA code improvement packages that address
uncertainties which significantly limit the understanding of both existing or future plant
behaviour (PWR, BWR, VVER, EPR) under Severe Accident conditions. Risk-relevant
topics selected are: BC4 oxidation and control rod degradation, high burn-up UO2 and MOX
behaviour, oxidation of U-O-Zr mixtures and liquefaction-collapse of fuel rods.

Specific objectives are.
- Experiments on selected topics at different scales for code developments and validation,
- Model development and coupling in SA codes used by Utilities, Industry and Safety
  Authorities,
- Evaluation of the consequences of results on safety, feedback on SAM for different plant
  designs.

Studies on risk-relevant core degradation topics are proposed for different plant designs:
PWR, BWR, VVER and EPR. Experiments at different scales up to integral tests (bundles)
are proposed on the following topics:
a) High burn-up UO2 and MOX dissolution by molten Zr and melting point of resulting U-O-
Zr mixtures and of a TMI-2 corium sample.
b) Simultaneous dissolution of UO2 and ZrO2 by molten Zr and rod collapse conditions for
prototypic PWR and VVER fuel rods (clad failure and loss of rod geometry due to UO2 and
ZrO2 dissolution phenomena).
c) Oxidation of B4C alone (pellets/powder from different plant designs) and degradation-
oxidation of prototypic B4C control rods representative of PWR, BWR and VVER rods. The
key point is the measure of gas and aerosols produced, in particular H2 and CH4 highly risk-
relevant for safety (H2-risk in containment, formation of Organic Iodine gas which cannot be
trapped in filters). Two integral tests are planned in QUENCH and CODEX facilities with a
fuel rod bundle and a central B4C control rod representative of PWR/BWR and VVER
designs.
d) Oxidation of U-O-Zr mixtures responsible of H2 production during late stages of core
degradation.

The aim of this experimental effort is to enable the development and validation of models
(B4C, MOX, U-O-Zr) which will be implement in European integral SA codes (ASTEC,
ICARE/CATHARE and ATHLET-CD/KESS). This analytical work is favoured by work-
packages, which include experimental and analyst teams. Finally emphasis has been put on
plant applications with different SA codes (ASTEC, ICARE/CATHARE, SCDAP/RELAP5,
MELCOR, MAAP-4), different plant designs sequences for PWR-1300, VVER-1000, BWR,
EPR and the TMI-2 reference accident. Calculations will be run by Designers, Utilities,
R&D and Safety authorities enabling benchmarks.

Plant calculations will be carried out to quantify the impact of new data and models on the
safety and to evaluate the consequences on risks and accident management measures. Special
emphasis will be put on H2-risk (oxidation of B4C and U-O-Zr), corium risk (burn-up effect,
MOX) and source term (CH4 -Organic Iodine production).



                                            70
Nuclear Energy Programme                                    Severe accident management
Operational safety of existing installations                Assessment of risks

Title:
LIMIT STRAINS FOR SEVERE ACCIDENT CONDITIONS

Acronym                 LISSAC

Proposal number FIS5-1999-00075                       Contract number FIKS-CT1999-00012

Type of action          Shared cost                   Duration           36 months

Starting date           1 February 2000               EC project officer A. Zurita

Total budget            2.685.000 €                   EC contribution    950.000 €


Co-ordinator
         Organisation          Forschungszentrum Karlsruhe GmbH (FZK)
         Address               Postfach 3640
                               D-76021 Karlsruhe
         Contact person        Dr. Rudolf Krieg
                               Tel:     49 7247 82 43 56
                               Fax:     49 7247 82 37 18
                               Email    maeule@irs.fzk.de

Partnership

          Country                     Organisations

            I                   Ente per le Nuove Tecnologie, l'Energia e l'Ambiente (ENEA)
            E                   Equipos Nucleares S.A.
            F                   Framatome ANP
            SI                  Institut "Josef Stefan"
            INT                 European Commission - JRC/IPSG
            D                   Staatliche Materialprufungsanstalt (MPA Stuttgart)
            NL                  Nuclear Research and Consultancy Group (NRG)
            CH                  Paul Scherrer Institute (PSI)
            D                   Framatome ANP GmbH
            EL                  University Aristotle of Thessaloniki
            FIN                 Technical Research Centre of Finland (VTT)

Project Summary
The local failure strains of essential reactor vessel components will be investigated. The size
influence of the components is of special interest. Typical severe accident conditions
including elevated temperatures are considered.




                                                 71
The main part of work consists of test families with specimens under uniaxial and biaxial
static and dynamic loads. Within one test family the specimen geometries are similar, but the
size is varied up to reactor dimensions. Special attention is given to geometries with a hole or
a notch causing non-uniform stress and strain distributions typical for reactor components.
There are indications that for such non-uniform distributions size effects may be stronger
than for uniform distributions. Thus size effects on the failure strains and failure processes
can be determined under realistic conditions.

Several tests with nominal identical parameters are planned for small size specimens. Thus
some information will be obtained about the scatter. A reduced number of tests is carried out
for medium size specimens and only a few tests are carried out for large size specimens to
reduce the costs to an acceptable level. For all specimens sufficient material is available from
a reactor pressure vessel. Thus the scatter of the material, which could impair the
interpretation of the test results, can be expected to be quite small. Nevertheless, an adequate
number of additional quality assurance tests are planned to cheek the material homogeneity.

To deepen the understanding of structural degradation and fracture and to allow
extrapolations, advanced computational method including damage models will be developed
and validated. In some cases so-called non-local concepts in other cases the description of
stochastic properties at the grain size level are considered.

Based on the results from the present research program and considering the findings in the
literature and the experience collected in industry, admissible strains will be proposed for
different severe accident requirements. If, for instance, leakages must be avoided, the
admissible strains will be moderate; if only the formation of missiles must be ruled out, the
values may be larger. In addition, more information will be gained about the failure process
of structures and the resulting damage.

Using these results a more realistic strain based evaluation concept can be employed and the
applicability of small-scale test results can be checked. Thus undue over- conservatism can
be avoided, accident management strategies can optimised, and it will become possible to
show that reactors are able to withstand many severe accidents. For demonstration selected
accident analyses will be performed.




                                              72
Nuclear Energy Programme                                    Severe accident management
Operational safety of existing installations                Assessment of risks

Title:
ASSESSMENT OF REACTOR VESSEL INTEGRITY

Acronym                 ARVI

Proposal number FIS5-1999-00006                       Contract number FIKS-CT1999-00011

Type of action          Shared cost                   Duration           36 months

Starting date           1 January 2000                EC project officer A. Zurita

Total budget            1.042.444 €                   EC contribution    700.000 €


Co-ordinator
         Organisation          Kungal Tekniska Högskalan
                               Nucl. Power Safety Div. of Dept. of Energy Technol.
         Address               Drotning Kristina Vag 33A
                               S-10044 Stockholm
         Contact person        Prof. Bal Raj Sehgal
                               Tel:      46 8 790 92 52
                               Fax:      46 8 790 9197
                               Email     sehgal@ne.kth.se

Partnership

          Country                     Organisations

            F                   Commissariat à l' Energie Atomique (CEA/DRN/DTP)
            FIN                 Fortum Nuclear Services Ltd
            F                   Framatome ANP
            CZ                  Nuclear Reseach Institute Rež plc (NRI)
            D                   Universität Stuttgart (IKE)
            US                  University Regents of California
            HU                  Institute for Electric Power Research Co.(VEIKI)
            FIN                 Technical Research Centre of Finland (VTT)

Project Summary
The project ARVI will be responsible for resolving the remaining issues of melt vessel
interactions after completion of the MVI project. The proposed work also includes the
application of the data and the validated methodology. The major focus of the project is on
determining (1) the creep behaviour of vessel, timing and modes of its failure with and
without penetrations, (2) effectiveness of the gap and the external cooling, (3) the effects of
the melt pool stratification observed in RASPLAV experiments. The ARVI project proposes



                                                 73
large-scale highly prototypic and innovative experiments for data on vessel creep and failure
modes, and on stratified pool convection.

The top level objective of the project ARVI is to resolve all the remaining issues that are
unresolved the melt vessel interaction during the late phase of the in-vessel progression of a
severe accident, resulting in accurate assessments about (a) the feasibility of promulgating
the in-vessel melt retention (IVMR) scheme in a plant or in its absence; (b) the time available
before vessel failure in which emergency accident management measures may terminate the
accident within the vessel. The second level objectives are (1) to determine the mode and
location of vessel failure, (2) to determine the effects of melt stratification, (3) to determine
the effectiveness of the gap and external cooling, (4) to determine the effect of an in-vessel
steam explosion on lower head, and (5) to apply the data and models for design of IVMR for
some specific plants.

The work is broken up into five packages. They are divided into tasks which are performed
by different partners. The work consists of experiments and analysis development. The major
experimental project is EC-FOREVER in which data is obtained on melt pool natural
convection and lower head creep and rupture. The EC-FOREVER experiments are the first
in the world in which vessels, containing heated melt, and the lower head walls maintained at
prototypic accident conditions, are ruptured. The products will be (1) the effectiveness of gap
cooling, (2) multiaxial creep laws for different vessel steels, (3) effect of penetrations, (4)
mode and location of lower head failure and (5) data for validation of computer codes.

Two other experimental projects are concerned with the effects of stratification and of the
metal layer on the thermal loads on the lower head wall during melt pool convection.
Another experimental project conducted at the ULPU facility will provide data and
correlations for the CHF for the external cooling of the lower head. The modelling activities
in the area of structural analyses are focussed on the support of EC-FOREVER experiments
as well as the exploitation of the data obtained from those experiments for creep modelling
and the validation of the industry structural codes. Work is also proposed for extension of
melt natural convection analyses to consideration of stratification, mixing and accurate
representation of turbulence (in the CFD codes). Other modelling activities are for (1) gap
cooling CHF, (2) lower head dynamic loading due to steam explosion inside and (3) simple
models for system code. Finally, the methodology and data will be applied to design of
IVMR severe accident management scheme for VVER-440/213s plants.

The results of the project ARVI will be disseminated to the partners in the regular project
meetings. In addition, specific workshops will be held to disseminate the results to
representatives from nuclear industry, nuclear utilities and nuclear regulators. Publications
from the project will be distributed to partners and to interested parties in the nuclear
enterprise in Europe.




                                               74
Nuclear Energy Programme                                    Severe accident management
Operational safety of existing installations                Assessment of risks

Title:
EUROPEAN NUCLEAR THERMODYNAMIC DATABASE (FOR IN- AND EX-VESSEL
APPLICATIONS)

Acronym                 ENTHALPY

Proposal number FIS5-1999-00001                       Contract number FIKS-CT1999-00001

Type of action          Shared cost                   Duration           36 months

Starting date           1 February 2000               EC project officer A. Zurita

Total budget            1.125.147 €                   EC contribution    599.863 €


Co-ordinator
         Organisation          Institut de Radioprotection et de Sûreté Nucléaire (IRSN/DRS)
         Address               BP 1
                               F-13108 Saint-Paul-lez-Durance
         Contact person        Dr - Ing Anne De Bremaecker
                               Tel:      33 4 42 25 35 01
                               Fax:      33 4 42 25 61 43
                               Email     anne.de-bremaecker@irsn.fr

Partnership

          Country                     Organisations

            UK                  AEA Technology Plc
            HU                  KFKI Atomic Energy Research Institute (AEKI)
            F                   Commissarait à l'Energie Atomique (CEA/DRN/DTP)
            F                   Commissarait à l'Energie Atomique (CEA/DTA/CEREM)
            F                   Electricité de France (EDF)
            B                   Belgian Nuclear Research Centre (SCK-CEN)
            D                   Framatome ANP GmbH
            CZ                  SKODA - UJP Praha a.s.
            F                   Thermodata
            B                   Université Catholique de Louvain (UCL)
            B                   Université Libre de Bruxelles (ULB)

Project Summary
The calculation of fuel degradation, melting, relocation, and ex-vessel spreading, and of
fission products retention/release are based on the physical properties of the corium
(viscosity, heat conductivity, density, solid/liquid fraction, etc.). These properties can be



                                                 75
deduced from the phase diagrams of the elements and systems present in the in- and ex-
vessel corium. Phase diagrams are obtained directly by experiments or indirectly by
thermodynamic measurements and models.

The objective of the ENTHALPY project is to obtain one unique European commonly
agreed thermodynamic database for in- and ex-vessel applications, well validated and to
develop methodologies to couple the database to Severe Accident codes used by end-users
i.e. at least utilities, Safety Authorities and nuclear designers.

In order to assemble the two existing nuclear thermodynamic databases in one database, the
thermodynamic modelling of the entire field from metal to oxide for a complex
multicomponent chemical system: O-U-Zr/(B-C)/Ag-In/Fe-Cr-Ni/Al-Ca-Mg-Si/Ba-(Ce)-La-
SrRu will first be done. The world-used CALPHAD method will be employed.

As reliable data are lacking in the key U-Zr-O and U-Zr-Fe-O systems and other in- an ex-
vessel sub-systems (B, B2O3, Pu, PuO2, Mo, Gd; Si, SiO2, Ca, CaO, Al203, etc) including
fission product with high decay heat (Ba), specific Separate Effect Tests will be performed to
obtain thermodynamic results (Tliq, Tsolidus, enthalpies, solubility limits).

The tests are performed to favour thermodynamic equilibrium and the instrumentation is
specifically oriented to measure directly or indirectly thermodynamic values or points of
diagram phases: control of po, minimisation of convection, closed ampoules, use of
thermogravimetry, thermal differential analysis for liquidus / solidus / eutectic temperatures
etc. Post-test analysis (metallography and chemical analysis) will be performed on all the
samples. Solidification process, segregation studies, diffusion layers, ablation studies,
reaction rates, tests with simulants etc. are out of the scope.

The new database will be improved by the results of the present proposal and validated
against global tests. Calculations on both condensed and gaseous phases will be performed
analysing experiments devoted to fuel degradation or fission product release from fuel pins
or from molten pools (RASPLAV, CIRMAT, etc.). The consequences of remaining
uncertainties on corium physical properties and behaviour will be evaluated.

Methodologies will be developed to effectively couple the database to severe accidents codes
and a recalculation of TMI2 with MAAP4 coupled to the database will be made. Finally, the
formal but necessary activities linked to the management, edition and documentation of the
database are also included in the project.




                                             76
Nuclear Energy Programme                                      Severe accident management
Operational safety of existing installations                  Assessment of risks

Title:
EX-VESSEL CORE MELT STABILISATION RESEARCH

Acronym                 ECOSTAR

Proposal number FIS5-1999-00016/70                    Contract number FIKS-CT1999-00003

Type of action          Shared cost                   Duration              48 months

Starting date           1 January 2000                EC project officer A. Zurita

Total budget*           4.439.000 €                   EC contribution* 2.399.000 €


Co-ordinator
         Organisation          Forschungszentrum Karlsruhe GmbH (FZK)
         Address               Postfach 3640
                               D-76021 Karlsruhe
         Contact person        Mr. Werner Scholtyssek
                               Tel:    49 7247 82 55 25
                               Fax:    49 7247 82 55 08
                               Email   werner.scholtyssek@psf.fzk.de

Partnership

          Country                     Organisations

            F                   Commissariat à l' Energie Atomique (CEA/DEN)
            F                   Electricité de France (EDF)
            I                   Ente per le Nuove Tecnologie, l'Energia e l'Ambiente (ENEA)
            D                   Forschungszentrum Rossendorf e.V. (FZR)
            D                   Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH
            CZ                  Nuclear Reseach Institute Rež plc (NRI)
            D                   Framatome ANP GmbH
            D                   Ruhr-Universität Bochum (RUB)
            D                   Universität Stuttgart (IKE)
            F                   Université de Provence
            D                   Universität Aachen
            S                   Royal Institute of Technology (KTH/RIT)
            D                   Becker Technologies GmbH




*
  Total eligible costs and EC contribution reduced respectively to 4.082.739 € and 2.279.836 €
following the changes in the consortium composition and the duration of the project.


                                                 77
Project Summary
In the frame of the overall challenge to realise additional safety margins for existing as well
as for future nuclear power plants the ECOSTAR project provides experimental and
theoretical investigations on core melt behaviour after failure of the reactor pressure vessel in
order to improve the accident mitigation concept.

The project programme is focussed on the completion of the understanding of the complex
phenomena involved as they are melt release, ex-vessel transport and long-term stabilisation.
This includes to improve modelling approaches for adequate computer codes.

Based on apparent R & D needs as well as requirements by the authorities a set of small- and
large-scale tests is performed using simulant as well as prototypic corium compositions
accompanied by detailed analytical work. Various European research teams and all relevant
facilities are involved providing a broad spectrum of technical and analytical resources.

The research involves experiments to quantify the dispersion effect as initiating ex-vessel
process step under various accidental conditions. Experimental and modelling investigations
are aimed at the erosion rate of the impinging jet on interacting structures.

A large-scale 2D-spreading experiment will provide further insight into the dominating
phenomena governing the core melt ex-vessel transport and demonstrate the suitability of a
dedicated spreading compartment as accident mitigation system. The accompanying
modelling work using the codes LAVA, THEMA and CORFLOW allows significant
contribution to their validation.

Specific experimental and theoretical work will be done to gain further detailed experience
on the corium solidification process and the phenomena acting during the interaction of
corium with structure materials.

For the long-term stabilisation of the melt top- and bottom-flooding concepts are further
developed to demonstrate the coolability of the spread melt.

Summarising the working programme, the demonstration of technical feasibility of
mitigation measures as well as the validation of a selected set of codes provides necessary
input for the definition of a convincing safety concept for both existing and future reactors.




                                               78
Nuclear Energy Programme                                    Severe accident management
Operational safety of existing installations                Assessment of risks

Title:
INTEGRAL LARGE SCALE EXPERIMENTS ON HYDROGEN COMBUSTION FOR
SEVERE ACCIDENT CODE VALIDATION

Acronym                 HYCOM

Proposal number FIS5-1999-00017                       Contract number FIKS-CT1999-00004

Type of action          Shared cost                   Duration          36 months

Starting date           1 February 2000               EC project officer G. Van Goethem

Total budget            1.420.521 €                   EC contribution   700.000 €


Co-ordinator
         Organisation          Forschungszentrum Karlsruhe GmbH (FZK)
         Address               Postfach 3640
                               D-76021 Karlsruhe
         Contact person        Mr. Werner Scholtyssek
                               Tel:    49 7247 82 55 25
                               Fax:    49 7247 82 55 08
                               Email   werner.scholtyssek@psf.fzk.de

Partnership

          Country                     Organisations

            D                   Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH
            F                   Institut de Radioprotection et de Sûreté Nucléaire (IRSN)
            INT                 European Commission - JRC/IE Petten
            RU                  Kurchatov Institute
            D                   Framatome ANP GmbH

Project Summary
The project aims at completion of the experimental data base needed for the verification of
newly developed analysis methods and codes to predict hydrogen combustion behaviour and
corresponding loads in complex multi-compartment geometry and on representative scale.
An experimental programme in the RUT facility of the Russian KURCHATOV Institute will
be performed with combustion modes, ranging from slow to fast turbulent deflagration, that
were not yet covered by previous experiments. The main focus will be on geometrical
aspects and inhomogeneous hydrogen concentrations. Although steam would generally be
present in accident atmospheres, test atmospheres will be dry and at ambient temperatures.
This is justified since the effect of steam on the reactivity of hydrogen-air mixtures was



                                                 79
studied extensively and is considered as being well known. In addition, tests at ambient
temperatures allow a more precise definition of the initial and boundary conditions.

The data base will be used for the validation of criteria, models and codes which were
developed by the partners to optimise practical implementation and operating conditions of
mitigation devices, and to qualify accident management measures.

Pre-test analysis will be performed for the planning of small and large scale experiments for
a) definition of boundary and initial conditions, which should allow to simulate typical
hydrogen specific situations in severe accidents, b) definition of instrumentation type,
number and location since the measured data must be suitable for validation of different
numerical tools and must allow detailed local and integral global interpretation of the
observed processes, and c) estimation of expected conditions during tests, e.g. combustion
regimes and loads, and prediction of performance of components and instrumentation.

Blind predictive calculations will be performed for a limited number of tests with typical
combustion regimes and in geometries of different complexity. Comparison with post-test
calculations is expected to give valuable information on code capabilities and on the range of
validity of model and code control parameters.

Experiments at relatively small scale will address characteristic features of turbulent flame
propagation and separate effects in relatively simple geometrical configurations. These
include venting, heat losses, concentration gradients, blockage ratio changes, channel cross-
section changes and multiple connections. Modification of measurement techniques will also
be tested at small scale. About 40 to 45 small scale tests will be conducted.

Large scale tests in multi-compartment geometries will be performed in the RUT facility to
examine processes of turbulent flame propagation in room chains. The geometry will include
up to 6 compartments with obstructions for effective flame acceleration. The total volume is
480 m3. An optional venting compartment can be added. Distribution rooms will allow
several possible directions of flame propagation. About 10 to 12 large scale tests will be
performed.
A selected number of suitable tests will be identified for benchmarking purposes. Relevant
data will be made available to users outside the project.

Post-test analysis will be made using best available tools, which include lumped parameter
codes, CFD codes and coupled systems. Detailed and careful analysis and comparison with
pre-test analysis results and with experimental data will allow to validate models and codes
for reactor typical applications. Final adjustment of some important parameters will give the
necessary confidence in the procedures and analysis tools so that reliable plant application
will be possible. Ranges of applicability, uncertainties and the degree of conservatism will be
given on the example of a full scale plant analysis.

The results of the experimental part of the project will complete the data base that is
necessary in the field of hydrogen combustion for the validation of numerical tools being
used for the analysis of hydrogen specific severe accident scenarios.
The analysis and validation work performed within the project will result in qualified
numerical tools with well defined ranges of applicability for hydrogen risk assessment and
mitigation in nuclear power plants applications.



                                              80
Nuclear Energy Programme                                    Severe accident management
Operational safety of existing installations                Assessment of risks

Title:
EUROPEAN VALIDATION OF THE INTEGRAL CODE ASTEC

Acronym                 EVITA

Proposal number FIS5-1999-00062                       Contract number FIKS-CT1999-00010

Type of action          Shared cost                   Duration          36 months

Starting date           1 February 2000               EC project officer G. Van Goethem

Total budget            2.279.380 €                   EC contribution   1.199.847 €


Co-ordinator
         Organisation          Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH
         Address               Postfach 10 15 64 / Schwertnergasse 1
                               D-50667 Köln
         Contact person        Dr. Hans-Josef Allelein
                               Tel:      49.221 206.86.68
                               Fax:      49.221 206.88.88
                               Email     all@grs.de

Partnership

          Country                     Organisations

            A                   Austrian Research Centre Seibersdorf (ARCS)
            D                   Becker Technologies GmbH
            F                   Electricité de France (EDF)
            E                   Empresarios Agrupados (EA)
            I                   Ente per le Nuove Tecnologie, l'Energia e l'Ambiente (ENEA)
            F                   Framatome ANP
            D                   Forschungszentrum Karlsruhe GmbH (FZK)
            F                   Institut de Radioprotection et de Sûreté Nucléaire (IRSN)
            SK                  Inzinierska Vypoctova Spolocnoast (IVS) Trnava
            INT                 European Commission - JRC/IE
            CZ                  Nuclear Reseach Institute Rež plc (NRI)
            D                   Framatome ANP GmbH
            SK                  Urad Jadroveho dozoru Slovenskej republiky (UJD SR)
            D                   Ruhr-Universität Bochum (RUB)
            D                   Universität Stuttgart (IKE)
            HU                  Institute for Electric Power Research Co.(VEIKI)
            SK                  Nuclear Power Plant Research Institute (VUJE) Trnava
            F                   Commissariat à l'Energie Atomique (CEA)



                                                 81
Project Summary
The ultimate intention beyond this shared cost action is to provide end-users like utilities,
vendors and licensing authorities with a well validated European Integral Code for the
simulation of severe accident sequences in NPPs. The main objective of this proposal is to
distribute the integral code ASTEC to European Partners in order to apply the validation
strategy issued from the VASA project (4th EC Framework Programme 1994-1998) to
ASTEC. Based on this a guidance of numerical modelling and validation will be developed
and applied to ASTEC, so that the reliable simulation of severe accident sequences and
severe accident management measures will be possible with the code. This will lead to the
use of the knowledge gained in 4th EC Framework Programme in the ASTEC development
and validation process. The close co-operation of code-developers, validating institutions,
and end-users is of special benefit for the proposed work.

Key experiments and severe accidents sequences, which form the basis of analysis, will be
selected and defined. Following the risk-oriented approach of the VASA project a guidance
for the ASTEC validation process fitting for specific end-user needs as well as for research
requirements will be established. The two ASTEC developing organisations have to supply
the other project partners with the code. The first release of ASTEC is foreseen for the mid
of 2000, the second one of an improved code version for the beginning of 2002. Both of
these versions will be made available for all the users on their different platforms. Then the
validation process of ASTEC based on the experiments defined before will be performed.
The experiments may be taken from the test series PBF-SFD, LOFT, STORM, BETA, HDR,
VANAM, PHEBUS and others. Furthermore plant applications with ASTEC for the severe
accident sequences defined before, and for the demonstration of the capability for studying
accident management measures will be performed. The sequences should be representative
for different types of reactors like PWR, BWR, VVER and future concepts like EPR. After
the discussion of the results the various European ASTEC users will give their feedback to
the developing organisations. Finally the specific users' needs concerning further ASTEC
development will be harmonised.

Guidelines for the validation of the integral code ASTEC will be commonly defined and
documented by developers, researchers, and end-users. The basis version of ASTEC and an
improved one will be available for the partners on different platforms. The extension and
quality of the ASTEC validation will be increased considerably. The validation status
reached and the needs for further ASTEC development will be defined by the partners with
special attention to specific end-users' requirements.




                                             82
Nuclear Energy Programme                                    Severe accident management
Operational safety of existing installations                Assessment of risks

Title:
LATE PHASE SOURCE TERM PHENOMENA

Acronym                 LPP

Proposal number FIS5-1999-00023                       Contract number FIKS-CT1999-00005

Type of action          Shared cost                   Duration           36 months

Starting date           1 February 2000               EC project officer A. Zurita

Total budget            1.341.519 €                   EC contribution    749.998 €


Co-ordinator
         Organisation          AEA Technology PLC
         Address               Winfrith Technology Centre
                               UK-DT2 8DH Dorchester, Dorset
         Contact person        Dr Christopher Benson
                               Tel:      44.13.05. 20 2751
                               Fax:      44.13.05.20.26.63
                               Email     christopher.benson@aeat.co.uk

Partnership

          Country                     Organisations

            F                   Institut de Radioprotection et de Sûreté Nucléaire (IRSN)
            RU                  Leningrad Special Integrated Plant "Radon"
            CZ                  Nuclear Reseach Institute Rež plc (NRI)
            D                   Framatome ANP GmbH
            D                   Ruhr-Universität Bochum (RUB)

Project Summary
The aim of the project is to quantify fission product and core materials released from molten
corium. This work will examine the kinetics of release of the key fission products,
lanthanides and actinides (as simulants), and also allow the importance of key phenomena
(e.g. sparging, slag formation, two-phase systems) to be determined. There will also be a
series of experiments to aid with understanding the chemistry of species released during the
late phase of an accident, together with experiments aimed at examining the long-term
behaviour of a solidified core immersed in a water pool. The results from these experiments
will be used to test and develop generic code models for the relevant phenomena. The
models and experimental results will also be used in plant calculations for Eastern, Western
and advanced reactor designs. The programme will aid in the development of severe accident



                                                 83
management strategies and provide data on fission product behaviour in the late phase of an
accident. The objectives of the project are as follows:
(i) to provide an experimental database on the kinetics of release of fission products and
      core materials from molten pools, with emphasis on sparging, slag formation, two
      phase pools, the oxygen potential of the atmosphere and melt composition;
(ii) to conduct experiments on the release behaviour of lanthanides and actinides, or their
      simulants;
(iii) to provide data on the identity and physico-chemical form of fission products released
      in the late phase of an accident;
(iv) to provide experimental data on the long-term behaviour of fission products leached
      from a solidified core immersed in a water pool, with consideration of the entrainment
      of leached fission products;
(v) to conduct experimental assessments, model development and code testing using
      experimental data;
(vi) to conduct sensitivity studies to address long-term coolability and impacts on
      radiological source term in the context of plant calculations for existing and future
      reactors.

Experiments will be conducted to study the processes affecting fission product and core
materials release from molten pools. These will study the effects of temperature, oxygen
potential, sparging, slag formation, two-phase pools and melt composition on the release.
The behaviour of simulant lanthanides and actinides will also be studied. In addition to the
parameters for the metallic melt experiments, crust effects may be studied in the ceramic
melt tests. Experiments will also be conducted on the transport and aerosol behaviour of
fission products released in the late-phase, with emphasis on the behaviour of Ru, Ba and Sr.
These studies will examine the species formed under accident conditions. In addition,
experiments will be conducted that utilise solidified simulant core material. This will be
immersed in water at ~ 100°C for long periods of time. An assessment of the experimental
conditions will be made, so that they will be as representative as possible.

The experimental and modelling studies will be integrated with plant assessments. These will
examine the consequences of an accident in terms of the plant behaviour (radiological source
term and coolability of the molten pool). These will also aid with the development of severe
accident management strategies (e.g. immersed core). Sensitivity studies will compare
different codes with the same data. Operational Western and Eastern European plants and
advanced reactors will be studied.




                                             84
Nuclear Energy Programme                                                Severe accident management
Operational safety of existing installations                            Assessment of risks

Title:
VALIDATION OF SEVERE ACCIDENT CODES AGAINST PHEBUS FP FOR PLANT
APPLICATIONS

Acronym                  PHEBEN 2

Proposal number FIS5-1999-00057                                Contract number FIKS-CT1999-00009

Type of action           Shared cost                           Duration                     48 months

Starting date            1 March 2000                          EC project officer A. Zurita

Total budget             1.271.048 € *                         EC contribution              619.078 € *

Co-ordinator
         Organisation              European Commission - Joint Research Centre - IE
         Address                   Postbus 2
                                   NL-1755 ZG Petten
         Contact person            Dr Alan Victor Jones
                                   Tel:     39.0332.78.96.29
                                   Fax:     39.0332.78.58.15
                                   Email    alan.jones@jrc.it
Partnership

          Country          Organisations
            UK          AEA Technology Plc
            F           Commissariat à l' Energie Atomique (CEA/DRN/DTP)
            E           Centro de Investigaciones Energéticas, Medioambientales y
                        Tecnológicas (CIEMAT)
        I               Ente per le Nuove Tecnologie, l'Energia e l'Ambiente (ENEA)
        D               Forschungszentrum Karlsruhe GmbH (FZK)
        D               Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH
        F               Institut de Radioprotection et de Sûreté Nucléaire (IRSN)
        EL              National Centre for Scientific Research "Demokritos"
        NL              Nuclear Research and Consultancy Group (NRG)
        CH              Paul Scherrer Institute (PSI)
        E               Universidad Politécnica de Madrid (UPM)
        I               Università di Pisa
        HU              VEIKI (*)
        BG              Technical University of Sofia (*)
        BG              ENPRO Consult (*)
        RO              INR (*)
___________________________
* Budget expanded following inclusion of new partners from Newly Associated States (NAS).




                                                         85
Project Summary
The Phebus tests are a unique source of representative integral source term data. In this
project the partners will apply detailed codes and partners' expertise to understand and
quantify the physical and chemical phenomena underlying the Phebus results. They will
combine this understanding with optimised calculations of the tests using integral codes such
as are generally applied for plant analysis to obtain objective information on the strengths
and weaknesses of such codes. By examining the code requirements for plant assessment and
the lessons learned from the Phebus validation work guidelines for code application and
indications of expected uncertainties of direct utility to the user community will be
developed.

The extensive data from the multifaceted Phebus programme will be analysed in two main
ways; detailed codes for circuit transport and chemistry, CF1) codes and deposition codes for
containment thermalhydraulics and aerosol behaviour, and iodine chemistry codes will be
applied by experts to gain understanding of the phenomena underlying the measurements
made in the Phebus experiments and to identify key mechanisms. The same codes are
applied to determine the main sensitivities and to identify the strengths and weaknesses of
state of the art models.

In parallel, integral codes commonly applied in plant analysis will be used to make
"best-shot" analyses of the Phebus tests, supplemented by sensitivity studies, as is usual in
plant assessments. Using the results and the understanding gained from the analyses with
detailed codes, judgements will be formulated on the models in the integral codes and on the
way in which they are integrated into an overall framework. Both absolute results (code to
Phebus data) and relative information (code to code) will be factored into the judgements.

Based on indications of risk importance criteria for the assessment of integral codes for plant
assessment will be developed, including information on uncertainties and the determination
of safety margins, and consideration of their application in the evaluation of severe accident
management measures.

Finally, by combining the foregoing elements guidelines for the optimum use of integral
plant assessment codes will be developed, with indications (quantitative where possible) of
the expected uncertainties. The intended audience includes all end-users: designers, operators
and regulators.

The results and conclusions of the detailed analyses will be documented in interpretation
reports for the individual Phebus experiments, as will the information from the validation of
integral codes. The Progress Reports will provide a more integrated view spanning several
tests. The outcome of the code criteria work will be reported separately, and the code user
guidelines and uncertainty information will be included in the Final Report.




                                              86
Nuclear Energy Programme                                  Severe accident management
Operational safety of existing installations              Assessment of risks

Title:
ARCHIVE MODELS FOR SOURCE TERM INFORMATION AND SYSTEM MODELS

Acronym                 ASTERISM II

Proposal number FIS5-1999-00019                     Contract number FIKS-CT1999-20001

Type of action          Concerted action            Duration           18 months

Starting date           1 February 2000             EC project officer A. Zurita

Total budget            299.178 €                   EC contribution    299.178 €


Co-ordinator
         Organisation         AEA Technology PLC
         Address              Winfrith Technology Centre
                              UK-DT2 8DH Dorchester, Dorset
         Contact person       Dr. Ann Tuson
                              Tel:     44.1305.20.21.95
                              Fax:     44.1305.20.25.08
                              Email    ann.tuson@aeat.co.uk

Partnership

          Country                   Organisations

            INT                 European Commission - JRC/IPSC
            UK                  National Nuclear Corporation (NNC) Limited

Project Summary
The main aim of this project is to extend a prototype database on source term phenomena
begun within the CEC IV Framework Programme to encompass all the data arising from the
source term cluster projects. Issues associated with the further development of the archive to
all nuclear fission safety areas (e.g. core degradation and containment) as well as more
substantial projects (i.e. Phebus-FP) will be addressed, together with defining a method to
ensure that data from all current and future nuclear fission safety projects could readily be
incorporated within the archive.

Under the 4th Framework Programme, a concerted action was put in place to develop the
Archive for Source Term Information and System Models (ASTERISM) prototype. This
project was concerned with ensuring that the output of the ten projects then within the
Source Term Cluster of the Nuclear Fission Safety programme were readily available to
future research projects, in the form of summaries of the projects, and the necessary data or



                                               87
models. Within the ASTERISM prototype project, a catalogue was compiled of the key
information arising from the source term projects, the archiving system for this information
was designed, and a pilot archive established based on data from 2 projects. User feedback
was sought at all stages, of the project. It is now proposed to extend the pilot archive to
encompass information from all the source term projects and to update the catalogue of
information issued under the ASTERISM project to include data generated after the issue of
the original catalogue. The importance of the end-user is increasingly apparent in all aspects
of research. The original ASTERISM project focused particularly on the requirements of the
research worker. However, it is essential that the key results, data and models from research
programmes are distilled into summaries designed to meet the requirements of both the
regulator and nuclear industry. The other part of this task is therefore to provide clear
summaries of the source term research projects undertaken to date for the different end users.
It is also proposed to develop the system to allow further extension in future, including
existing data/models from all nuclear fission projects (4th Framework Programme); data from
more substantial projects, notably Phebus-FP; ongoing (5th Framework) and future projects.




                                             88
Nuclear Energy Programme                                   Severe accident management
Operational safety of existing installations               Assessment of risks

Title:
EUROPEAN EXPERT NETWORK FOR THE REDUCTION OF UNCERTAINTIES IN
SEVERE ACCIDENT SAFETY ISSUES

Acronym                 EURSAFE

Proposal number FIS5-2001-00044/46                  Contract number FIKS-CT2001-20147

Type of action          Thematic network            Duration            24 months

Starting date           1 December 2001             EC project officer A. Zurita

Total budget            540.046 €                   EC contribution     400.000 €


Co-ordinator
         Organisation         Institut de Radioprotection et de Sûreté Nucléaire (IRSN)
         Address              Bât. 219
                              F-13108 Saint-Paul-lez-Durance
         Contact person       Dr. Daniel Magallon
                              Tel:      (33-4)42254920
                              Fax:      (33-4)42256465
                              Email     magallon@drncad.cea.fr

Partnership
          Country                   Organisations
            F                  Commissariat à l'Energie Atomique (CEA)
            F                  Electricité de France (EDF)
            FIN                Teollisuuden Voima Oy (TVO)
            UK                 AEA Technology Plc
            CZ                 Nuclear Reseach Institute Rež plc (NRI)
            D                  Forschungszentrum Karlsruhe GmbH (FZK)
            D                  Universität Stuttgart
            D                  Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH
            D                  Framatome ANP GmbH
            S                  Royal Institute of Technology (KTH/RIT)
            HU                 KFKI Atomic Energy Research Institute (AEKI)
            E                  Consejo de Seguridad Nuclear (CSN)
            E                  Universidad Politécnica de Madrid (UPM)
            INT                European Commission - JRC/IE
            HU                 Institute for Electric Power Research (VEIKI)
            UK                 Health and Safety Executive (HSE)
            CH                 Paul Scherrer Institut (PSI)
            US                 Nuclear Regulatory Commission



                                               89
Project Summary
The objective of the EURSAFE network is to establish a large consensus on the Severe
Accident issues where large uncertainties still subsist and to propose a possible Severe
Accident Networking of Excellence structure to address these issues through concerted and
optimised Research programmes. First, a PIRT (Phenomena Identification and Ranking
Tables) is establish for each phase of Severe accident from core degradation up to release of
fission products in the containment, taking into account any possible counter-measures and
the evolution of fuel management. Second, the PIRT implications and actions are determined
taking into account existing and planned European facilities, codes and programmes. Third,
recommendations are made for the structure of a future Networking of Excellence. Fourth, a
consolidated framework for the preservation of integral severe accident data used for the
assessment of computer codes performance in nuclear reactor conditions data is proposed.

The project is divided into five work packages: Project management and reporting, PIRT,
PIRT implications, Networking of Excellence structure, Severe Accident Data Base
structure. The “PIRT” Work package is divided into three safety oriented sub-groups
(Primary circuit, Containment, Source term) and five phenomena oriented sub-groups (In-
Vessel phenomena, Ex-Vessel phenomena, Dynamic loading, Long term loading, Fission
products). The safety oriented groups are in charge of identifying and ranking the
phenomena according to their importance for safety. The phenomena oriented groups will
rank these phenomena in terms of knowledge Ratio (KR) will be established during the
quick-off meeting. Each sub-group of the PIRT work package is co-ordinated by a chairman
and a vice-chairman, one specialist of reactors, one specialist of phenomena. The participants
to a sub-group are nominated on a case by case basis. The chairman of the sub-groups report
to the co-ordinator of the work package by means of meeting reports and a final report. The
“PIRT implications” work package is in charge of defining R&D needs in terms of
objectives and priorities, identifying the required R&D tasks, reviewing the European
facilities and codes which could be used for these tasks, taking into account the existing and
planned programmes. The “Networking of Excellence structure” work package proposes a
conceptual organisation of a European Networking Of Excellence for Severe Accidents to
address the remaining uncertainties on the key safety issues by optimising the use of the
resources available in Europe. The "Severe Accident Data Base structure" work package
assesses the current practices for preservation of the data, identify the access requirement for
developers and users, formulates guidelines and design a platform for the best preservation
and access to the data.

Five two-day meetings will establish the PIRT within 16 months after project start. A final
report including all the phases of the PIRT is issued. Two two-day meetings will derive the
PIRT implications and review the existing European capabilities and programmes within 6
months after the PIRT. A document is produced. One two-day meeting will elaborate the
recommendations for the Networking of Excellence within the last 3 months of the project
with the edition of a final report. The work on the data base will be performed in parallel
with three two-day meetings each six months, and edition of a specific report and
distribution of a platform accessible on the WEB at the end of the project.




                                              90
Nuclear Energy Programme                                                Severe accident management
Operational safety of existing installations                            Assessment of risks

Title:
THEMATIC NETWORK FOR A PHEBUS FPT-1 INTERNATIONAL STANDARD
PROBLEM

Acronym                  THENPHEBISP

Proposal number FIS5-2001-00048                                Contract number FIKS-CT2001-20151

Type of action           Thematic network                      Duration                     24 months

Starting date            1 December 2001                       EC project officer A. Zurita

Total budget             240.899 € *                           EC contribution              240.899 € *


Co-ordinator
         Organisation              Institut de Radioprotection et de Sûreté Nucléaire (IRSN)
                                   Dept. de Recherches en Sûreté (DRS)
         Address                   Centre d'Etudes Nucléaires Bat. 702
                                   F-13108 Saint-Paul-lez-Durance
         Contact person            Dr. Bernard Clément
                                   Tel:      (33-4)42257646
                                   Fax:      (33-4)42252929
                                   Email     bernard.clement@irsn.fr

Partnership

          Country                       Organisations

        INT                          European Commission - JRC/IE
        UK                           AEA Technology Plc.
        HU                           KFKI Atomic Energy Research Institute (AEKI)
        CZ                           Nuclear Reseach Institute Rež plc (NRI)
        I                            Ente per le Nuove Tecnologie, l'Energia e l'Ambiente (ENEA)
        I                            University of Pisa
        E                            Universidad Politecnica de Madrid (UPM)
        D                            Forschungszentrum Karlsruhe GmbH (FZK)
        D                            Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH
        CH                           Paul Scherrer Institute (PSI)
        B                            Belgian Nuclear Research Centre (SCK-CEN)
        F                            Electricité de France (EDF)
        SI                           Jozef Stefan Institute (*)
        BG                           Enpro Consulting (*)
____________________
* Budget expanded following inclusion of new partners from Newly Associated States (NAS).




                                                         91
Project Summary
The Phébus-FP programme comprises six integral experiments on reactor severe accidents
dealing with fuel degradation, hydrogen production, fission product release, transport and
behaviour in the containment. The aim of the Project is to perform an International Standard
Problem, following the OECD/NEA methodology, on the second Phébus-FP test, FPT-1,
performed in July 1996. The ultimate goal of such an exercise is to provide insight on the
applicability of severe accident codes to the reactor case, by benchmarking them on an
integral experiment, closer to real situations than any experiment performed so far. In
addition, the Project will help the dissemination of the knowledge acquired by the Phébus-FP
Programme throughout the European (and international) community.

The first step will be the production of a specification report, including the data needed to
model the experiment, the experimental boundary conditions, the experimental data, required
information on the codes used (models, assumptions…), and required results from
calculations. This work will be made by the co-ordinator. The report will be reviewed and
revised as necessary taking into account participants' comments.

In the second step, each participant will perform calculations. As Phébus-FP experiments
are integral, the participants will be encouraged to perform integral calculations.
Nevertheless, it will be possible to calculate only one part of the experiment: fuel
degradation and associated fission product and hydrogen release, transport in the circuit,
thermal-hydraulic and aerosol behaviour in the containment, and iodine chemistry. The
calculation results will be presented during an intermediate comparison workshop.

In the third step, the whole set of calculations will be compared with the experimental
results. This work will be compiled in a final comparison report, made by the co-ordinator.
The report will be reviewed during a final workshop. Following an additional meeting on the
lessons learnt regarding plant calculations, it will then pass through the OECD/NEA/CSNI
review process (Working Group on Analysis and Management of Accidents) and be issued
as an OECD report.

The Project should lead to conclusions on the adequacy of models and computer codes to
reproduce the main results of an integral experiment such as Phébus FPT-1. It should also
lead to recommendations for improvement to the codes, if necessary. Finally, it should
provide insights on the applicability of codes to predict the consequences of severe accidents
in a nuclear power plant.




                                             92
Nuclear Energy Programme                                  Severe accident management
Operational safety of existing installations - RI         Assessment of risks

Title:
SCALING OF CONTAINMENT EXPERIMENTS

Acronym                 SCACEX

Proposal number FIS5-2001-00017                     Contract number FIR1-CT2001-20127

Type of action          Thematic network            Duration           12 months

Starting date           1 January 2002              EC project officer A. Zurita

Total budget            372.812 €                   EC contribution    372.812 €


Co-ordinator
         Organisation         Becker Technologies GmbH
         Address              Koelnerstrasse 6
                              D-65760 Eschborn
         Contact person       Dr. Karsten Fischer
                              Tel:      (49-6196)936116
                              Fax:      (49-6196)936100
                              Email     fischer@becker-technologies.com

Partnership

          Country                   Organisations

            D                   Framatome ANP GmbH
            D                   Forschungszentrum Karlsruhe GmbH (FZK)
            UK                  The Victoria University of Manchester
            F                   Institut de Radioprotection et de Sûreté Nucléaire (IRSN)
            I                   Università di Pisa
            I                   Università di Roma
            CZ                  Nuclear Research and Consultancy Group (NRG)
            D                   Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH
            SK                  Nuclear Power Plant Research Institute (VUJE) Trnava Inc
            E                   Empresarios Agrupados Internacional SA
            F                   Electricité de France (EDF)

Project Summary
The project addresses the problem how results from experiments can be transferred to real
reactor conditions in a systematic and reliable way. A network of European experts shall
apply scaling methods to a variety of containment-related thermal-hydraulic and material
behaviour experiments. Objectives are:



                                               93
- Identify applicability and requirements of existing scaling theory
- Document example cases of scaling analysis as reference for future applications
- Find out common features and rules of scaling analyses in different application fields
- Identify the significance of scaling for experiments and modelling
- Assess the needs, potentials and benefits of a Scaling Handbook in nuclear reactor
technology

The network of European experts will conduct scaling analyses on a number of experiments
that have been or are being performed in the EURATOM research programme, like e.g. the
projects DABASCO (containment data base), VOASM (containment flows), ATHERMIP
(containment penetration sealings), CESA (concrete containment leakage), and others. The
work of the group will be supported by an internationally acknowledged scientist in the field
of scaling methods, who will summarise the theoretical basis and scrutinise the applications.
The results will be assembled in a report with the main parts
- Introduction
- Theoretical framework
- Scaling of thermal-hydraulic processes
- Scaling of material behaviour
- Conclusions

In addition to the traditional dimensional analysis of small-scale thermal hydraulic
experiments, tests with artificial materials (helium instead of hydrogen), time scales (e.g.
accelerated thermal ageing), or material loads (irradiation by gamma rays instead of
neutrons) will be discussed. Special applications cover heat transfer, spray and bubble
condenser effects, natural convection flow, cable ageing, containment sealings and leaks.
The role of computer codes in the scaling method will be identified and illustrated. While the
present work is concentrated on containment aspects, the perspectives for extension to other
fields like primary system, core melt and severe accident will be discussed, that may result in
a new activity to establish a more comprehensive Scaling Handbook in nuclear reactor
technology. The working period will be one year, and it will be structured by 3 meetings of
the project members.




                                              94
Nuclear Energy Programme                                     Severe accident management
Operational safety of existing installations - RI            Assessment of risks

Title:
PLATFORM FOR IMPROVEMENTS IN NUCLEAR INDUSTRY AND UTILITY SAFETY

Acronym                 PLINIUS

Proposal number FIS5-2001-00049                        Contract number FIR1-CT2001-40152

Type of action          Research infrastructure        Duration           36 months

Starting date           1 December 2001                EC project officer A. Zurita

Total budget            699.300 €                      EC contribution    699.300 €


Co-ordinator
         Organisation          Commissariat à l'Energie Atomique (CEA) - Direction de
                               l'Energie Nucléaire
                               Dept. Thermohydraulique et Physique
         Address               BP1
                               F-13108 Saint-Paul-lez-Durance
         Contact person        Mr. Christophe Journeau
                               Tel:      (33-4)42254121
                               Fax:      (33-4)42256465
                               Email     cjourneau@cea.fr

Partnership


Project Summary
This project is aimed at providing support for European research to conduct experiments
with prototypic corium in the PLINIUS experimental facility.

The PLINIUS platform at CEA Cadarache is made of 4 facilities for the experimental study
of molten mixtures containing depleted UO2.
 VULCANO: a 300 kW plasma arc furnace able to reach 3000 K and to melt and pour 50-
    100 kg of prototypic corium. Spreading test sections and crucible for sustained heating
    have been used in VULCANO. Specific instrumentation including high temperature
    thermometers and up to 8 video and infrared cameras follow the corium evolution.
 COLIMA: a smaller scale facility in which a few kilogram of corium can be molten by
    induction (150 kW available). The crucible is installed in an instrumented enclosure with
    a temperature controlled wall capable of representing accidental containment
    configuration. It is devoted to aerosol and material interaction studies as well as to the
    determination of some physical properties.
 VITI: a facility for the determination of corium viscosity and surface tension, using the


                                                  95
    levitated droplet technique. It uses only a few millilitres of corium.
   KROTOS: devoted to steam explosions, this facility in which a few kilograms of
    prototypic corium are poured into water has been developed and operated by the Joint
    Research Centre. It is now CEA property and is being transferred to the PLINIUS
    platform.

The team operating this platform for 5 years has gained a valuable experience in the making
and measurement of corium. Currently, it is the sole platform in the EU operating with
prototypic corium.

Since Experiments with prototypic corium are a necessary step for maintaining the European
R&D potential necessary to the mastering of sever accidents, transnational access will be
proposed by public calls for European users which will be publicised on the web and in
scientific/technical journals. The choice of the user groups will be made by an international
expert panel. The results of the experiments will be made publicly available (except in the
case of first access by Small and Medium-sized Enterprises).




                                             96
Nuclear Energy Programme                                      Severe Accident Management
Operational safety of existing installations - RI             Assessment of risks

Title:
LARGE SCALE EXPERIMENTS ON CORE DEGRADATION, MELT RETENTION AND
COOLABILITY

Acronym                 LACOMERA

Proposal number FIS5-2002-00007                         Contract number FIR1-CT2002-40158

Type of action          Research Infrastructures        Duration           36 months

Starting date           1 September 2002                EC project officer A. Zurita

Total budget            1.250.000 €                     EC contribution    500.000 €


Co-ordinator
         Organisation          Forschungszentrum Karlsruhe, GmbH,
                               Institut fuer Nukleare Entsorgungstechnik
         Address               Hermann-von-Helmholtz-Platz, 1; PO Box 3640
                               D-76021 Karlsruhe
         Contact person        Mr. Alexei Miassoedov
                               Tel:     (49-7247) 822553 or 824981
                               Fax:     (49-7247) 822095 or 824567
                               Email    alexei.miassoedov@imf.fzk.de


Project Summary
During the last years, concerns of nuclear safety experts have concentrated on residual safety
problems associated with core quenching and melt retention. To improve our understanding
of core melt formation and corium behaviour and to allow qualified accident management to
terminate the accident, further experiments are proposed from different institutions. Such
experiments can also be used to validate and improve computer models, which are being
developed in the area of quenching behaviour, molten pool formation, cooling in the lower
head and ex-vessel melt behaviour.
Four large-scale experimental facilities at FZK with a broad experience on severe accident
research are offered to external partners from EU member countries and associated states.
These facilities are QUENCH, LIVE, DISCO and COMET. Their overall purpose is to
investigate core melt scenarios from the beginning of core degradation to melt formation and
relocation in the vessel, possible melt dispersion to the reactor cavity, and finally corium
concrete interaction and corium coolability in the reactor cavity. These help in the
understanding of core degradation and quenching, melt formation and relocation as well as
melt coolability in real reactors in two ways – firstly directly by scaling-up and secondly
indirectly by providing data for the improvement and validation of computer codes.
Although the facilities can only perform experiments with simulant materials, the tests can


                                                   97
be considered as prototypic since the selected materials represent in important physical
properties the real core materials. The large masses used allow extrapolation to the reactor
case. Moreover, the flexibility and variability of the facilities is high due to the rather simple
handling. Pre-tests, parallel separate-effects tests and post-test analysis can be performed in
one hand. These tests can be seen as complementary to tests with UO2 in other research
centres.
Three calls for proposals (third call is optional) will be announced by FZK as provider of the
infrastructure during the 36 months period of the project, inviting interested users to specify
the experimental requirements and conditions. Following each call for proposals, the user
group selection panel will evaluate the proposals and recommend a short-list of user groups
that should benefit under this project. Careful consideration will be given to features not
already considered in experiments in this facility and will depend on the outcome of the
“user requirements” study. By the evaluation of the proposals, recommendations and results
of the EURSAFE project will be taken into account.
The overall objectives of the LACOMERA project are to offer research institutions from the
EC access to large scale experimental facilities which shall be used to increase the
knowledge of the quenching of a degraded core and regaining melt coolability in the Reactor
Pressure Vessel, of possible melt dispersion to the cavity, of Molten Core Concrete
Interaction and of ex-vessel melt coolability. One major aspect is to understand how these
events affect the safety of European reactors so as to lead to soundly based accident
management procedures. The project will bring together interested partners of different
European member states in the area of severe accident analysis and control, with the goal to
increase the public confidence in the use of nuclear energy. Moreover, partners from the
newly associated states will be included as far as possible, and therefore the needs of
Eastern, as well as Western, reactors will be considered in LACOMERA project. The project
offers a unique opportunity to get involved in the networks and activities supporting VVER
safety, and for Eastern experts to get an access to large-scale experimental facilities in a
Western research organisation to improve understanding of material properties and core
behaviour under severe accident conditions.




                                               98
Nuclear Energy Programme                                    Severe accident management
Operational safety of existing installations                Severe accident management

Title:
EUROPEAN GROUP FOR ANALYSIS OF CORIUM RECOVERY CONCEPTS

Acronym                 EUROCORE

Proposal number FIS5-1999-00050                     Contract number FIKS-CT1999-20003

Type of action          Thematic network            Duration           24 months

Starting date           1 March 2000                EC project officer A. Zurita

Total budget            385.404 €                   EC contribution    385.404 €


Co-ordinator
         Organisation         Commissariat a l'Energie Atomique (CEA)
                              DRN / DTP
         Address              17 rue des Martyrs
                              F-38054 Grenoble Cedex 9
         Contact person       Dr. Jean Marie Seiler
                              Tel:      33. 4 76.88.30.23
                              Fax:      33. 4 76.88.52.51
                              Email     seiler@dtp.cea.fr

Partnership

          Country                   Organisations

            UK                 Serco Assurance
            E                  Análisis y Soluciones Tecnológicas (AST)
            F                  Electricité de France (EDF)
            FIN                Fortum Nuclear Services Ltd
            F                  Framatome ANP
            D                  Forschungszentrum Karlsruhe GmbH (FZK)
            D                  Framatome ANP GmbH
            D                  Universität Stuttgart (IKE)
            S                  Royal Institute of Technology (KTH/RIT)

Project Summary
The role of the EUROCORE concerted action is to obtain a much clearer view of the state of
the art of European actual reactor corium recovery concepts and to better identify Research
and Development needs, taking into account current technical knowledge and reactor
situations. A prioritisation of R&D actions coming from the synthesis of real users needs
should result from this action which involves public and private research organisations, but



                                               99
also utilities and nuclear reactors vendors. In order to be successful, it will be paid a great
attention to the fact that debates are based on technical objectivity and a strong consensus
finally emerges, because a weak consensus makes no sense for handling such an important
problem.

To meet the action objectives, the following methodology is proposed:

First, the requirements attached to the considered retention concepts have, to be clearly
defined and relevant physical criteria for the concept to be efficient have to be derived, based
on the requirements. Second, analyses of reactor scenarios have to be conducted in order to
identify the generic situations of interest for the considered retention concepts. Realistic
initial and boundary conditions for these situations have to be defined. Third, these situations
have to be analysed with the actual technical knowledge, a synthesis has to be written,
including the identification of remaining unknowns and phenomena and the estimated
ranking of these phenomena considering their relative importance and causality. The final
step is to propose a set of research and development actions (modelling & experiments) with
associated priorities and a tentative timetable for issuing these actions in close connection
with the end users needs. The proposed work programme of the concerted action is
structured by the different existing or innovative reactor corium recovery concepts which are
or might be studied in Europe. Five different classes of concepts have been distinguished,
1eading to five different workpackages (WP2 to WP6). An additional workpackage is related
to the project management and the organisation of the workshops. For each technical
workpackage, three major tasks are conducted
- Collection and synthesis of most recent results
- Highlights over related open problems
- Recommendations for further research and development needs and ranking of the
priorities

The issues are:
• requirements attached to the corium considered retention concepts,
• analyses of reactor scenarios (Top-Down and Bottom-Up),
• a set of research and development actions (modelling & experiments),
• recommendations for calculation methodologies and physical models.

Measurable objectives of this project are:
• the production of consensus reports on the four dominant corium recovery concepts
• the production of consensus reports on alternative concepts
• the identification of further R&D needs for assessing, these concepts
• the identification of practical measures for severe accident management




                                              100
Nuclear Energy Programme                                    Severe accident management
Operational safety of existing installations                Severe accident management

Title:
STEAM GENERATOR TUBE RUPTURE SCENARIOS

Acronym                 SGTR

Proposal number FIS5-1999-00031                       Contract number FIKS-CT1999-00007

Type of action          Shared cost                   Duration           36 months

Starting date           1 January 2000                EC project officer A. Zurita

Total budget            2.260.000 €                   EC contribution    800.000 €


Co-ordinator
         Organisation          Technical Research Centre of Finland (VTT)
         Address               P.O. Box 1401 / Biologinkuja 7
                               FIN-02044 Espoo VTT
         Contact person        Dr. Jorma Jokiniemi
                               Tel:      358 9 456.61.58
                               Fax:      358 9 456.70.21
                               Email     jorma.jokiniemi@vtt.fi

Partnership

          Country                     Organisations

            E                   Centro de Investigaciones Energéticas, Medioambientales y
                                Tecnológicas (CIEMAT)
            FIN                 Fortum Nuclear Services Ltd
            NL                  Nuclear Research and Consultancy Group (NRG)
            CZ                  Nuclear Reseach Institute Rež plc (NRI)
            CH                  Paul Scherrer Institut (PSI)

Project Summary
The objective of this work is to generate a comprehensive database and to develop and verify
models to support accident management interventions in steam generator tube rupture
sequences leading to severe accident conditions. Experimental investigations in four
different facilities to study fission product retention in such scenarios before direct by-pass
release to the environment are planned. The current accident management actions foresee
flooding of the secondary side through the emergency feed water system in an attempt to
arrest the activity. Effective accident management actions may significantly reduce the
source term in these accident types. There is currently no appropriate database and associated
model estimating the source term from these types of accidents.



                                                101
The first task (WP1) includes definition of the most important steam generator tube rupture
accident sequences. This will be done by using the existing PSA studies for the PWR and
VVER-440 and by performing additional integral code calculations. The range of boundary
conditions for the experimental programmes will then be determined based on expected
conditions in the reference PWR and VVER plants.

In WP2 experiments will be conducted in a scaled-down version of steam generators
representing western PWRs and VVERs. An improved mechanistic understanding of the
local deposition will be achieved with an experimental programme to be carried out in
small-scale facilities. The separate-effect tests will be conducted in support of the integral
tests. This data is necessary for the model development and verification. Effect of different
secondary side flooding procedures (timing, flooding rate, etc.) will be investigated. The
relevant range of source term will be established. This will produce a reliable database for
PSA Level 3 evaluations.

In WP3 the experimental results will be applied to real steam generators by utilising system
codes, which will be equipped with the new developed and verified models. Models for
individual phenomena will be based on separate effect tests, which will then be implemented
in calculations of the integral experiments. Finally these models will be scaled up for full
size steam generators and expressed as mathematical correlations.

In WP4 the database and the analytical model will be used to assess the effectiveness of
different accident management procedures and to provide proposals for further
improvements. Important accident scenarios for the reference PWR and VVER-440 plants
will be analysed.

This project will provide experimental data and a validated model for fission product
transport in steam generator tube rupture (SGTR) sequences as well as the effect of different
accident management procedures to mitigate source terms in SGTR scenarios for western
PWR and VVER nuclear power plants.




                                             102
Nuclear Energy Programme                                    Severe accident management
Operational safety of existing installations                Severe accident management

Title:
IODINE CHEMISTRY AND MITIGATION METHODS

Acronym                 ICHEMM

Proposal number FIS5-1999-00038                       Contract number FIKS-CT1999-00008

Type of action          Shared cost                   Duration           36 months

Starting date           1 February 2000               EC project officer A. Zurita

Total budget            1.244.696 €                   EC contribution    548.707 €


Co-ordinator
         Organisation          AEA Technology plc
         Address               A32, Winfrith Technology Centre
                               UK-DT2 8DH Dorchester, Dorset
         Contact person        Dr. Shirley Dickinson
                               Tel:      44-1305-20 28 55
                               Fax:      44-1305-20 26 63
                               Email     shirley.dickinson@aeat.co.uk

Partnership

          Country                     Organisations

            F                   Institut de Radioprotection et de Sûreté Nucléaire (IRSN)
            INT                 European Commission - JRC/IE
            CZ                  Nuclear Reseach Institute Rež plc (NRI)
            CH                  Paul Scherrer Institut (PSI)
            D                   Framatome ANP GmbH
            S                   University Chalmers

Project Summary
Reliable models for the behaviour of iodine in a reactor containment following a severe
nuclear reactor accident are essential to the prediction of the potential release to the
environment, and thus to the development and qualification of appropriate mitigation
strategies and devices. Whilst most aspects of iodine chemistry are now adequately
understood, particularly for PWR conditions, some outstanding issues remain. Firstly, some
of the processes leading to the destruction of volatile forms of iodine are not well quantified.
An improved knowledge of these destruction rates will allow their importance to be assessed,
in terms of natural mitigation processes and accident management interventions. Secondly,
the effects on the iodine behaviour of certain materials and conditions, which are specific to



                                                103
BWR systems, are unknown. An understanding of these specific effects will allow data and
models developed mainly for PWR systems to be applied with confidence to BWR source
term predictions.

The proposed work programme comprises the following main elements.

i.   Provision of new kinetic data on volatile iodine destruction or transmutation reactions,
     which are not routinely included in severe accident iodine chemistry modelling codes.
     This will involve experimental measurements of the rate of molecular iodine destruction
     by ozone in the gas phase, and of the rate of methyl iodide destruction under irradiation
     in the gaseous and aqueous phases

ii. Investigation of other possible mitigation mechanisms or accident management measures
    to favour the conversion of volatile iodine species to non-volatile forms under severe
    accident conditions. This will involve experimental studies of the effects of candidate
    additive materials on iodine volatility from irradiated iodine solutions.

iii. Provision of experimental data on iodine behaviour under conditions specific to BWR
     containments under accident conditions, including the effect of reactive materials, and

iv. Quantification of the effects of the identified mitigation mechanisms on the predicted
    iodine source term for representative accident sequences. This will include the
    development of kinetic models based on the results of the experimental programmes,
    incorporation into severe accident modelling codes and evaluation of the impact on the
    calculated source term for some prototypical accident sequences.




                                             104
Nuclear Energy Programme                                      Severe accident management
Operational safety of existing installations                  Severe accident management

Title:
HYDROGEN REMOVAL FROM LWR CONTAINMENTS BY CATALYTIC COATED
THERMAL INSULATION ELEMENTS

Acronym                 THINCAT

Proposal number FIS5-1999-00029                       Contract number FIKS-CT1999-00006

Type of action          Shared cost                   Duration              28,5 months

Starting date           1 January 2000                EC project officer A. Zurita

Total budget*           1.200.000 €                   EC contribution* 600.000 €


Co-ordinator
         Organisation          Forschungszentrum Jülich (FZJ)
         Address               ISR - 2
                               D-52425 Jülich
         Contact person        Dr. Ernst-Arndt Reinecke
                               Tel:      49.2461.615530
                               Fax:      49.2461.614059
                               Email     e.-a.reinecke@fz-juelich.de

Partnership

          Country                     Organisations

            E                   Centro de Investigaciones Energéticas, Medioambientales y
                                Tecnológicas (CIEMAT)
            E                   Empresarios Agrupados
            D                   Kaefer Isoliertechnik GmbH & Co
            S                   Swedpower AB

Project Summary
An alternative concept for hydrogen mitigation in a LWR containment shall be developed,
based on catalytic coated thermal insulation elements of the main coolant loop components
instead of or in addition to backfitting passive autocatalytic recombiner devices. A first
estimate shows that there are sufficient insulation surfaces to achieve adequate
recombination rates when equipped with a catalytic coating. The project shall prove the
enhanced safety levels with respect to local high hydrogen concentrations, unintended
ignitions and recombination start-up delay. Economic advantages shall be demonstrated the
*
  Total eligible costs and EC contribution reduced respectively to 794.580 € and 355.370 € following
the changes in consortium composition and the duration of the project.


                                                105
respect to containment space obstructions, backfitting and licensing procedures. An
experimental database and suitable models to predict the hydrogen concentration transients
shall be developed. Safety and cost-benefit analyses shall be prepared to assess the economic
perspectives of the concept. The concept shall be developed to the level of commercial
usability.

The following tasks shall be performed to establish the catalytic thermal insulation concept:
-Catalytic coating of thermal insulation elements
Selection of materials and manufacturing process to generate catalytic surfaces on insulation
elements, with due consideration of backfitting existing insulations and coating new ones.
-Recombination efficiency experiments with coated insulation elements
In small-scale experiments with forced flow conditions, the hydrogen recombination rates for
selected coatings are measured. The influence of aerosols upon the rates is investigated. In
large-scale experiments with natural convection conditions, the overall hydrogen
recombination rate is determined for selected geometric element shapes. A 3-dimensional
code is used to simulate the experiments and evaluate the overall rate, using local rate
expressions from the small-scale data.
- Containment behaviour and thermal hydraulics analysis
A model to simulate the hydrogen distribution and recombination in the entire containment is
established, using recombination rate correlations derived from the large-scale experiments
and 3-d code analyses. The model is applied to a selected PWR for various accident transient
analyses.
- Local flow and heat transport processes near leaks
Numerical simulation of hydrogen jet release from a leak and jet contact with coated surface,
typical for lower containment rooms. Assessment of recombination start-up behaviour.
- Cost-benefit analysis
Evaluation of cost conditions for hydrogen management using coated insulation elements.
Comparison with hydrogen management by traditional recombiners, with due consideration
of operational and licensing aspects.

The research activities form an element of an integral hydrogen severe accident mitigation
strategy. This strategy aims to reduce most combustible hydrogen concentrations in the
containment by catalytic recombination, using the best available technology (coated thermal
insulators, possibly combined with passive recombiners), to minimise duration and levels of
enhanced concentrations. This will minimise the risk for fast deflagration. Since high
concentration levels cannot be fully excluded, the catalytic devices shall be designed such
that their probability for unintended ignition is low. Economic aspects of implementation,
maintenance and plant operation shall be taken into account.

Note:
After the bankruptcy of the previous co-ordinating organisation Battelle Ingenieurtechnik
GmbH, it was withdrawn from the project in October 2001. As Battelle was a project key
partner because of the planned integral experiments to be performed at the THAI facility, the
project could not achieve all initial objectives and was terminated untimely. As a
consequence some work packages were skipped and some activities were not performed.
This concerns especially the large-scale experiments and some of the small-scale tests, which
were intended to demonstrate the performance of the new concept under relevant conditions.




                                            106
Nuclear Energy Programme                                  Severe accident management
Operational safety of existing installations              Severe accident management

Title:
HYDROGEN HAZARD - PASSIVE AUTOCATALYTIC RECOMBINERS STATE-OF-THE-
ART

Acronym                 PARSOAR

Proposal number FIS5-1999-00030                     Contract number FIKS-CT1999-20002

Type of action          Thematic network            Duration          24 months

Starting date           1 February 2000             EC project officer G. Van Goethem

Total budget            150.000 €                   EC contribution   150.000 €


Co-ordinator
         Organisation         Technicatome

         Address              BP 34000 / 1100, Avenue Jean-René Guillibert Gautier De La
                              Lauzière
                              F-13791 Aix-en-Provence Cedex 03
         Contact person       Mr. François Arnould
                              Tel:     33 4 42 60 28 50
                              Fax:     33 442 60 25 11
                              Email    arnouldf@ta-aix.tecatom.fr

Partnership

          Country                   Organisations

            B                   Association Vinçotte Nuclear
            CA                  Atomic Energy of Canada Limited
            F                   Commissariat à l'Energie Atomique (CEA/DRN/DTP)
            CH                  Electrowatt Engineering Ltd
            D                   Framatome ANP GmbH

Project Summary
Environmental safety of nuclear power plants may be severely affected by uncontrolled
hydrogen-oxygen reactions in case of severe accidents. The introduction of passive
autocatalytic recombiners is an interesting method to remove together hydrogen and oxygen.
This thematic network consists of performing the state of the art in passive autocatalytic
recombiners, which seem nowadays to represent the best solution in mitigating the hydrogen
hazard. The main purposes of the work are:
 to expand knowledge about passive autocatalytic recombiners among the main users,



                                              107
   to build up a large synthesis on such devices in order to create an aid tool for users,
   to assess PAR applications in different fields (fission or fusion reactors, chemical
    industry).

This work will be carried out by a EU workgroup including the main actors in this area like
recombiner manufacturers, safety authorities, research experts, nuclear power plant designers
and utilities.

During the last ten years, several safety authorities have recognised recombiners to be an
efficient solution to reduce hydrogen hazard, and many papers have been published
concerning PARs, but no synthesis has been yet performed. So this thematic network aims at
answering five purposes:
 to build a large synthesis about the existing PARs in order to create an aid tool for users,
 to compare qualification tests with severe accident conditions and licensing procedures,
 to estimate hydrogen explosion hazard induced by passive autocatalytic recombiners,
 to evaluate the main numerical model needs for PAR designers and utilities,
 to assess possible PAR applications in order to develop new markets for manufacturers.

This thematic network progresses in four sections. The first one makes up a complete
presentation of each type of passive autocatalytic recombiner according to three topics that
are description, implementation and maintenance. The second one proposes an exhaustive
assessment of the current qualification processes. The third one attempts to determine the
main numerical model needs to approach recombiner behaviour during accidental situations.
The last one explores the potential new markets for PARs like small nuclear reactors, nuclear
waste storage, nuclear fuel transport in wet conditions, fusion reactors, and conventional
applications like chemical industry or hydrogen storage.

The present thematic network is further development of the hydrogen hazard global study,
which was done during the third Framework Programme FP-3 (1990-1994) by Pr.
FINESCHI’s workgroup (University of Pisa). The expected results of this thematic network
are:
 to perform a synthesis on PARs, which may become an aid tool for nuclear reactor
   utilities,
 to define qualification tests which would be necessary for being in accordance with SAC,
 to determine the main numerical model needs in order to improve numerical codes,
 to draw the attention on attractive industrial PAR applications in nuclear and conventional
   fields.




                                             108
Nuclear Energy Programme                                    Severe accident management
Operational safety of existing installations                Severe accident management

Title:
OPTIMISATION OF SEVERE ACCIDENT MANAGEMENT STRATEGIES FOR THE
CONTROL OF RADIOLOGICAL RELEASES

Acronym                 OPTSAM

Proposal number FIS5-1999-00074                       Contract number FIKS-CT1999-00013

Type of action          Shared cost                   Duration           24 months

Starting date           1 June 2000                   EC project officer A. Zurita

Total budget            1.011.260 €                   EC contribution    499.843 €


Co-ordinator
         Organisation          National Nuclear Corporation Limited (NNC)
         Address               Booths Hall, Chelford Road
                               UK-WA16 8QZ Knutsford, Cheshire
         Contact person        Dr. Ming Leang Ang
                               Tel:     44.15.65.84 37 89
                               Fax:     44.15.65.84 38 89
                               Email    ming.ang@nnc.co.uk

Partnership

          Country                     Organisations

            F                   Electricité de France (EDF)
            FIN                 Fortum Nuclear Services Ltd
            D                   Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH
            E                   Iberdrola
            CZ                  Nuclear Reseach Institute Rež plc (NRI)
            S                   Swedpower AB
            S                   Sycon Energikonsult
            B                   Tractebel S.A.
            HU                  Institute for Electric Power Research Co.(VEIKI)

Project Summary
This study has the following objectives:
1. To evaluate the impact of SAM strategies on the radiological behaviour and to examine
    the potential for the optimisation of their implementation to minimise any adverse
    radiological effects.




                                                109
2.   To establish a comprehensive technical basis for the definition of source terms for
     operating reactors across Europe.
3.   To define a set of realistic source terms for operating reactors, with provision made for
     the implementation of SAM measures and to compare them with the existing source
     term definitions.

This study is based on a broad spectrum of analysis involving 9 reactor designs that are
regarded as representative of the reactors operating in both current and prospective member
states of the European Union. The scope is summarised as follows:

WP1: The key core damage accident sequences, for the 11 reference plants, would be
identified and described. In addition to the accident sequences more usually involving
releases into the primary containment, emphasis is also placed on containment bypass
sequences.

WP2: The key issues to be addressed in the sensitivity analysis would be identified and
defined. This would be in two parts:
- Review of information: This would identify the major issues, both in terms of the impact
   on plant status and the associated radiological releases.
- Definition of the sensitivity study matrix.

WP3: For each of the key sequences baseline source terms would be defined, calculated
predominantly using integrated severe accident analysis codes. For each sensitivity study, the
predictions would be evaluated against the baseline values to assess the impact of the SAM
strategies on either the plant status or radiological releases. Issues concerning the operational
and design aspects would be examined.

WP4: The key findings of WP3 would be discussed and evaluated. This would be conducted
in two ways: in the context of risk reduction potential and recommendations on their optimal
implementation and operation.

Finally, an outline methodology would be developed to determine representative source
terms for operating reactors; as, for example, would be required as input to a Level 3 PSA.




                                              110
Nuclear Energy Programme                                    Severe accident management
Operational safety of existing installations                Severe accident management

Title:
A PERSPECTIVE ON               COMPUTERIZED         SEVERE     ACCIDENT     MANAGEMENT
OPERATOR SUPPORT

Acronym                 SAMOS

Proposal number FIS5-2001-00104                     Contract number FIKS-CT2001-20189

Type of action          Thematic network            Duration           18 months

Starting date           1 December 2001             EC project officer A. Zurita

Total budget            177.044 €                   EC contribution    107.160 €


Co-ordinator
         Organisation         Nuclear Service Corporation Netherlands (NSC)

         Address              Akenwerf 35
                              NL-2317DK Leiden
         Contact person       Dr. George Vayssier
                              Tel:     (31-71)523245
                              Fax:     (31-71)5232341
                              Email    vayssier@hetnet.nl

Partnership

          Country                   Organisations

            B                  Westinghouse Electric Europe
            SK                 Nuclear Regulatory Authority of the Slovak Republic
            E                  Tecnatom
            SI                 Krsko Nuclear Power Plant

Project Summary
In recent years, many NPPs have developed and implemented severe accident management
guidance (SAMG), which is aimed at prevention and mitigation of accidents involving core
degradation and core melt. A good overview of SAMG approaches in Europe and the USA
has been documented under the SAMIME Concerted Action.

In all these programmes, there is a set of severe accident management guidelines, which are
to be executed by qualified personnel. In many cases, this is a group of people within the
Emergency Response Organisation (ERO) and the group is subdivided in 'evaluators',
'decision makers' and 'implementers'. They are usually located in a separate location, called



                                              111
the Technical Support Centre. The 'evaluators' assess the situation at the plant, on the basis of
information which they receive from the control room operators, and they come up with
recommendations what to do to mitigate the accident. Decision is made on a higher level,
where also information from outside the plant is available, as planned actions may have
consequences offsite (e.g. releases). Finally, actions decided upon are mostly carried out by
licensed control room personnel, the 'implementers'.

The tools available at the TSC are the severe accident guidelines of the plant, plus
appropriate other tools (computational aids, simplified formulae), and sometime specific
guidance for the TSC (such as the Technical Support Guidelines for the TSC).
Computational Aids (CAs) are precalculated curves and graphs that supply quantitative
information which may be needed during the course of actions.

As limited equipment is available (either through the initiating event or as a consequence of
the severe accident), it is not clear whether this will be sufficient to mitigate the
consequences, or which failed equipment should be brought back to service with priority. So
all tools are paper tools, and all judgement is human judgement, based on incomplete or
invalid information and made under high stress conditions.

The situation may be improved by the use of computerised support to the TSC and the
operator. The proposed project envisages to investigate the possibilities of this approach, and
to indicate both the benefits and drawback of such advanced methods. Before embarking on
a larger project where this will actually be developed, a feasibility study will be done to
identify the optimum approach and developing the tools for that. The present project limits
itself to this study. The central tool to be used is the CAMS programme developed in the
OECD Halden Reactor Project. It is a further development of the work done by
Tecnatom/Iberinco at Cofrentes NPP (Spain) and Halden, and of work performed by
Tractebel (OPA-system) and others under the EC Reinforced Concerted Action on Reactor
Safety.




                                              112
Nuclear Energy Programme                                    Severe accident management
Operational safety of existing installations                Severe accident management

Title:
CONCERTED UTILITY REVIEW OF VVER-440 SAFETY RESEARCH NEEDS

Acronym                 VERSAFE

Proposal number FIS5-1999-00181                     Contract number FIKS-CT2000-20044

Type of action          Concerted action            Duration          24 months

Starting date           1 September 2000            EC project officer P. Manolatos

Total budget            186.000 €                   EC contribution   186.000 €


Co-ordinator
         Organisation         Fortum Nuclear Services Ltd
                              Nuclear Power
         Address              Rajatorpantie 8
                              FIN-00048 Vantaa
         Contact person       Dr. Harri Tuomisto
                              Tel:      (358-1045)32464
                              Fax:      (358-1045)92464
                              Email     harri.tuomisto@fortum.com

Partnership

          Country                   Organisations

            HU                  PAKS Nuclear Power Plant Ltd.
            CZ                  Czech Energy Company, Inc
            SK                  Slovenske Elektrarne, a.s.

Project Summary
The recent and current safety improvement programmes of the VVER-440 plants in Czech
Republic, Hungary and Slovakia have been successful in enhancing the level of management
of design basis accidents and thus bringing the prevention of severe accidents to high
standards. After demonstration of the effective accident prevention, the next level of the
defence-in-depth is to reduce the risks associated with severe accidents. It is the
responsibility of the plant owner or licensee to develop an overall approach to the Severe
Accident Management (SAM). Another current issue of the plant owners' safety management
is to develop the approach to maintaining the achieved safety level until to the end of
economically and technically justified lifetime of the plant.
The high-level objective of this Concerted Action is to create a network of the safety
managers and experts of the plant licensees that are foreseen to operate the VVER-440


                                              113
reactors within the European Union during the first decades of the 21st century. The aim is to
contribute the utility views to the preconditions to operate the VVER-440 reactors. For this
purpose, the specific features of the VVER's in the Central European countries in respect to
the already obtained high safety level will be taken into account.
The first task of the proposed Concerted Action is to have an overview on the comprehensive
approaches to two safety management areas that are of concern for the near future, i.e.
               1. Severe Accident Management (SAM)
               2. Plant Life Management (PLIM)
Existing results from the related national research projects and from the related EU
sponsored Phare projects will be reviewed and taken into account, when applicable.
The selection of the final approach for the individual plants has to be consistent with the
plant-specific features and with the national and utility requirements. However, the
harmonisation of the utility views is sought in order to obtain maximal benefits of the unified
approaches.
The primary objective of VERSAFE is to define what are the needs for the additional
information from the safety research, when developing a generic approach and the plant-
specific approaches to the SAM and PLIM issues. The role of national research institutes and
organisations of the partner countries is of crucial importance in performing such research in
order to create and maintain the expertise also on the national level. Thus, the project will
also account for the education and training needs in these specific areas. A further objective
is to enhance possibilities of well-defined research projects that are oriented to the needs of
the end-users, to be accepted into the Phare programme.
Common recommendations of the utilities are collected into the Final Report that is the main
result and deliverable of the Concerted Action. The Final Report will be written in such a
way that it can be utilised as a Handbook of Good Practices of SAM and PLIM. Thus, the
objectives of the Final Report are, in addition to outlining the outcome of discussions among
the partners in the common workshops, (1) to collect basic information needed for defining
commonly agreed methodologies to deal with SAM and PLIM, (2) to provide guidance for
the utilities in their decisions of the SAM and PLIM research needs for VVER-440/213
plants, (3) to discuss the results of the related Phare projects and their application to the
plants, and (4) to provide information that can be used to facilitate and support the
negotiations of the EU applicant countries operating VVER-440 reactors.




                                             114
Nuclear Energy Programme                                    Evolutionary concepts
Operational safety of existing installations                Evolutionary safety concepts

Title:
ADVANCED THREE-DIMENSIONAL TWO-PHASE FLOW SIMULATION TOOL FOR
APPLICATION TO REACTOR SAFETY

Acronym                 ASTAR

Proposal number FIS5-1999-00204                       Contract number FIKS-CT2000-00050

Type of action          Shared cost                   Duration          36 months

Starting date           1 September 2000              EC project officer G. Van Goethem

Total budget            1.888.691 €                   EC contribution   799.723 €


Co-ordinator
         Organisation          Commissariat à l'Energie Atomique (CEA)
                               Department of Mechanics and Technology
         Address               CEA Saclay, DMT / Sysco
                               F-91191 Gif-sur-Yvette Cedex
         Contact person        Dr. Henri Paillere
                               Tel:      (33-1) 69088409
                               Fax:      (33-1) 69089696
                               Email     henri.paillere@cea.fr

Partnership

          Country                     Organisations

            F                   Electricité de France (EDF)
            INT                 European Commission - JRC/IE Petten
            D                   Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH
            UK                  Manchester Metropolitan University
            B                   Von Karman Institute for Fluid Dynamics (VKI)
            CH                  Paul Scherrer Institute (PSI)

Project Summary
The objectives of the ASTAR project are to substantially enhance the three-dimensional two-
phase flow prediction capabilities of current Thermal-Hydraulic (TH) codes and to lay the
scientific and technical basis in terms of numerical schemes and modelling strategy for the
next generation of TH codes, while taking into account industrial requirements.

The basis for the development of improved two-phase flow simulation is the 3D two-fluid
model which can be coupled to additional transport equations for interfacial area and



                                                115
turbulent kinetic energy and dissipation rate. A well-designed experiment of confined bubble
plume will be performed in the framework of this project, providing a comprehensive 3D
field measurement data set for validation purposes. High resolution numerical schemes with
very low numerical diffusion compared to methods currently used in TH codes will be
further developed. Their improved accuracy will be assessed by extensive benchmarking on
safety-relevant two-phase flow test case problems as well as the newly performed
experiment, and by comparing with prediction capabilities of numerical methods found in
the current generation of codes.

The methodology of the project may be summarised as follows:
 identify the shortcomings of 3D two-phase flow models, closure laws and interface
   transfer processes, in line with the EUROFASTNET concerted action project (contract n°
   FIKS-CT2000-20100). Emphasis will be placed on the bubble flow regime, in a first
   step towards the simulation of more complex flows of relevance to reactor safety.
 define a common generic model on which the numerical developments will be based, and
   which will be improved during the course of the project by further physical modelling
   and experimental work.
 develop advanced numerical methods for 3D two-phase flow using high-order upwind
   schemes, unstructured grid formulations and efficient implicit time-integration schemes;
 test and validate the numerical developments on a pertinent set of test cases of industrial
   relevance, including the newly obtained experimental data.
 demonstrate the usability of the 3D module components by coupling to an existing
   system code.

The measurable objectives of this project are:
 improved 3D two-phase flow models and closure laws;
 experimental data set for confined bubble plume flow;
 improved numerical methods with low artificial diffusion, suitable for structured and
   unstructured meshes;
 development of simulation module components with flexible data structure for future
   coupling to existing system codes;
 demonstration of integration of new 3D models in existing system code;
   dissemination of the work.




                                            116
Nuclear Energy Programme                                    Evolutionary concepts
Operational safety of existing installations                Evolutionary safety concepts

Title:
TESTING AND ENHANCED MODELLING OF PASSIVE EVOLUTIONARY SYSTEMS
TECHNOLOGY (FOR CONTAINMENT COOLING)

Acronym                 TEMPEST

Proposal number FIS5-1999-00273/317                   Contract number FIKS-CT2000-00095

Type of action          Shared cost                   Duration          36 months

Starting date           1 December 2000               EC project officer G. Van Goethem

Total budget            3.238.785 €                   EC contribution   1.000.015 €


Co-ordinator
         Organisation          Nuclear Research and Consultancy Group (NRG)
                               Plant, Performance and Technology
         Address               Westerduinweg 3
                               NL-1755 ZG Petten
         Contact person        Mr. Victor A. Wichers
                               Tel:      (31-224)564656
                               Fax:      (31-224)563490
                               Email     wichers@nrg-nl.com

Partnership

          Country                     Organisations

            F                   Commissariat à l'Energie Atomique (CEA)
            CH                  Paul Scherrer Institute (PSI)
            FIN                 Technical Research Centre of Finland (VTT)
            D                   Gesellschaft für Anlagen - und Reaktorsicherheit (GRS) mbH
            D                   Forschungszentrum Karlsruhe GmbH (FZK)
            D                   Framatome ANP GmbH
            US                  University of California, Berkeley
            US                  General Electric Company

Project Summary
The primary objective of the TEMPEST project is to validate and improve advanced
modelling methods for evaluating pressure safety margins of the containment of Boiling
Water Reactors (BWRs). Accurate prediction of containment pressure transients during
severe accidents requires capabilities for modelling effects such as three-dimensional (3D)
mixing and stratification, since these strongly affect the performance of passive cooling



                                                117
systems. Modern Computational Fluid Dynamics (CFD) codes possess the desired 3D
modelling capabilities. In previous projects performed in the 4FWP (1994-1998), the
potential of these methods for detailed analysis of containment systems was shown, but also
the need for new experiments to validate CFD models became apparent. In the TEMPEST
project the applicability of CFD tools to model passive containments of evolutionary BWRs
will be further investigated.

The project has adopted two BWR reference designs currently under development for
Europe, the SWR1000 and the ESBWR. The passive decay heat removal systems applied in
these designs are respectively the Building Condenser (BC) and the Passive Containment
Cooling System (PCCS). The approach followed in the project is to use and generate
experimental data on the operation of passive containment cooling systems dedicated to CFD
model validation and to investigate in detail the performance of CFD containment models.

The experimental data will be obtained from new integral (containment) experiments of the
PCCS to be performed in the PANDA facility and from earlier BC experiments in the
PANDA facility. The new test series will be focussed on investigating the distribution of the
non-condensable gases inside the containment, their effect on the effectiveness of the passive
containment cooling systems, and on improvements of these systems. The integral PANDA
tests will cover all physics dominating passive containments, i.e. mixing and stratification,
steam condensation in the presence of non-condensables and boiling. Therefore the integral
system tests will be complemented by separate effect tests of mixing of buoyant steam-
hydrogen plumes in a gas atmosphere in the KALI facility.

The modelling methods to be validated fall into two broad categories: CFD codes and
advanced system codes. These two categories differ in degree of modelling detail,
complexity of use and computational efficiency. Since in practice a trade-off must be made
between these aspects, both categories have representatives in the project: commercial CFD
codes (CFX, FLUENT, STAR), dedicated CFD codes (GASFLOW, TONUS, GOTHIC),
advanced system codes (WAVCO, SPECTRA), as well as a coupled approach (CFX-
WAVCO). Code assessment will result both in model improvement, validated models,
guidelines on best practice as well as recommendations for model improvement.

The improved modelling methods will be used in ESBWR and SWR1000 plant evaluations
in order to assess the potential reduction of design pressure due to improved modelling. The
generic results of this project are applicable to all pressure-suppression type BWRs with
either active decay heat removal systems (operating plants) or with passive decay heat
removal systems (future plants).




                                             118
Nuclear Energy Programme                                    Evolutionary concepts
Operational safety of existing installations                Evolutionary safety concepts

Title:
EVALUATION OF COMPUTATIONAL FLUID DYNAMIC METHODS FOR REACTOR
SAFETY ANALYSES

Acronym                 ECORA

Proposal number FIS5-2001-00051/67                    Contract number FIKS-CT2001-00154

Type of action          Shared cost                   Duration          36 months

Starting date           1 October 2001                EC project officer G. Van Goethem

Total budget            1.623.803 €                   EC contribution   753.480 €


Co-ordinator
         Organisation          Gesellschaft für Anlagen - und Reaktorsicherheit (GRS) GmbH
                               Thermalhydraulics and Process Engineering
         Address               PO Box 1328
                               Forschungsgelände
                               D-85739 Garching
         Contact person        Mrs. Martina Scheuerer
                               Tel:     (49-89)32004401
                               Fax:     (49-89)32004599
                               Email    bam@grs.de

Partnership

          Country                     Organisations

            DE                  AEA Technology GmbH
            HU                  KFKI Atomic Energy Research Institute (AEKI)
            F                   Commissariat à l'Energie Atomique (CEA)
            F                   Electricité de France (EDF)
            D                   Forschungszentrum Rossendorf e.V. (FZR)
            NL                  Nuclear Research and Consultancy Group (NRG)
            CZ                  Nuclear Research Institute Rež plc (NRI)
            CH                  Paul Scherrer Institute (PSI)
            S                   Vattenfall Utveckling AB
            FIN                 Technical Research Centre of Finland (VTT)
            UK                  Serco Assurance/Serco Ltd




                                                119
Project Summary
The overall objective of the present project is to evaluate the capabilities of Computational
Fluid Dynamic (CFD) software packages in relation to simulating flows in the primary
system and containment of nuclear reactors. The interest in the application of CFD methods
arises from the importance of three-dimensional effects in these flows which can not be
predicted by traditional one-dimensional system codes. Perspective areas of the application
of detailed three-dimensional CFD calculations will be identified and recommendations for
code improvements will be provided which are necessary for comprehensive simulations of
safety-relevant accident scenarios for future research. In the ECORA project the experience
of twelve partners is combined from European industry and research organisations in the
field of nuclear safety applying the CFD codes CFX, Fluent, Saturne, STAR-CD and Trio-U.

The assessment will include the establishment of Best Practice Guidelines and standards
regarding the use of CFD software and evaluation of results for safety analysis. CFD quality
criteria will be standardised prior to the application of different CFD software, and results
will only be accepted when the set quality criteria are satisfied. Thus, a general basis will be
formed for assessing merits and weaknesses of particular models and codes on a European
basis. CFD simulations achieving an accepted quality level will increase confidence in the
application of CFD-tools.

Furthermore, a comprehensive and systematic software engineering approach for extending
and customising CFD codes for nuclear safety analyses will be formulated and applied. The
adaptation of CFD software for nuclear reactor flow simulations will be shown by
implementing enhanced two-phase flow, turbulence, and energy transfer models relevant for
pressurised thermal shock (PTS) applications into CFX, Saturne and Trio_U. An analysis of
selected UPTF and PANDA experiments will be performed to validate CFD software in
relation to PTS phenomena in the primary system and severe accident management in the
containment.

The methodology of the project can be summarised as follows:
 Best Practice Guidelines for the use of CFD software and for the formalised judgement
   of CFD results and experimental data will be established.
 CFD simulations of three-dimensional flows in LWR primary systems and containments
   will be assessed.
 Quality controlled CFD simulations for selected UPTF and PANDA test cases will be
   performed.
 CFD code customisation and improvement will be demonstrated for PTS relevant
   applications.

The expected outcome of the project will be a comprehensive evaluation of CFD software
for nuclear reactor safety applications, 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 CFD codes in nuclear safety
which will be constituted at the end of the project in a workshop. The goal of the network
will be to establish, maintain and extend the Best Practice Guidelines, and to collaborate on
an European level in transforming the defined CFD requirements into software solutions.




                                              120
Nuclear Energy Programme                                  Evolutionary concepts
Operational safety of existing installations              Evolutionary safety concepts

Title:
EUROPEAN GROUP FOR FUTURE ADVANCES IN SCIENCES AND TECHNOLOGY
FOR NUCLEAR ENGINEERING THERMALHYDRAULICS

Acronym                 EUROFASTNET

Proposal number FIS5-1999-00324                     Contract number FIKS-CT2000-20100

Type of action          Concerted action            Duration           18 months

Starting date           1 September 2000            EC project officer G. Van Goethem

Total budget            318.725 €                   EC contribution    199.988 €


Co-ordinator
         Organisation         Commissariat à l'Énergie Atomique (CEA)
                              Departement de Thermohydraulique et Physique
         Address              17 Rue des Martyrs
                              F-38054 Grenoble Cedex 9
         Contact person       Dr. Dominique Bestion
                              Tel:     (33-4)76883645
                              Fax:     (33-4)76885179
                              Email    dominique.bestion@cea.fr

Partnership

          Country                   Organisations

            CH                  Paul Scherrer Institute (PSI)
            I                   Università degli Studi di Pisa
            D                   Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH
            F                   Electricité de France (EDF)
            FIN                 Lappeenranta University of Technology
            SI                  Institut "Jozef Stefan"
            UK                  AEA Technology
            UK                  The Imperial College of Science, Technology and Medicine
                                (ICSTM)
            INT                 European Commission - JRC/IE Petten
            CZ                  Nuclear Research Institute (NRI) Rež plc
            I                   Ente per le Nuove Tecnologie, l'Energia e l'Ambiente (ENEA)
            F                   Framatome ANP S.A.
            F                   Institut de Radioprotection et de Sûreté Nucléaire (IRSN)




                                              121
Project Summary
EUROFASTNET means European project for Future Advances in Sciences and Technology
for Nuclear Engineering Thermal-Hydraulics. The point of view of the industry, the utilities,
the R&D institutes, the safety institutes, and research laboratories of Universities will be
associated. This Concerted Action should result in a common understanding on what are the
key problems and which R&D work should be initiated to solve them. It should give rise to
an enhanced European co-operation for present and future development work.

The objective of this Concerted Action is clearly to identify a set of R&D actions in thermal-
hydraulic, which can put the European nuclear industry in the best position to meet industrial
challenges of tomorrow. A generic need appears to develop new thermal-hydraulic
computer models, and new experiments to have a more detailed and better qualified
description of the physics.

This proposal of R and D actions in thermal-hydraulics will contribute to put in common the
efforts, which are currently spent, in several European countries in a dispersed way and it
will help to better define in a consensual manner the R&D priorities.

In order to identify the factors of progress from the industrial point of view, a review is made
of the R&D needs in Thermal hydraulics for all types of European reactors in operation
(western PWR, BWR and WWER), as well as for innovative reactors designs, which are
under study. This review is to be fully assessed by vendors, electricity producers and safety
authorities. It should address problems related to reactor performance, availability, reactor
life span, and problems associated with reactor safety, code validation, uncertainty
evaluation.

Then a state of the art and an analysis of the limitations of present available tools (numerics
as well as instrumentation) for thermal-hydraulics seen from the R&D side will be made.
The industrial problems are analysed and they are confronted with present numerical or
physical model limitations and instrumentation capabilities.

In a third step, an R&D programme is elaborated to answer the industrial problems. The
ways to pass beyond the limitations of presently available design and safety analysis tools
will be identified. This R&D programme will address the physical modelling, the numerical
improvements, the experiments and the uncertainty evaluation




                                              122
Nuclear Energy Programme                                    Evolutionary concepts
Operational safety of existing installations                Evolutionary safety concepts

Title:
REVISITING CRITICAL ISSUES IN NUCLEAR REACTOR DESIGN / SAFETY BY USING
3-D NEUTRONICS / THERMALHYDRAULICS MODELS: STATE-OF-THE-ART

Acronym                 CRISSUE-S

Proposal number FIS5-2001-00099                       Contract number FIKS-CT2001-00185

Type of action          Shared cost                   Duration          24 months

Starting date           1 January 2002                EC project officer G. Van Goethem

Total budget            288.489 €                     EC contribution   150.000 €


Co-ordinator
         Organisation          University of Pisa
                               Dept. of Mechanical, Nuclear & Production Engineering
         Address               Via Diotisalvi
                               I-56100 Pisa
         Contact person        Prof. Francesco d'Auria
                               Tel:      (39-050)836653
                               Fax:      (39-050)836665
                               Email     dauria@ing.unipi.it

Partnership

          Country                     Organisations

            E                   Asociacion Nuclear Asco Vandellos
            S                   Studsvik Eco & Safety AB
            S                   Swedish Nuclear Power Inspectorate (SKI)
            E                   Universidad Politécnica de Madrid (UPM)
            E                   Universidad Politécnica de Valencia
            CZ                  Nuclear Research Institute Rež plc (NRI)

Project Summary
The CRISSUE-S project deals with the techniques of applying coupled 3-D neutronics-
thermalhydraulics models to the technology of LWR. The design and the safety evaluation
of Nuclear Power Plants are concerned. The recent availability of powerful techniques and
of suitable computational resources opens new horizons in the technology. Advanced safety
evaluations and design optimisations can be performed that were not possible a few years
ago.




                                                123
The lack of immediate industrial interest, owing to the stop in the construction of nuclear
units in the majority of Western Countries, and the natural caution from the regulatory
bodies in accepting innovations, prevented so far the wide exploitation of the techniques here
concerned. This justifies the proposed project that aims at establishing the current state of
the art in the area and at formulating recommendations to the users of the nuclear
technology. In wider terms, the strategic objective of the project is to show the advantages in
using the considered techniques and to promote their diffusion. The use of plutonium and of
MOX fuel, the use of fuel elements of different types within the same core and the practices
of current interest of increasing the average burn-up and the core thermal power constitute
further reasons for the application and the diffusion of these techniques.

The evaluation of results from the analysis of critical transients that have challenged nuclear
engineers and researchers in the last decades constitutes the mean to achieve the mentioned
objectives. Off-normal conditions that have the potential to increase neutron power are of
interest and can be characterised by the term RIA (Reactivity Initiated Accident). For the
PWR class, emphasis is given to transients originated by the Main Steam Line Break that
may be the source of localised power rises of the core. For the BWR class, emphasis is
given to transients originated by the closure of the Main Steam Isolation Valve (and to
similar transients, e.g. originated by turbine trip) and to those originated by instability.
Anticipated transients without scram (ATWS) are of interest in both cases.

Relevant international activities recently completed or in progress and critical evaluation of
the related results are at the centre of the attention in the proposed project. The group of
partners of CRISSUE-S includes all actors in the nuclear technology, i.e., utilities, designers,
licensing and research institutions. In addition, the presence among the partners of the
International Institution OECD and of US institutions that are directly linked with the US
NRC (namely: Pennsylvania State University and University of Illinois at Urbana
Champaign, should be emphasised. The presence of one partner from the former Eastern
Countries and the connection with the VALCO project (see below) ensure proper
consideration of the WWER technology. The approval by the EC of a project, named
VALCO having objectives ‘complementary’ to those of CRISSUE-S confirms the
importance of the subject within the current technology. Tight connections have been
established between CRISSUE-S and VALCO.

The eight areas listed below, part of the ‘state-of-the-art-report’ (SOAR) at the centre of the
CRISSUE-S project, give an idea of the main features of the project and of the cross
connections with other disciplines and parallel techniques.
1) Probabilistic Safety Assessment,
2) System thermalhydraulics,
3) 3-D neutronics,
4) Fuel behaviour (fundamentals),
5) Achievement of high Burn-up and use of High Burn-up fuel,
6) Exploitation of Plutonium,
7) Operator training and control room design (including Emergency Operating Procedures),
8) Regulatory requirements (current and future) and actual safety margins (including relevant
statement about uncertainty)

The result expected from the project is the evaluation of the safety status of the current LWR
as resulting from the application of the considered techniques. Recommendations to utilities
and to regulators for the most fruitful use of those techniques are the final outcome of the


                                              124
activity. Improvements in the design and the use of Engineered Safety Features and of
Emergency Operating Procedures can be envisaged based on the achieved results. Areas of
the NPP design can be identified where the design/safety requirements can be relaxed owing
to improved knowledge in the area of neutronics/thermalhydraulics coupling.




                                           125
Nuclear Energy Programme                                    Evolutionary concepts
Operational safety of existing installations                Evolutionary safety concepts

Title:
VALIDATION OF COUPLED NEUTRONICS/THERMAL HYDRAULICS CODES FOR
VVER REACTORS

Acronym                 VALCO

Proposal number FIS5-2001-00070                       Contract number FIKS-CT2001-00166

Type of action          Shared cost                   Duration          24 months

Starting date           1 January 2002                EC project officer G. Van Goethem

Total budget            1.092.045 €                   EC contribution   672.902 €


Co-ordinator
         Organisation          Forschungszentrum Rossendorf E.V. (FZR)
                               Institute of Safety Research
         Address               D-01314 Dresden
         Contact person        Prof. Frank-Peter Weiss
                               Tel:      (49-351)2603480
                               Fax:      (49-351)2603440
                               Email     F.P.Weiss@fz-rossendorf.de

Partnership

          Country                     Organisations

            D                   Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH
            FIN                 Technical Research Centre of Finland (VTT)
            HU                  KFKI Atomic Energy Research Institute (AEKI)
            CZ                  Nuclear Research Institute Rež plc (NRI)
            SK                  Nuclear Power Plant Research Institute (VUJE) Trnava Inc
            BG                  Institute of Nuclear Research and Nuclear Energy
            UA                  State Scientific and Technical Centre on Nuclear and Radiation
                                Safety/KIEV
            SK                  Slovenske Elektrarne, a.s. (EBO), BOHUNICE
            SK                  Slovenske Elektrarne, a.s. (EBO), MOHOVCE
            RU                  Russian Research Centre 'Kurchatov Institute '

Project Summary
The project is aimed at the improvement of the validation of coupled neutronics/thermal
hydraulic codes for VVERs. Modern safety standards require a modelling of complex
accident processes with significant interaction between thermal hydraulic system behaviour



                                                126
and space-dependent reactor kinetics. To perform the analysis of such events, thermal
hydraulic system codes have to be coupled to three-dimensional core models. These coupled
codes need to be validated against well specified transient scenarios. The implementation of
advanced codes for accident analysis in associated states and NIS will represent a
contribution to safety evaluation of NPPs in central and eastern European countries who will
be member of EU in nearest future.
The work is based on results obtained within the EU Phare project "Improvement of the
verification of coupled thermal hydraulics/neutron kinetics codes" (SRR1/95).
The first objective is to extend the measurement data base for the validation of coupled
neutronics / thermal hydraulic codes for VVER type reactors covering processes which were
not considered in previous analyses. Based on the experience obtained in the frame of the
Phare project the measurement data base for validation of coupled codes will be extended
and qualified. While the transients analysed in the Phare SRR1/95 project were initiated by
perturbations in the secondary circuit, transients caused by hardware actions in the primary
circuit are of special interest to this project. Two data sets, one for each VVER-440 and
VVER-1000, will be selected for transient calculation. The analysis of these transients will
be accomplished with different available code systems (e.g. DYN3D-ATHLET, KIKO3D-
ATHLET, BIPR8-ATHLET, HEXTRAN-SMABRE). The calculated results will be
compared with measurement values.
The second objective is to develop a new methodology of uncertainty analysis for coupled
codes and to apply it to selected transients. Up to now uncertainty analyses were performed
only for thermal hydraulic code systems. The application of methods which were developed
for thermal hydraulic codes will be extended to coupled codes. The uncertainty analysis
methodology will be applied to transients considered within the above mentioned Phare
project. Uncertainty bands of relevant output parameters of the codes will be obtained and
compared with the results of previous analyses. On basis of this comparison weak points of
validation procedure are identified.
The third objective is to validate neutron kinetics models and nuclear cross section libraries
which are used in the different coupled code systems against kinetics experiments. The
selection and calculation of two kinetics experiments performed at V-1000 zero power test
facility in the Kurchatov Institute Moscow will be realised. The detailed experimental data
and macroscopic cross section data will be made available. The experiments will be
modelled with the three-dimensional neutron kinetics codes DYN3D, HEXTRAN and
KIKO3D.
In the framework of the VALCO project a tight co-operation will be realised with the
participants of CRISSUE-S project (“Revisiting Critical Issues in Nuclear Reactors
Design/Safety by using 3-D Neutronics/Thermalhydraulics Models: State-of-the-Art”).
As a result of the project, various coupled neutronics/thermal hydraulic code systems will be
qualified for the application to transient analyses for VVER type reactors. The results of the
Phare project will be completed by systematically extending the validation base including
additional neutron kinetics experiments without thermal hydraulic feedback. Clarifying
reasons for deviations between measurements and calculations, directions of further code
improvements will be shown. A methodology for the uncertainty assessment of coupled
codes will be developed and used to quantify the uncertainties of safety relevant parameters.




                                             127
Nuclear Energy Programme                                                Evolutionary concepts
Operational safety of existing installations                            Evolutionary safety concepts

Title:
RELIABILITY METHODS FOR PASSIVE SAFETY FUNCTIONS

Acronym                  RMPS

Proposal number FIS5-1999-00250                                Contract number FIKS-CT2000-00073

Type of action           Shared cost                           Duration                     36 months

Starting date            1 February 2001                       EC project officer S. Casalta

Total budget             944.581 € *                           EC contribution              550.000 € *


Co-ordinator
         Organisation              Commissariat à l'Énergie Atomique (CEA)
                                   DRN/DER/SSAE/LAER
         Address                   C.E. Cadarache - Building 238
                                   F-13108 Saint-Paul-lez-Durance
         Contact person            Mr. Flavio De Magistris
                                   Tel:      (33)442256336
                                   Fax:      (33)442252408
                                   Email     flavio.magistris@cea.fr

Partnership

          Country                       Organisations

             I          Consorzio Interuniversitario per la Ricerca Tecnologica
                        Nucleare (CIRTEN)
        I               Ente per le Nuove Tecnologie, l'Energia e l'Ambiente (ENEA)
        D               Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH
        INT             European Commission - JRC/IE
        F               Technicatome
        BG              Technical University of Sofia (TUSO)*
_____________________________
* Budget expanded following inclusion of new partners from Newly Associated States (NAS).


Project Summary
The functions of the passive B systems are based on thermal-hydraulic (T-H) principles,
which are not currently considered to be subject to any kind of failure. But due to the
environment and to the physical phenomena that may deviate from expectation, the passive
system may fail to meet its required passive function. The quantification of the T-H
unreliability is often still a difficult process due to the numerous uncertainties.



                                                        128
The objective of the project is to propose a specific methodology to assess the passive
system thermal-hydraulic reliability. This methodology will answer the following questions:
      a) How to identify and evaluate the sources of uncertainties and how to determine the
important variables, i.e. those variables whose uncertainty has a significant impact on the T-
H performance of passive systems,
      b) How to propagate efficiently the uncertainties through T-H codes and how to assess
the unreliability of the T-H passive system,
      c) How to link in an accident sequence, the passive system T-H unreliability with
others unreliability (failure of active systems, human errors...) in order to evaluate the
influence of the passive system on the sequence?

The methodology will be tested on an example of industrial T-H passive system. The T-H
calculations will be performed using the RELAP5 and ATHLET computer codes.

In order to address these questions the project is structured in three main work packages:
a)    Identification and quantification of the sources of uncertainties and determination
of the important variables: In this task, a systematic methodology will be defined from a
state-of-the-art:
 To ensure that uncertainties associated with the T-H performance of passive systems in
    code input variables as well as in software correlation and models are considered.
    Particular attention will be paid in selecting the range of uncertainty and the probability
    density function for these variables. The influence of the choice of the distribution on the
    model response will be assessed
 To rank and to quantify the relative contribution of each uncertain parameter on the whole
    response uncertainty. A special attention will be paid on non-linear sensitivity analysis.
The different methods will be tested and compared on the industrial study case.

b) Propagation of the uncertainties through a T-H model and reliability assessment
of the T-H passive system. The following methods for reducing the number of T-H
calculations will be tested and compared:
 variance reduction techniques in Monte-Carlo simulation: Latin Hypercube, Importance
    Sampling, Directional Simulation…
 response surface techniques: polynomial surfaces, non-linear response surface obtained
    by neural networks...
Improvement specific to the problems of T-H systems will be realised.
        The First and Second Order Reliability Methods (FORM/SORM) used in structural
mechanics to evaluate the reliability of components and structures will be analysed from the
point of view of their application to the T-H passive systems. Specific coupling scheme
between a T-H code and the FORM/SORM algorithm will be developed.
c)    Introduction of passive system unreliability in the accident sequence analysis. This
task is a state-of-the-art on the approaches made to incorporate physical uncertainties in
conventional PSA. In particular, hybrid approaches where the stochastic phenomena are
treated as junction in the event tree and the subjective probabilities reflecting the lack of
knowledge are treated by the way of Monte-Carlo simulations, will be tested and analysed on
a simplified event tree in relationship with the concerned passive system. A methodology
will be proposed.




                                              129
Nuclear Energy Programme                                    Evolutionary concepts
Operational safety of existing installations                Evolutionary safety concepts

Title:
NATURAL CIRCULATION AND STABILITY PERFORMANCE OF BWRS

Acronym                 NACUSP

Proposal number FIS5-1999-00175                       Contract number FIKS-CT2000-00041

Type of action          Shared cost                   Duration           48 months

Starting date           1 December 2000               EC project officer S. Casalta

Total budget            2.973.033 €                   EC contribution    1.197.508 €


Co-ordinator
         Organisation          Nuclear Research and Consultancy Group (NRG)
                               Plant Performance & Technology
         Address               Westerduinweg 3
                               NL-1755 ZG Petten
         Contact person        Dr. Kees Ketelaar
                               Tel:     (31-224) 564342
                               Fax:     (31-224) 563490
                               Email    ketelaar@nrg-nl.com

Partnership

          Country                     Organisations

            F                   Commissariat à l'Energie Atomique (CEA)
            NL                  Delft University of Technology
            CH                  Swiss Federal Institute of Technology Zuerich
            S                   Forsmarks Kraftgrupp AB
            D                   Forschungszentrum Rossendorf e.V. (FZR)
            E                   Iberdrola, S.A.
            CH                  Paul Scherrer Institut (PSI)
            E                   Universidad Politecnica de Valencia

Project Summary
The goal of this project is to improve the economics of operating and future plants through
improved operational flexibility, enhanced availability, and increased confidence level on the
safety margins regarding the stability issues in Boiling Water Reactors (BWRs).
The next generation reactors typically use large-sized cores to produce higher power output
as one of the key measures to reduce cost. It is well established that as the core size
increases, nuclear coupling between different parts of the core becomes weaker, and the core


                                                130
becomes more susceptible to out-of-phase (regional) oscillations. This issue presents a
constraint on the power level, core design, and a possible limitation on the maximum core
size that is feasible for the future plants.
For the operating BWR plants, reactor operators use their own approach as to the stability
issue. Some try to avoid unstable operational conditions based on numerical predictions.
Others try to detect and suppress power and flow oscillations before exceeding specified
acceptable fuel design limits. Most utilities use a combination of these two strategies. These
solutions often use conservative inputs and assumptions to account for uncertainties in
models and confidence in the methodology, leading to conservative exclusion region in
which operation will be precluded. Moreover, it is well known that the unstable operational
regime changes with fuel burnup and depends on the power profile.
Development of a common European approach is needed here, which will be one of the
outcomes of this project. Another measure to reduce cost is the use of natural circulation in
the next generation reactor designs. These designs eliminate the need for circulation pumps
and the associated piping and systems. In addition to simplification and cost saving, the
natural-circulation reactor together with the passive safety systems has demonstrated the
potential to realise the inherently safe concept in reactor design. The experience with natural-
circulation reactors is limited to small reactor cores with modest powers. One concern on this
type of reactor design, besides the large-sized core, is the start-up process. Some experiments
and analyses have indicated that hydraulic oscillations can occur at low-pressure and low-
power operating conditions. Two-phase natural circulation is a key in most of the advanced
light water reactor designs, which use automatic depressurisation systems to change the
operating state of the reactor from high power/pressure to low power/pressure conditions.
These concepts depend on natural circulation flow for the long-term cooling of the core.
This project addresses the above stability issues by expanding the basic understanding
through well structured testing and analyses of experimental data as well as analyses of
existing operational stability data of 3 different European reactors (Forsmark, Cofrentes,
Leibstadt); by applying this knowledge via efficient models and validated computer codes to
operating reactors and reactor designs; and by developing general guidelines for reactor
operation and design on how to avoid reactor instabilities.
The involved experimental facilities (DESIRE, CIRCUS, CLOTAIRE, PANDA) range from
low-power and low-pressure to nominal power and nominal pressure, and from small-scale
to large-scale, thereby covering the whole relevant range of possible natural-circulation
operating points. These data will be used to verify and improve the capability of existing
tools (analytical methods and general-purpose transient thermohydraulic computer codes),
and to develop an efficient frequency-domain tool, over a broad range of flow, power, and
pressure natural circulation conditions. Improved codes and validated tools can lead to better
defined operating procedures and margins, and as a result, to a more economic core design.
The tasks proposed in this project will enhance the basic understanding on stability issues,
generate new experimental data, provide guidelines, develop efficient models and validate
computer codes. The majority of the results is applicable to both forced and natural-
circulation cooled BWRs. The emphasis is concentrated on the thermal-hydraulic aspects of
the stability features. The results of this project will improve operational flexibility, increase
confidence level on the safety margins, and consequently, increase the overall economics of
the operating BWRs and future designs. In order to guarantee proper dissemination of the
results, the end-users, being BWR utilities, are well represented within the consortium.




                                               131
Nuclear Energy Programme                                    Evolutionary concepts
Operational safety of existing installations                Evolutionary safety concepts

Title:
DESIGN AND DEVELOPMENT OF A STEAM GENERATOR EMERGENCY
FEEDWATER PASSIVE SYSTEM FOR EXISTING AND FUTURE PWR'S USING
ADVANCED STEAM INJECTORS

Acronym                 DEEPSSI

Proposal number FIS5-1999-00262                       Contract number FIKS-CT2000-00113

Type of action          Shared cost                   Duration           36 months

Starting date           1 December 2000               EC project officer S. Casalta

Total budget            1.568.840 €                   EC contribution    700.000 €


Co-ordinator
         Organisation          Commissariat à l'Énergie Atomique (CEA)
                               DRN / DER / SERSI
         Address               CEA Cadarache - Building 212
                               F-13108 Saint-Paul-lez-Durance Cedex
         Contact person        Dr. Patrick Dumaz
                               Tel:       (33) 4 42 25 40 98
                               Fax:       (33) 4 42 25 40 46
                               Email      patrick.dumaz@cea.fr

Partnership
          Country                     Organisations

            CZ                  Nuclear Research Institute Rež plc (NRI)
            I                   Centro Electrotecnico Sperimentales Italiano (CESI)
            I                   Ente per le Nuove Tecnologie, l'Energia e l'Ambiente (ENEA)
            I                   Società Italiana Esperienze Termoidrauliche (SIET) S.p.A.
            PL                  Polska Akademia Nauk (IMP-PAN)

Project Summary
The use of passive systems to remove decay heat in advanced light water reactors is one way
to improve the safety of nuclear systems. Among these systems, the steam injectors (often
called “condensing ejectors” or “steam jet pumps”) appear to have promising capabilities
regarding their operating principle: i.e. to expand pressurised steam through a converging-
diverging nozzle in order to suck low pressure cold water and to pressurise it through a
second converging-diverging nozzle. Since the end of the fifties, many studies were
undertaken to extend the operating range of steam injectors widely used in the past in



                                                132
conventional steam engines. Although, significant progress was made, it was not enough to
make possible and attractive the steam injector utilisation in nuclear reactors.

The DEEPSSI project proposes to attain an innovative high pressure steam injector design
and the required reactor implication evaluation using a qualified computational model. The
reference reactor application is the Steam Generator Emergency FeedWater System (EFWS)
of Pressurised Water Reactors (PWR’s). Both western and eastern types PWR’s
(WWER440) will be considered.

Experimental and theoretical studies will be conducted. The experimental studies will
address the development of an innovative steam injector design and to provide enough data
to constitute a significant data base allowing the elaboration of dedicated steam injector
correlations and the validation of a computational model. Three experimental facilities will
be involved:
- a small scale facility located in Poland (IMP-PAN) which can provide steam up to 0.5
    MPa and 0.042 kg/s. The power scaling factor will be about 8 and the tests conducted
    will be mainly devoted to the understanding of some basic thermalhydraulic phenomena
    (direct contact condensation, two-phase shock wave).
- an industrial scale facility, CLAUDIA, located in France (CEA-Cadarache) which can
    provide steam up to 3 MPa and 11 kg/s. Here, the power scaling factor will be about one.
    A first test series will be devoted to the steam injector design using an existing test
    section slightly modified. A second test series using advanced two-phase flow
    instrumentation will deliver experimental data at a realistic scale required by modelling.
- a second industrial scale facility, IETI, located in Italy (SIET) which can provide steam
    up to 9 MPa and 5.5 kg/s. Here, the steam injector design obtained will be assessed at
    high pressure (component tests) and then in a realistic system configuration (system
    tests) using a new fabricated test section.

The theoretical studies will be undertaken in the frame of the CATHARE thermalhydraulic
computer code and in particular its one-dimensional module which uses the classical two-
phase six equations model. Previous works have demonstrated the feasibility of a steam
injector computational model based on this CATHARE module. Furthermore, with
CATHARE, it will be possible to model complex systems including tanks, valves, heat
exchangers … The code adaptations still necessary (numerics, geometrical modelling
capabilities) will be carried out. Correlations dedicated to the steam injector nominal
functioning will be derived including a pertinent physical description of the domain where
this nominal functioning can be obtained. In addition to the present project experimental
data, one will benefit of two significant sources of information: the 4th European framework
programme SYNTHESIS and the CEA steam injector programme DIVA.
These latter theoretical studies will lead to the new CATHARE module describing steam
injectors. This module will be qualified with the DEEPSSI database (CLAUDIA and IETI
tests). A limited use of other computational tools is foreseen on purpose of better
understanding the SI basic thermalhydraulic phenomena (CFD codes) and benchmarking the
CATHARE plant calculations (RELAP).

Plant models of the reference PWR reactors will be used to calculate some accidental
transients and to assess the plant responses with steam injector based EFWS. The evaluation
of this new safety system will be made by making comparisons to existing solutions. A
limited economic evaluation will also be made to assess the potential cost reduction.



                                             133
Nuclear Energy Programme                                     Evolutionary concepts
Operational safety of existing installations                 Evolutionary safety concepts

Title:
FAST-ACTING BORON INJECTION SYSTEM

Acronym                 FABIS

Proposal number FIS5-2001-00116                        Contract number FIKS-CT2001-00195

Type of action          Shared cost                    Duration           24 months

Starting date           1 September 2001               EC project officer S. Casalta

Total budget            602.891 €                      EC contribution    348.962 €


Co-ordinator
         Organisation           Technical Research Center of Finland (VTT)
         Address                Tekniikantie 4C
                                FIN-02044 VTT Espoo
         Contact person         Dr. Jari Tuunanen
                                Tel:       (358-9)4565081
                                Fax:       (358-9)4565000
                                Email      jari.tuunanen@vtt.fi

Partnership

          Country                     Organisations

            FIN                  Lappeenranta University of Technology
            D                    Framatome ANP GmbH

Project Summary
In the existing Boiling Water Reactors (BWRs), a common cause failure may lead to a
situation where the nuclear fission process can’t be stopped by the active fine motion control
system or by the passive scram system. This is possible because both of them use the same
control rods as neutron absorbing elements. Hence, malfunction of the control rod system
due to a common cause failure has a significant effect on the risk of a core melt accident. To
avoid this, a diverse fast-acting boron injection system is proposed, which injects solution of
sodium pentaborate into the reactor pressure vessel (RPV) or directly to the core. This new
system is passive like the scram system and uses similar working principles. With
comparably short shutdown duration, less than 30 seconds from activation, this system reacts
faster than the existing fine motion control system, where the shutdown duration may be as
long as 90 seconds. The purpose of this project is to confirm that such a fast shutdown can be
realised both in the existing and future BWRs. In this project, SWR 1000 concept from
Framatome ANP has been selected as a reference reactor.



                                                 134
The aim of the project is to find answers to the following questions:
(1) In which time the injected boron spreads over the core to such concentrations that the
nuclear fission reaction stops?
(2) Is the boron mixed so well with the RPV water after the end of injection that the core
entrance flow is borated?
(3) What are the forces acting on the lines between the boron tank and RPV due to thermal
shocks and pressure wave propagations?
(4) How can the results be transformed from the laboratory to the full scale?

Answer to the first question will be found in two steps. First, mixing of injected boron with
core bypass flow will be calculated for a section of the core using a computational fluid
dynamics (CFD) code. The main parameters in the CFD calculations are the mass flow rate
in the core bypass, the diameter, flow velocity and direction of the local ejection nozzles and
the pulse of the entrance flow to the core bypass. The optimisation calculations will lead to a
proper combination of those parameters. Second step is to test the optimum solution in a 1:1
scale perspex test rig, where coloured water simulates boron. This gives the possibility to
compare the optical view of the test with the calculated boron distribution. The test rig will
be built in Erlangen and Framatome ANP will perform the tests.

Answer to the second question will be found using a lumped parameter code. The code is
used to simulate the flows in RPV with neutron coupling in the core. The heat (or steam)
production and in consequence the core mass flow will be calculated over the time. So, the
time period will be calculated until boron solution washed out at the upper end of the core
will again enter the core from below after one period of internal recirculation.

Answer to the third question will be found experimentally. The experiments will be
performed in a test rig, which was earlier used to simulate the flow from a hydraulic scram
system. The experiments will be performed in a test rig at the Lappeenranta University of
Technology. The rig was earlier used to simulate the flow from a hydraulic scram system.
The reaction of the connecting pipe to the pressure and temperature transients will be
measured by wire strain gages. During these tests, the start-up procedure and the stand-by
modus of the passive system will also be simulated because they do not differ principally
from those in the original boron injecting system.

Answer to the fourth question will be found using dimensional analyses and engineering
judgement and performing thermal-mechanical calculations with the code KWUROHR.




                                             135
Nuclear Energy Programme                                  Evolutionary concepts
Operational safety of existing installations - RI         Evolutionary safety concepts

Title:
EUROPEAN NETWORK FOR THE CONSOLIDATION OF THE INTEGRAL SYSTEM
EXPERIMENTAL DATA BASES FOR REACTOR THERMAL-HYDRAULIC SAFETY
ANALYSIS

Acronym                 CERTA

Proposal number FIS5-1999-00213                     Contract number FIR1-CT2000-20052

Type of action          Thematic network            Duration          36 months

Starting date           1 October 2000              EC project officer G. Van Goethem

Total budget            297.800 €                   EC contribution   250.000 €


Co-ordinator
         Organisation         European Commission (EC)
                              JRC/IE
         Address              Via Enrico Fermi 1
                              I-21020 Ispra (VA)
         Contact person       Mr. Carmelo Addabbo
                              Tel:    (39-0332) 789812
                              Fax:    (39-0332) 785584
                              Email   carmelo.addabbo@jrc.it

Partnership
          Country                   Organisations

            I                   Universita' degli Studi di Pisa
            D                   Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH
            D                   Framatome ANP GmbH
            HU                  KFKI Atomic Energy Research Institute (AEKI)
            FIN                 Technical Research Centre of Finland (VTT)
            I                   Società Italiana Esperienze Termoidrauliche (SIET S.p.A.)
            F                   Commissariat à l'Energie Atomique (CEA)
            S                   Studsvik Eco & Safety AB
            CH                  Paul Scherrer Institute (PSI)

Project Summary
The safety evaluation of existing reactors and, in perspective, of evolutionary or innovative
reactor concepts, is generally supported by a wide spectrum of experimental and analytical
efforts aimed at 1) the acquisition of representative experimental data bases in integral
system effect and/or separate effect test facilities and 2) the development of computer codes


                                              136
in order to provide realistic predictions of system and/or component behaviour under
accident conditions.

The extent to which the existing reactor safety experimental data bases are preserved and can
be eventually accessed and/or recovered is an issue often debated in the nuclear community.
In addition to the loss of skilled human resources, a compounding problem is the rapid
advancement of computer hardware and software technology which is making several of the
storage methods obsolete and as such access to the data practically impaired.

The programmatic objective of the proposed network is thus aimed at providing a
consolidated framework for the preservation of the integral system experimental data bases
for reactor thermal-hydraulic safety analysis acquired in the context of the research programs
carried out by European institutional and industrial research organisations. The specific
objectives include:

   assessment of current practices adopted within the participating organisations in the
    storage of the reactor safety experimental thermal-hydraulic data bases and in the
    maintenance of the related documentation,
   definition of optimised data storage and access requirements for the verification and
    validation of system codes used in reactor thermal-hydraulic safety analysis,
   establishment of a user-friendly, web-based distributed informatic platform based on
    modern informatic technologies and provision of a demonstration package for remote
    data access and retrieval.

As structured, the network includes experimental programs and data bases relevant to
reactors in operation within the EU member countries (i.e. PWRs and BWRs) as well as to
reactors in operation within the Central and Eastern-European Countries and the New
Independent States (i.e. VVERs).




                                             137
Nuclear Energy Programme                                  Evolutionary concepts
Operational safety of existing installations - RI         Evolutionary safety concepts

Title:
IMPROVEMENT OF TECHNIQUES FOR MULTISCALE MODELLING OF IRRADIATED
MATERIALS

Acronym                 ITEM

Proposal number FIS5-2001-00031                     Contract number FIR1-CT2001-20136

Type of action          Thematic network            Duration             48 months

Starting date           1 November 2001             EC project officer G. Van Goethem

Total budget            442.933 €                   EC contribution      397.733 €


Co-ordinator
         Organisation          Electricité de France (EDF)
                               Departement Etude des Matériaux
         Address               EDF Site des Renardières
                               F-77818 Moret sur Loing
         Contact person        Prof. Jean-Claude Van Duysen
                               Tel:      (33-1)60736813
                               Fax:      (33-1)60737369
                               Email     jean-claude.van-duysen@edf.fr

Partnership

          Country                   Organisations

            F                   Commissariat à l'Energie Atomique (CEA)
            UK                  University of Liverpool
            E                   Universidad Complutense de Madrid
            F                   Centre National de la Recherche Scientifique/ONERA
            I                   Ente per le Nuove Tecnologie, l'Energia e l'Ambiente (ENEA)
            B                   Belgian Nuclear Research Centre (SCK-CEN)
            F                   Université de technologie de Compiègne
            E                   Universidad Politecnica de Catalunya
            UK                  University of Edinburgh
            FIN                 University of Helsinki
            D                   Max-Planck-Institut für Metallforschung
            RU                  Russian Research Centre "Kurchatov Institute"
            RU                  Ioffe Institut
            S                   Kungliga tekniska högskolan
            F                   Laboratoire de Recherches sur la Réactivité des Solides
            FIN                 GPM2 Unité Mixte de Recherche



                                              138
          F                   Ecole Nationale Supérieure des Mines de St-Etienne
          F                   LTPCM Unité Mixte de Recherche
          CZ                  Charles University in Prague
          RU                  St-Petersburg State Technical University
          I                   Instituto Nazionale per la Fisica della materia
          NL                  University of Groningen
          F                   Ecole Centrale de Paris
          F                   Centre National de la Recherche Scientifique
          UK                  Kings College London
          E                   Universidad Politecnica de Madrid (UPM)
          B                   Université Libre de Bruxelles (ULB)
          E                   Centro de Investigaciones Energéticas, Medioambientales y
                              Tecnológicas (CIEMAT)
          CH                  Paul Scherrer Institut (PSI)
          CH                  CRPP Fusion Technology - Materials
          DK                  Risoe National Laboratory
          E                   University of Seville
          RU                  State Research Center of Russian Federation - Troitsk Institute
          D                   Hahn-Meitner-Institut Berlin GmbH
          D                   Universität Augsburg

Project Summary
The use of Test reactors and hot cell facilities for studying irradiated materials becomes more
and more problematic. One way of partially solving this problem is to develop tools for
computer simulation of radiation effects in materials, applying multiscale modelling
techniques. The development of tools of this type specifically for the nuclear domain (called
Virtual Test reactors : VTRs) began recently (1998) on specific issues with current
simulation techniques. However, it is paramount to prepare a new generation of techniques,
which will allow to enlarge the application field of VTRs within few years. The objective of
the proposed Thematic network is to ensure that these developments are performed rapidly
and in a co-ordinated way in Europe. It is also a great opportunity for the European Nuclear
Industry to lead such an international effort in the direction of its interest. Multiscale
modelling approach is also being aggressively pursued in the US and in Japan. The Network
will allow Europe to exchange information at an equal level with these countries.

The activity of the Network will be carried out within seven Technical Areas which are listed
below with their main objectives :
1) Website and Database construction and Maintenance : the Database will contain codes,
   values of parameters and any other information relevant to enable interested research
   groups to perform simulations, without having to retrieve these data from the literature.
2) Radiation damage Modelling : the objective is to study the formation and evolution of
   radiation induced defects in simple systems (clustering of point defects...) at different time
   and space scales. The interaction between those defects and dislocations will also be
   studied.
3) Simulation of Mechanical Properties of Single Crystals : the objective consists in solving
   issues concerning the treatment of Dislocation Dynamics problems in relation to single-
   crystal plasticity.
4) Polycrystal Simulation Methods : the objective is to try to link atomic-level information
   about extended defects and microstructure evolution into new and/or improved


                                              139
   mesoscopic and macroscopic multiscale models of the thermal/mechanical/radiation
   response of polycrystalline materials
5) Multiscale Modelling Applications : the objective is to improve the VTRs from the point
   of view of code-coupling efficiency and computing speed.
6) Experimental Validation : this will give to the participants of the Network the possibility
   of validating step by step key points of the developments of the computational tools, by
   carrying out mechanical tests and/or microstructure characterisations.
7) Phase Stability and Kinetics of Phase Transformations in Alloys under Irradiation : the
   objective is to assess new methods to study the stability of phases of alloys under
   irradiation.

For each Technical Area, the Network will produce one or more deliverables (models, code,
consensual opinion, ....) for solving a key issue or defining the direction of further work.
These tools will be used to improve one or more of the VTRs currently under development
within the European Union. All these results will be made available through the Website and
European Database.




                                             140
Nuclear Energy Programme                                    Evolutionary concepts
Operational safety of existing installations                High burn-up and MOX fuel

Title:
THE INFLUENCE OF MICROSTRUCTURE OF MOX FUEL ON ITS IRRADIATION
BEHAVIOUR UNDER TRANSIENT CONDITIONS

Acronym                 MICROMOX

Proposal number FIS5-1999-00149                       Contract number FIKS-CT2000-00030

Type of action          Shared cost                   Duration           48 months

Starting date           1 October 2000                EC project officer A. Zurita

Total budget            2.346.011 €                   EC contribution    800.027 €


Co-ordinator
         Organisation          Belgonucléaire S.A.
                               Engineering
         Address               Avenue Ariane, 4
                               B-1200 Brussels
         Contact person        Mr. Marc Lippens
                               Tel:     (32-2) 7740625
                               Fax:     (32-2) 7740547
                               Email    m.lippens@belgonucleaire.be

Partnership

          Country                     Organisations

            UK                  British Nuclear Fuels plc (BNFL)
            INT                 European Commission - JRC/IE
            INT                 European Commission - JRC/ITU
            NL                  Nuclear Research and Consultancy Group (NRG)
            CH                  Paul Scherrer Institute (PSI)

Project Summary
Achievement of high burnup with MOX fuel-economically recommended - is presently
limited due to excessive fission gas release, leading to large consumption of margins
relatively to design criteria. Reduction of gas release and rod inner pressure are possible by
combining or using one of the following methods: increase of rod inner free volume,
reduction of fuel central temperature, use of fuel with increased capability of gas retention.

The objectives of the MICROMOX project are to fabricate, irradiate and test in transient
conditions MOX fuels potentially presenting different degrees of gas retention.



                                                141
Today, MOX fuels are prepared by mechanical blending of UO2 and PuO2 powders giving a
Pu distributed almost everywhere in the UO2 matrix and Pu-rich zones of a few tens micron
size finely dispersed in the UO2 matrix. Alternative fuels showing an enhanced Pu
homogeneity over the previous ones are contemplated as potentially having a better
capability for gas retention.

MOX fuel having larger grain size than usual is also considered as having better retention
capability. The impact of Pu homogeneity in the fuel on the fission gas release will be
studied in the project. For this, ITU will fabricate MOX fuels with homogeneous and
heterogeneous Pu distributions. MOX with a large grain size will also be fabricated, together
with UO2 as reference material.

These fuels will be loaded in rodlets instrumented for temperature and- pressure
measurements, and irradiated at moderate rating in the High Flux Reactor (HFR) to achieve a
burnup of 60 GWd/tM. The end of the irradiation will consist in a temperature transient
allowing following fission gas release as a function of fuel temperature. Post-irradiation
examinations of fuel will be made at NRG and PSI, focusing on fission gas release and fuel
microstructural investigations.

Fuel modellers (BN, BNFL, ITU, PSI) having developed codes to simulate the in-reactor
behaviour of MOX fuel will contribute in project by performing specific calculations.




                                            142
Nuclear Energy Programme                                    Evolutionary concepts
Operational safety of existing installations                High burn-up and MOX fuel

Title:
OXIDE FUELS: MICROSTRUCTURE AND COMPOSITION VARIATIONS

Acronym                 OMICO

Proposal number FIS5-2001-00037                       Contract number FIKS-CT2001-00141

Type of action          Shared cost                   Duration           36 months

Starting date           1 October 2001                EC project officer A. Zurita

Total budget            2.047.945 €                   EC contribution    1.023.972 €


Co-ordinator
         Organisation          Belgian Nuclear Research Centre (SCK-CEN)
                               Reactor Materials Research
         Address               Boeretang 200
                               B-2400 Mol
         Contact person        Dr. Marc Verwerft
                               Tel:     (32-14)333048
                               Fax:     (32-14)321216
                               Email    mverwerf@sckcen.be

Partnership

          Country                     Organisations

            INT                 European Commission - JRC/ITU
            F                   Framatome ANP

Project Summary
The project OMICO compares the behaviour of oxide fuels with homogeneous and
heterogeneous microstructure and with three different chemical compositions. It addresses
fundamental questions on the mechanisms that govern the release of fission gas. This is to be
achieved through the irradiation and in-pile measurement of centreline temperature and
internal pressure of a small bundle of experimental fuel pins. At regular intervals, a sub-
assembly will be unloaded and measured non-destructively. The irradiation conditions will
be fine-tuned on the basis of concurrent modelling of the fuel behaviour and the comparison
of calculated predictions with experimental results of both the in-pile and out-of-pile
measurements.

The proposed test matrix compares in a systematic way the behaviour of three different fuel
compositions (UO2, (U,Pu)O2 and (Th,Pu)O2), and for each composition, two different



                                                143
microstructures are inter-compared (homogeneous and fine dispersed ceramic-in-ceramic).
This results in six different fuel types, which will be assembled in a small experimental
assembly and irradiated in a pressurised water loop of the BR-2 materials testing reactor
(SCK•CEN) that simulates the thermo-hydraulic conditions of a typical Pressurised Water
Reactor. During irradiation, the fuel temperature and gas pressure will be monitored. The
primary objective is to provide insight in the separate effects of fuel chemistry (matrix
composition) on the one hand and the degree of dispersion of the fissile material
(microstructure) on the other hand.

The design of the fuel rods and irradiation conditions will be performed using currently
available models for LWR fuel. Especially the (Th, Pu)O2 type of fuel will require efforts to
incorporate it in the fuel performance codes. The fuels will be fabricated by ITU, loaded in
two sets of rodlets (giving thus a total of twelve rodlets), one set of which is instrumented
with a central thermocouple and a pressure transducer. A detailed characterisation of the
fuel, with emphasis on its microstructure and thermal properties will complete the fuel
fabrication. It is foreseen to irradiate the samples during ten cycles of 21 days, and to achieve
a burnup of about 25GWd/tM and 2-5% fission gas release. The set of non-instrumented rods
will be unloaded intermittently to study critical fuel performance indicators (cladding
corrosion, creep, fuel swelling) as well as to perform an independent experimental control of
the power of the fuel rods.

Using the experimental results (detailed characterisation, in-pile data and out-of-pile non
destructive analyses), a benchmarking of the codes for the different fuels will be performed
(Framatome, SCK•CEN and ITU). The project will also focus on the development of a
model for fission gas release that takes better account for the microstructure of the fuel.




                                              144
Nuclear Energy Programme                                    Evolutionary concepts
Operational safety of existing installations                High burn-up and MOX fuel

Title:
VALIDATION OF HIGH BURNUP MOX FUELS CALCULATIONS

Acronym                 VALMOX

Proposal number FIS5-2001-00107                       Contract number FIKS-CT2001-00191

Type of action          Shared cost                   Duration           30 months

Starting date           1 October 2001                EC project officer A. Zurita

Total budget            1.000.000 €                   EC contribution    500.000 €


Co-ordinator
         Organisation          Belgonucléaire S.A.
         Address               Avenue Ariane, 4
                               B-1200 Brussels
         Contact person        Mr. Servais Pilate
                               Tel:      (32-2)7740569
                               Fax:      (32-2)7740547
                               Email     s.pilate@belgonucleaire.be

Partnership

          Country                     Organisations

            B                   Belgian Nuclear Research Centre (SCK-CEN)
            F                   Commissariat à l'Energie Atomique (CEA)
            NL                  Nuclear Research and Consultancy Group (NRG)

Project Summary
Achievement of high-burnup with MOX fuel is desirable for economic reasons: the fuel
cycle cost contributes for about 25 % to the electricity generation cost in present-day LWRs,
and an increase in fuel burnup reduces the fuel cycle cost.

A gradual raise in burnup is authorised by the Safety Authorities on the basis of experimental
followed by post-irradiation examinations, which give as major results, the fuel isotopic
mass-balances following irradiation, and also the amount of helium gas and fission product
gases produced in the fuel, which determine the internal pin pressure at the end of life.

The partners in this project will evaluate high-burnup MOX fuel irradiations recently
performed in large LWRs, using JEF nuclear data files together with state-of-the-art
neutronics codes. Important experimental data banks are available in Belgium and in France.



                                                145
BN and SCK-CEN have available results of irradiations measured for PWRs (ARIANE
project) and for BWRs. In France, EdF and CEA have obtained irradiation results for two
modern PWR plants (Saint-Laurent-des-Eaux B1 and Dampierre). In both countries, the fuel
burnup reached 45 to 60 GWd/t (average at discharge), well over the values presently
licensed.

The partners will first calculate the mass balances in their own irradiation experiments (WP1
and WP2 in parallel) using well-validated computing procedures. They will later
intercompare the trends observed in the calculated-over-experimental (C/E) ratios for Pu and
minor actinide isotopes, and for fission products.

NRG will contribute to these evaluations by performing sensitivity and uncertainty
calculations, so as to be able to relate the C/E discrepancies to possible deficiencies in the
JEF nuclear data.

Particular attention will be given to the formation of helium gas, as its build-up gives rise to
part of the pin internal pressure, which is often the major parameter limiting the fuel
irradiation.

Ultimately, revisions of cross-sections will be proposed to the OECD/NEA group in charge
of the new JEFF-3 nuclear data file.




                                              146
Nuclear Energy Programme                                     Evolutionary concepts
Operational safety of existing installations                 High burn-up and MOX fuel

Title:
SIMULATION OF RADIATION EFFECTS IN ZR-NB ALLOYS: APPLICATION TO
STRESS CORROSION CRACKING BEHAVIOUR IN IODINE-RICH ENVIRONMENT

Acronym                 SIRENA

Proposal number FIS5-2001-00032                       Contract number FIKS-CT2001-00137

Type of action          Shared cost                   Duration           36 months

Starting date           1 January 2002                EC project officer G. Van Goethem

Total budget            641.787 €                     EC contribution    398.617 €


Co-ordinator
         Organisation          Electricité de France (EDF)
                               Dept. Etude des Matériaux
         Address               Site des Renardières
                               F-77818 Moret sur Loing
         Contact person        Ms. Stéphanie Jumel
                               Tel:     (33-1)60736174
                               Fax:     (33-1)60737369
                               Email    stephanie.jumel@edf.fr

Partnership

          Country                     Organisations

            E                   Universidad Politecnica de Madrid (UPM)
            UK                  The University of Liverpool
            B                   Université Libre de Bruxelles (ULB)
            F                   Commissariat à l'Energie Atomique (CEA)
            S                   Westinghouse Atom AB
            E                   Universidad Politecnica de Madrid (UPM)
            F                   Centre National de la Recherche Scientifique/ONERA

Project Summary
In order to optimise cladding materials for safe operating conditions of nuclear reactors and
to predict long-term performance for fuel assembly storage, it is important to quantify the
influence of the parameters controlling this type of cracking. The experimental work
necessary to reach this objective is being carried out by nuclear plant operators, fuel
assembly manufacturers, and regulatory safety authorities. Such work is extremely long and
costly (i.e. the cost of a power ramp to assess the resistance of cladding during operation is in



                                                147
the range of 1 million euros). It is therefore interesting to complement it by using state-of-
the-art computer simulation. For this purpose the development of a powerful suite of
simulation tools is necessary.

The objective of the project is to build a simulation suite allowing the modelling of :
- firstly, neutron irradiation effects in the Zr-Nb alloys used to manufacture LWR fuel
   assembly claddings; and
- subsequently, the stress-corrosion cracking behaviour of these irradiated alloys (annealed
   or not) in an iodine-rich environment.

In the scope of the project, this suite will be used to solve issues proposed by the industrial
partners on Zr-Nb alloys. Later, it will be possible to extend its use to other alloys (e. g. Zr-
Nb-Sn-Fe), also used to make fuel cladding.

The planned simulation suite will consist of two modules: a "neutron irradiation" module and
an "iodine-assisted stress-corrosion" module. Both will be built by assembling state-of-the-art
codes and models and will be able to provide fundamental physical insight into the effects of
radiation on these materials, as well as quantitative data for this specific application field.
The production of quantitative results for broader applications (different materials,
conditions,...) will require the development of more advanced computer codes and models. A
European Thematic network proposal devoted to this development (ITEM/contract n° FIR1-
CT2001-20136) is also funded by the European Commission.

Each organisation will bring specific and complementary competencies and simulation tools
in the project ; thus, files of simulation data will have to be exchanged between them. This
will require to use the pan-European Gigabit Research Network GEANT, funded by the
European Commission as part of the IST Programme.

While being developed, the suite of simulation tools will be validated by continuous
comparison with available experimental results, provided by the partners of the project. This
experimental database will have to cover an ‘as-large-as-possible’ range of (1) materials
features (texture, chemical composition,...) and (2) irradiation and corrosion conditions.

Once developed, in the framework of the project the above-described tool will be used in
close collaboration with the partners to: (1) interpret previous test results not yet fully
understood and (2) explore storage conditions for which the consequences on fuel assemblies
are not well known.




                                              148
Nuclear Energy Programme                                    Evolutionary concepts
Operational safety of existing installations                High burn-up and MOX fuel

Title:
EXTENSION OF TRANSURANUS CODE APPLICABILITY WITH NB CONTAINING
CLADDING MODELS

Acronym                 EXTRA

Proposal number FIS5-2001-00083                       Contract number FIKS-CT2001-00173

Type of action          Shared cost                   Duration           24 months

Starting date           1 December 2001               EC project officer A. Zurita

Total budget            438.999 €                     EC contribution    219.499 €


Co-ordinator
         Organisation          Atomic Energy Research Institute (KFKI)
                               Fuel & Reactor Materials Department
         Address               Konkoly Thege str. 29-33
                               HU-1121 Budapest
         Contact person        Dr. Csaba Gyori
                               Tel:      (36-1)3922294
                               Fax:      (36-1)3959293
                               Email     gyori@sunserv.kfki.hu

Partnership

          Country                     Organisations

            SK                  Nuclear Power Plant Research Institute (VUJE) Trnava Inc
            INT                 European Commission - JRC/ITU

Project Summary
The main objective of the project is to provide a widely validated computer code for the
accident assessment of nuclear reactors, especially VVER type reactors, and to improve the
safety culture this way.

Due to the comprehensive materials data bank for different fuels, claddings and coolant, the
TRANSURANUS fuel code (developed by the Institute for Transuranium Elements) is
widely used in the safety evaluation of different types of nuclear reactors (PWR, VVER,
BWR, FBR, HWR, GCR) in East- and West-European countries. The scope of the covered
phenomena and the numerical solution methods of the equation systems make the
TRANSURANUS code capable to simulate both long fuel cycles under normal operating
conditions and hypothetical accidents even in the time scale of milliseconds.



                                                149
However, the application of the TRANSURANUS code to simulate hypothetical accidents
requires the extension of the materials functions up to the failure limits. From the point of
view of fission product release, the simulation of the different cladding failure mechanisms
(overstress, ballooning, oxide layer wall thinning, etc.) is also necessary.

Due to the lack of appropriate tools, the project aims at the extension of the
TRANSURANUS code applicability for accident analyses. The conception is to implement
newly developed oxidation and mechanical models to simulate the high temperature
behaviour of Nb containing cladding materials, applied in VVERs and lately in PWRs as
well. The goal is to be achieved through the following measurable objectives:
1. Database development
In order to provide appropriate background for model development and code validation, an
electronic database will be compiled from the available data of numerous separate effect
tests accomplished at the AEKI with Zr1%Nb cladding samples.
2. Model development
To simulate the high temperature oxidation, the plastic deformation and the burst of Zr1%Nb
claddings, new correlations will be developed and integrated into the TRANSURANUS code
structure.
3. Code validation
The extended TRANSURANUS code will be validated through the comparison of code
results of post-test computations with experimental data. The code validation will cover all
the physical phenomena modelled in the project.
4. Plant application




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