Nuclear Energy Programme

NUCLEAR FISSION AND RADIATION PROTECTION PROJECTS SELECTED FOR FUNDING 1999-2002 ANNEX I OPERATIONAL SAFETY OF EXISTING INSTALLATIONS 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 Integrity of equipment and structures FRAME - Fracture mechanics based embrittlement trend curves for the characterisation of nuclear pressure vessel materials RETROSPEC - Retrospective dosimetry focussed on the reaction 93nb(n,n') PISA - Phosphorus influence on steel ageing RENION - Reactor neutronic investigations on LR-0 reactor INTERWELD - Irradiation effects on the evolution of the microstructure, properties and residual stresses in the heat affected zone of stainless steel welds PRIS - Properties of irradiated stainless steels for predicting lifetime of nuclear power plant components CASTOC - Crack growth behaviour of low alloy steel for pressure boundary components under transient light water reactor (LWR) operating conditions ADIMEW - Assessment of aged piping dissimilar metal weld integrity VOCALIST - Validation of constraint-based assessment methodology in structural integrity SMILE - Structural margin improvements in aged-embrittled RPV with load history effects THERFAT - Thermal fatigue evaluation of piping system "Tee"- connections WAHALOADS - Two-phase flow water hammer transients and induced loads on materials and structures of nuclear power plants FLOMIX-R - Fluid mixing and flow distribution in the reactor circuit FEUNMARR - Future EU needs in materials research reactors MAECENAS - Modelling of ageing in concrete nuclear power plant structures 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 On-line monitoring and maintenance GRETE - Evaluation of non destructive testing techniques for monitoring of material degradation LIRES - Development of Light Water Reactor (LWR) reference electrodes SPIQNAR - Signal processing and improved qualification for non-destructive testing of ageing reactors 41 43 45 6 8 10 12 14 16 18 20 22 25 28 30 33 35 37 4 39 1 REDOS - Reactor Dosimetry: accurate determination and benchmarking of radiation field parameters, relevant for reactor pressure vessel monitoring VRIMOR - Virtual reality for inspection, maintenance, operation, and repair of nuclear power plant NURBIM - Nuclear risk-based inspection methodology ENPOWER - Management of nuclear plant operation by optimising weld repairs Organisation and management of safety BE-SECBS - Benchmark exercise on safety evaluation of computer-based systems CEMSIS - Cost effective modernisation of systems important to safety LearnSafe - Learning organisations for nuclear safety SPI - Evaluation of alternative approaches for assessment of safety performance indicators for nuclear power plants Safety of VVER reactors IMPAM-VVER - Improved accident management of VVER nuclear power plants VERLIFE - Unified procedure for lifetime assessment of components and piping in VVER NPPS ATHENA - AMES thematic network on ageing SEVERE ACCIDENT MANAGEMENT Assessment of severe accident risks COLOSS - Core loss during a severe accident LISSAC - Limit strains for severe accident conditions ARVI - Assessment of reactor vessel integrity ENTHALPY - European nuclear thermodynamic database (for in- and ex-vessel applications) ECOSTAR - Ex-vessel core melt stabilisation research HYCOM - Integral large scale experiments on hydrogen combustion for severe accident code validation EVITA - European validation of the integral code ASTEC LPP - Late phase source term phenomena PHEBEN 2 - Validation of severe accident codes against Phebus FP for plant applications ASTERISM II - Archive models for source term information and system models EURSAFE - European expert network for the reduction of uncertainties in severe accident safety issues THENPHEBISP - Thematic network for a Phebus FPT-1 international standard problem SCACEX - Scaling of containment experiments PLINIUS - Platform for improvements in nuclear industry and utility safety LACOMERA - Large scale experiments on core degradation, melt retention and coolability 47 49 51 53 55 57 59 61 63 65 67 69 71 73 75 77 79 81 83 85 87 89 91 93 95 97 2 Severe accident management measures EUROCORE - European group for analysis of corium recovery concepts SGTR - Steam generator tube rupture scenarios ICHEMM - Iodine chemistry and mitigation methods THINCAT - Hydrogen removal from LWR containments by catalytic coated thermal insulation elements PARSOAR - Hydrogen hazard - passive autocatalytic recombiners state-of-the-art OPTSAM - Optimisation of severe accident management strategies for the control of radiological releases SAMOS - A perspective on computerized severe accident management operator support VERSAFE - Concerted utility review of VVER-440 safety research needs EVOLUTIONARY CONCEPTS Evolutionary safety concepts ASTAR - Advanced three-dimensional two-phase flow simulation tool for application to reactor safety TEMPEST - Testing and enhanced modelling of passive evolutionary systems technology (for containment cooling) ECORA - Evaluation of computational fluid dynamic methods for reactor safety analyses EUROFASTNET - European group for future advances in sciences and technology for nuclear engineering thermalhydraulics CRISSUE-S - Revisiting critical issues in nuclear reactor design / safety by using 3-D neutronics / thermalhydraulics models: state-of-the-art VALCO - Validation of coupled neutronics/thermal hydraulics codes for VVER reactors RMPS - Reliability methods for passive safety functions NACUSP - Natural circulation and stability performance of BWRs DEEPSSI - Design and development of a steam generator emergency feedwater passive system for existing and future PWR's using advanced steam injectors FABIS - Fast-acting boron injection system CERTA - European network for the consolidation of the integral system experimental data bases for reactor thermal-hydraulic safety analysis ITEM - Improvement of techniques for multiscale modelling of irradiated materials 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 VALMOX - Validation of high burnup mox fuels calculations SIRENA - Simulation of radiation effects in Zr-Nb alloys: application to stress corrosion cracking behaviour in iodine-rich environment 143 145 147 115 117 119 121 123 126 128 130 132 134 136 138 99 101 103 105 107 109 111 113 EXTRA - Extension of transuranus code applicability with Nb containing cladding models 149 3 Contents Nuclear Energy Programme Operational safety of existing installations - RI Title: JOINT SAFETY RESEARCH INDEX Plant life extension and management General Acronym JSRI Contract number FIR1-CT2000-20089 Duration 30 months Proposal number FIS5-1999-00302 Type of action Starting date Total budget* Concerted action 1 January 2001 307.784 € EC project officer G. Van Goethem EC contribution* 299.909 € Co-ordinator Organisation Address Contact person Gesellschaft für Anlagen - und Reaktorsicherheit (GRS) GmbH Research Management Division Schwertnergasse 1 D-50667 Köln Dr. Axel Breest Tel: (49-221) 2068667 Fax: (49-221) 2068629 Email bre@grs.de Partnership Country INT E RO NL I D UK F CZ HU CH * Organisations European Commission - JRC/ISIS Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT) Center of Technology and Engineering for Nuclear Projects (CITON) Nuclear Research and Consultancy Group (NRG) Ente per le Nuove Tecnologie, l'Energia e l'Ambiente (ENEA) Forschungszentrum Karlsruhe GmbH (FZK) Health and Safety Executive (HSE) Institut de Radioprotection et de Sureté Nucléaire (IRSN) Nuclear Research Institute Rež plc (NRI) Nuclear Services Ltd 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 Contents DK B S FIN RISOE National Laboratory Belgian Nuclear Research Centre (SCK-CEN) Swedish Nuclear Power Inspectorate (SKI) 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 cooperation 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 Contents Nuclear Energy Programme Operational safety of existing installations Plant life extension and management Integrity of equipment and structures Title: FRACTURE MECHANICS BASED EMBRITTLEMENT TREND CURVES FOR THE CHARACTERISATION OF NUCLEAR PRESSURE VESSEL MATERIALS Acronym FRAME Contract number FIKS-CT2000-00101 Duration 36 months Proposal number FIS5-1999-00325 Type of action Starting date Total budget Shared cost 1 September 2000 858.905 € * EC project officer P. Manolatos EC contribution 429.453 € * Co-ordinator Organisation Address Contact person Technical Research Centre of Finland (VTT) Kemistintie 3 FIN-02044 Espoo 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 Contents 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 nonconservatism 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 Contents Nuclear Energy Programme Operational safety of existing installations Plant life extension and management Integrity of equipment and structures Title: RETROSPECTIVE DOSIMETRY FOCUSSED ON THE REACTION 93NB(N,N') Acronym RETROSPEC Contract number FIKS-CT2000-00091 Duration 25 months Proposal number FIS5-1999-00305 Type of action Starting date Total budget Shared cost 1 October 2000 318.982 € EC project officer P. Manolatos EC contribution 159.491 € Co-ordinator Organisation Address Contact person Nuclear Research and Consultancy Group (NRG) Materials, Monitoring & Inspection Westerduinweg 3 NL-1755 ZG Petten Dr. Willem Voorbraak Tel: (31-224) 564295 Fax: (31-224) 564457 Email voorbraak@nrg-nl.com Partnership Country B FIN Organisations Belgian Nuclear Research Centre (SCK-CEN) 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 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations Plant life extension and management Integrity of equipment and structures Title: PHOSPHORUS INFLUENCE ON STEEL AGEING Acronym PISA Contract number FIKS-CT2000-00080 Duration 36 months Proposal number FIS5-1999-00269 Type of action Starting date Total budget Shared cost 1 December 2000 1.618.840 € EC project officer P. Manolatos EC contribution 794.912 € Co-ordinator Organisation Address Contact person AEA Technology Plc Nuclear Science 220 Harwell UK-OX11 0RA Didcot, Oxfordshire Dr. Colin English Tel: (44-1235) 434342 Fax: (44-1235) 435941 Email colin.english@aeat.co.uk Partnership Country INT F D UK HU FIN CZ E UK D Organisations European Commission - JRC/IE Electricité de France (EDF) Framatome ANP GmbH British Nuclear Fuel plc (BNFL) KFKI Atomic Energy Research Institute (AEKI) Technical Research Centre of Finland (VTT) Nuclear Research Institute Rež plc (NRI) Tecnatom S.A. The University of Liverpool 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 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations - RI Plant life extension and management Integrity of equipment and structures Title: REACTOR NEUTRONIC INVESTIGATIONS ON LR-0 REACTOR Acronym RENION Contract number FIR1-CT2002-40157 Duration 24 months Proposal number FIS5-2002-00052 Type of action Starting date Total budget Research Infrastructures 1 February 2003 137.444 € EC project officer A. Zurita EC contribution 137.444 € Co-ordinator Organisation 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 Ústav Jaderného Výzkumu Rež A.S. Nuclear Power and Safety Divison 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 cooperation should allow to maintain and to extend the EU competence in the area of reactor experimental research. 12 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations Plant life extension and management 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 Contract number FIKS-CT2000-00103 Duration 42 months Proposal number FIS5-1999-00332 Type of action Starting date Total budget Shared cost 1 September 2000 1.739.960 € EC project officer P. Manolatos EC contribution 660.631 € Co-ordinator Organisation Address Contact person Nuclear Research and Consultancy Group (NRG) Materials, Monitoring & Inspection Westerduinweg 3 NL-1755 ZG Petten Mr. Bob Van der Schaaf Tel: (31-224) 564665 Fax: (31-224) 568490 Email vanderschaaf@nrg-nl.com Partnership Country D INT CH B E Organisations Framatome ANP GmbH European Commission - JRC/IE Paul Scherrer Institute (PSI) Belgian Nuclear Research Centre (SCK-CEN) 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 Contents - 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 Contents Nuclear Energy Programme Operational safety of existing installations Plant life extension and management Integrity of equipment and structures Title: PROPERTIES OF IRRADIATED STAINLESS STEELS FOR PREDICTING LIFETIME OF NUCLEAR POWER PLANT COMPONENTS Acronym PRIS Contract number FIKS-CT2000-00084 Duration 36 months Proposal number FIS5-1999-00277 Type of action Starting date Total budget Shared cost 1 October 2000 1.143.531 € EC project officer P. Manolatos EC contribution 499.966 € Co-ordinator Organisation Address Contact person ABB Atom Ab Nuclear Services Gideonsbergsgatan, 2 S-72163 Västeras Mr. Henrik Westermark Tel: (46-21)347000 Fax: (46-21)348500 Email henrik.westermark@se.westinghouse.com Partnership Country E F B D S FIN Organisations Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT) Framatome ANP Belgian Nuclear Research Centre (SCK-CEN) Framatome ANP GmbH Studsvik Material AB 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 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations Plant life extension and management 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 Contract number FIKS-CT2000-00048 Duration 36 months Proposal number FIS5-1999-00198 Type of action Starting date Total budget Shared cost 1 September 2000 1.431.759 € EC project officer P. Manolatos EC contribution 600.003 € Co-ordinator Organisation Address Contact person Staatliche Materialpruefungsanstalt (MPA Stuttgart) Department for Environmental Effects Pfaffenwaldring 32 D-70569 Stuttgart Dr. Juergen Foehl Tel: (49-711) 6852564 Fax: (49-711) 6852761 Email foehl@mpa.uni-stuttgart.de Partnership Country E CZ CH D FIN Organisations Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT) Nuclear Research Institute Rež plc (NRI) Paul Scherrer Institute (PSI) Framatome ANP GmbH 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 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations Plant life extension and management Integrity of equipment and structures Title: ASSESSMENT OF AGED PIPING DISSIMILAR METAL WELD INTEGRITY Acronym ADIMEW Contract number FIKS-CT2000-00047 Duration 36 months Proposal number FIS5-1999-00187 Type of action Starting date Total budget Shared cost 1 November 2000 1.152.578 € * EC project officer P. Manolatos EC contribution 512.100 € * Co-ordinator Organisation Address Contact person Electricité de France (EDF) Engineering and Service Division - Septen 12 -14 Avenue Dutrievoz F-69628 Villeurbanne Cedex 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 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations Plant life extension and management Integrity of equipment and structures Title: VALIDATION OF CONSTRAINT-BASED STRUCTURAL INTEGRITY ASSESSMENT METHODOLOGY IN Acronym VOCALIST Contract number FIKS-CT2000-00090 Duration 36 months Proposal number FIS5-1999-00303 Type of action Starting date Total budget Shared cost 1 October 2000 1.747.338 € * EC project officer P. Manolatos EC contribution 746.169 € * Co-ordinator Organisation Address Contact person Serco Assurance Engineering Integrity Group Risley UK-WA3 6AT Warrington 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 Contents 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. Cracktip 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 largescale 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 constraintbased 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 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations Plant life extension and management Integrity of equipment and structures Title: STRUCTURAL MARGIN IMPROVEMENTS IN AGED-EMBRITTLED RPV WITH LOAD HISTORY EFFECTS Acronym SMILE Contract number FIKS-CT2001-00131 Duration 36 months Proposal number FIS5-2001-00023 Type of action Starting date Total budget Shared cost 1 January 2002 1.725.053 € EC project officer P. Manolatos EC contribution 787.526 € Co-ordinator Organisation Address Contact person Electricité de France (EDF) Division Production Nucléaire Site Cap Ampère, 1 Place Pleyel F-93282 Saint-Denis Mr. Georges Bezdikian Tel: (33-1)43693848 Fax: (33-1)43693482 Email georges.bezdikian@edf.fr Partnership Country UK F F D UK F D INT D US Organisations Serco Assurance Commissariat à l'Energie Atomique (CEA) Ministere de l'Economie, des Finances et de l'Industrie Staatliche Materialprufungsanstalt (MPA Stuttgart) British Energy Generation Ltd Framatome ANP Framatome ANP GmbH European Commission - JRC/IE E. ON Kernkraft GmbH Oak Ridge National Laboratory 25 Contents 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 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations Plant life extension and management Integrity of equipment and structures Title: THERMAL FATIGUE EVALUATION OF PIPING SYSTEM "TEE"- CONNECTIONS Acronym THERFAT Contract number FIKS-CT2001-00158 Duration 36 months Proposal number FIS5-2001-00043/58 Type of action Starting date Total budget Shared cost 1 December 2001 1.679.925 € * EC project officer P. Manolatos EC contribution 839.963 € * Co-ordinator Organisation Address Contact person E.ON Kernkraft GmbH Dept. of mechanics, materials, non destructive testing (TTF) Tresckowstrasse 5 D-30457 Hannover Mr. Klaus-Jürgen Metzner Tel: (49-511)4394009 Fax: (49-511)4394377 Email klaus-juergen.metzner@eon-energie.com Organisations Partnership Country F F F E D Electricité de France (EDF) Framatome ANP Commissariat à l'Energie Atomique (CEA) Tecnatom S.A. 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 Contents 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, thermohydraulic 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 neurofuzzy 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 Contents Nuclear Energy Programme Operational safety of existing installations Plant life extension and management 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 Contract number FIKS-CT2000-00106 Duration 36 months Proposal number FIS5-1999-00114/341 Type of action Starting date Total budget Shared cost 1 October 2000 2.089.758 € EC project officer G. Van Goethem EC contribution 1.269.869 € Co-ordinator Organisation Address Contact person Université Catholique de Louvain (UCL) Unité Thermodynamique Place du Levant, 2 B-1348 Louvain-la-Neuve Prof. Michel Giot Tel: (32-10) 472210 Fax: (32-10) 452692 Email giot@term.ucl.ac.be Partnership Country F E SI B F E D HU D D Organisations Commissariat à l'Energie Atomique (CEA) Iberdrola S.A. Institute "Josef Stefan" Tractebel S.A. Electricité de France (EDF) Empresarios Agrupados Internacional S.A. Framatome ANP GmbH KFKI Atomic Energy Research Institute (AEKI) Fraunhofer-Gesellschaft zur Förderung der angewandten Forschung e.V./ UMSICHT Forschungszentrum Rossendorf (FZR) 30 Contents 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 nonhomogeneous, 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 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations Plant life extension and management Integrity of equipment and structures Title: FLUID MIXING AND FLOW DISTRIBUTION IN THE REACTOR CIRCUIT Acronym FLOMIX-R Contract number FIKS-CT2001-00197 Duration 36 months Proposal number FIS5-2001-00119 Type of action Starting date Total budget Shared cost 1 October 2001 1.256.915 € * EC project officer G. Van Goethem EC contribution 703.788 € * Co-ordinator Organisation Address Contact person Forschungszentrum Rossendorf E.V. (FZR) Institute of Safety Research PO Box 511019 D-01314 Dresden 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 Contents 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 restart 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 Contents Nuclear Energy Programme Operational safety of existing installations - RI Plant life extension and management Integrity of equipment and structures Title: FUTURE EU NEEDS IN MATERIALS RESEARCH REACTORS Acronym FEUNMARR Contract number FIR1-CT2001-20122 Duration 12 months Proposal number FIS5-2001-00007 Type of action Starting date Total budget Thematic network 1 November 2001 119.950 € EC project officer A. Zurita EC contribution 119.950 € Co-ordinator Organisation Address Contact person Commissariat à l'Energie Atomique (CEA) Nuclear Energy Direction C.N. Cadarache BT.224 F-13108 Saint-Paul-les-Durance Mr. Daniel Parrat Tel: (33-4)42257572 Fax: (33-4)42254777 Email daniel.parrat@cea.fr Partnership Country B INT UK CZ D F UK F F Organisations Belgian Nuclear Research Centre (SCK-CEN) European Commission - JRC/IE NAC International Nuclear Research Institute Rež plc (NRI) Framatome ANP GmbH Technicatom Independent Consultant OECD - Nuclear Energy Agency 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 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations Plant life extension and management Integrity of equipment and structures Title: MODELLING OF AGEING IN CONCRETE NUCLEAR POWER PLANT STRUCTURES Acronym MAECENAS Contract number FIKS-CT2001-00186 Duration 36 months Proposal number FIS5-2001-00100 Type of action Starting date Total budget Shared cost 1 November 2001 1.160.424 € EC project officer G. Van Goethem EC contribution 1.099.257 € Co-ordinator Organisation Address Contact person University of Sheffield Dept. of Civil and Structural Engineering Sir Frederick Mappin Building, Mappin Street UK-S1 3JD Sheffield Dr. Roger Crouch Tel: (44-144)2225716 Fax: (44-144)2225700 Email r.crouch@sheffield.ac.uk Partnership Country CZ UK UK F UK I I Organisations Ceské Vysoké Uceni Techniké v Praze (Technical University) University of Glasgow Health and Safety Executive (HSE) Ecole Centrale de Nantes British Energy Generation Ltd University of Rome "La Sapienza" 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 Contents 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 objectoriented 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 Contents Nuclear Energy Programme Operational safety of existing installations Plant life extension and management 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 Contract number FIKS-CT2001-00204 Duration 36 months Proposal number FIS5-2001-00125 Type of action Starting date Total budget Shared cost 1 January 2002 1.334.760 € EC project officer G. Van Goethem EC contribution 454.000 € Co-ordinator Organisation Address Contact person Force Institute Division for Materials and Chemical Analysis Park Alle 345 DK-2605 Broendby Mr. Oskar Klinghoffer Tel: (45-43)267255 Fax: (45-43)267011 Email osk@force.dk Partnership Country S S F Organisations Scanscot Technology AB Barsebaeck Kraft AB 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 Contents 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-theart 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 Contents Nuclear Energy Programme Operational safety of existing installations Plant life extension and management On-line monitoring and maintenance Title: EVALUATION OF NON DESTRUCTIVE TESTING TECHNIQUES FOR MONITORING OF MATERIAL DEGRADATION Acronym GRETE Contract number FIKS-CT2000-00086 Duration 36 months Proposal number FIS5-1999-00280 Type of action Starting date Total budget Shared cost 1 October 2000 1.527.358 € EC project officer P. Manolatos EC contribution 669.975 € Co-ordinator Organisation Address Contact person Electricité de France (EDF) - R&D Division Materials Studies Branch Route de Sens, Ecuelles B.p. 1 F-77818 Moret-sur-Loing Cedex Mr. Marc Delnondedieu Tel: (33-1) 60736315 Fax: (33-1) 60736889 Email marc.delnondedieu@edf.fr Partnership Country NL FIN E D INT UK E A HU D CH Organisations Nuclear Research and Consultancy Group (NRG) Technical Research Centre of Finland (VTT) Tecnatom S.A. Fraunhofer-Gesellschaft zur Förderung der angewandten Forschung e.V. (FhG-IZFP) European Commission - JRC/IE Serco Assurance Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT) Österreichisches Forschungszentrum Seibersdorf Ges.m.b.H. (ARCS) KFKI Atomic Energy Research Institute (AEKI) University of Hannover Paul Scherrer Institut (PSI) 41 Contents CZ D F RU D Nuclear Research Institute Rež plc (NRI) Siempelkamp Pruf- und Gutachter - Gesellschaft MBH INSAVALOR All-Russian Institute for Nuclear Power Plants Operation (VNIIAES) 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 Contents Nuclear Energy Programme Operational safety of existing installations Plant life extension and management On-line monitoring and maintenance Title: DEVELOPMENT OF LIGHT WATER REACTOR (LWR) REFERENCE ELECTRODES Acronym LIRES Contract number FIKS-CT2000-00012 Duration 48 months Proposal number FIS5-1999-00113 Type of action Starting date Total budget Shared cost 1 October 2000 1.256.416 € EC project officer P. Manolatos EC contribution 649.305 € Co-ordinator Organisation Address Contact person Belgian Nuclear Research Centre (SCK-CEN) Reactor Materials Research Boeretang 200 B-2400 Mol Dr. Rik-Wouter Bosch Tel: (32-14) 333428 Fax: (32-14) 321336 Email rbosch@sckcen.be Partnership Country HU F E B CZ D S FIN Organisations KFKI Atomic Energy Research Institute (AEKI) Commissariat à l'Energie Atomique (CEA) Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT) Katholieke Universiteit Leuven (KUL) Nuclear Research Institute Rež plc (NRI) Framatome ANP GmbH Studsvik Material AB 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 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations Plant life extension and management On-line monitoring and maintenance Title: SIGNAL PROCESSING AND IMPROVED QUALIFICATION FOR NON-DESTRUCTIVE TESTING OF AGEING REACTORS Acronym SPIQNAR Contract number FIKS-CT2000-00065 Duration 36 months Proposal number FIS5-1999-00233/251 Type of action Starting date Total budget Shared cost 1 October 2000 1.746.368 € EC project officer P. Manolatos EC contribution 999.995 € Co-ordinator Organisation Address Contact person Mitsui Babcock Technology Centre High Street UK-PA4 8UW Renfrew Mr. Neil Cameron Tel: +44 141 886 4141 Fax: +44 141 885 3338 Email ncameron@mitsuibabcock.com Partnership Country UK B S D F INT F CZ E D S UK Organisations British Energy Generation Ltd AIB - VINCOTTE International Uppsala University Universität Gesamthochschule Kassel CEA - Centre d'Etudes et de Recherches sur les Matériaux (CEREM) European Commission - JRC/IE Intercontrôle Nuclear Reseach Institute Rež plc (NRI) Tecnatom S.A. E.ON Kernkraft GmbH SQC Swedish NDT Qualification Centre Serco Assurance 45 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations Plant life extension and management On-line monitoring and maintenance Title: REACTOR DOSIMETRY: ACCURATE DETERMINATION AND BENCHMARKING OF RADIATION FIELD PARAMETERS, RELEVANT FOR REACTOR PRESSURE VESSEL MONITORING Acronym REDOS Contract number FIKS-CT2001-00120 Duration 36 months Proposal number FIS5-2001-00004 Type of action Starting date Total budget Shared cost 1 November 2001 916.919 € EC project officer S. Casalta EC contribution 499.949 € Co-ordinator Organisation Address Contact person Tecnatom S.A. Inspection Engineering Division Avda. Montes de Oca, 1 E-28709 San Sebastian de los Reyes (Madrid) Mr. Antonio Ballesteros Tel: (34-91)6598723 Fax: (34-91)6598677 Email aballesteros@tecnatom.es Partnership Country D CZ HU BG INT CZ D Organisations Forschungszentrum Rossendorf e.V. (FZR) Nuclear Research Institute (NRI) KFKI Atomic Energy Research Institute (AEKI) Institute of Nuclear Research and Nuclear Energy (INRNE) European Commission - JRC/IE Skoda JS a.s. 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 Contents 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 mockup 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 Contents Nuclear Energy Programme Operational safety of existing installations Plant life extension and management On-line monitoring and maintenance Title: VIRTUAL REALITY FOR INSPECTION, MAINTENANCE, OPERATION, AND REPAIR OF NUCLEAR POWER PLANT Acronym VRIMOR Contract number FIKS-CT2000-00114 Duration 24 months Proposal number FIS5-1999-00328 Type of action Starting date Total budget Shared cost 1 February 2001 1.198.947 € EC project officer S. Casalta EC contribution 599.474 € Co-ordinator Organisation Address Contact person National Nuclear Corporation Limited (NNC) Simulation Business Team Booths Hall, Chelford Road UK-WA16 8QZ Knutsford, Cheshire Dr. David Lee Tel: (44-156)5843732 Fax: (44-156)5843441 Email david.lee@nnc.co.uk Partnership Country E E E B UK Organisations Tecnatom Universidad Politécnica de Madrid (UPM) Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT) Centre d'Etude de l'Energie Nucléaire (SCK-CEN) 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 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations Plant life extension and management On-line monitoring and maintenance Title: NUCLEAR RISK-BASED INSPECTION METHODOLOGY Acronym NURBIM Contract number FIKS-CT2001-00172 Duration 32 months Proposal number FIS5-2001-00082 Type of action Starting date Total budget Shared cost 1 November 2001 1.209.094 € EC project officer S. Casalta EC contribution 604.546 € Co-ordinator Organisation Address Contact person Gesellschaft für Anlagen - und Reaktorsicherheit (GRS) GmbH Schwertnergasse 1 D-50667 Köln Dr. Helmut Schulz Tel: (49-221)2068603 Fax: (49-221)2068888 Email suh@grs.de Partnership Country F S D S CZ UK UK FIN UK INT E Organisations Electricité de France (EDF) OKG AB E. ON Kernkraft GmbH Det Norske Veritas AB Nuclear Research Institute Rež plc (NRI) OJV Consultancy Ltd The Welding Institute (TWI) Ltd Technical Research Centre of Finland (VTT) Mitsui Babcock Energy Ltd European Commission - JRC/IE 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 Contents 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. 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 1 52 Contents Nuclear Energy Programme Operational safety of existing installations Plant life extension and management On-line monitoring and maintenance Title: MANAGEMENT OF NUCLEAR PLANT OPERATION BY OPTIMISING WELD REPAIRS Acronym ENPOWER Contract number FIKS-CT2001-00167 Duration 36 months Proposal number FIS5-2001-00071 Type of action Starting date Total budget Shared cost 1 December 2001 1.713.251 € EC project officer P. Manolatos EC contribution 919.263 € Co-ordinator Organisation Address Contact person Institut de Soudure 90, rue des Vanesses F-93420 Villepinte Dr. Christian Boucher Tel: (33-1)49903633 Fax: (33-1)49903628 Email c.boucher@institutdesoudure.com Partnership Country UK UK D INT UK F Organisations British Energy Generation Ltd Mitsui Babcock Energy Ltd Framatome ANP GmbH European Commission - JRC/IE University of Bristol 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 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations Plant life extension and management Organisation and Management of safety Title: BENCHMARK EXERCISE ON SAFETY EVALUATION OF COMPUTER-BASED SYSTEMS Acronym BE-SECBS Contract number FIKS-CT2000-00054 Duration 30 months Proposal number FIS5-1999-00216 Type of action Starting date Total budget Shared cost 1 January 2001 791.785 € EC project officer G. Van Goethem EC contribution 395.892 € Co-ordinator Organisation Address Contact person European Commission JRC/IE, Petten Postbus 2 NL-1755 ZG Petten Dr. Christian Kirchsteiger Tel: (31-224) 565118 Fax: (31-224) 565641 Email christian.kirchsteiger@jrc.nl Partnership Country FIN FIN D F D Organisations Radiation and Nuclear Safety Authority (STUK) Technical Research Centre of Finland (VTT) Inst. für Sicherheitstechnologie (ISTec) GmbH Institut de Radioprotection et de Sûreté Nucléaire (IRSN) 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 Contents 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 safetycritical 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 Contents Nuclear Energy Programme Operational safety of existing installations Plant life extension and management Organisation and Management of safety Title: COST EFFECTIVE MODERNISATION OF SYSTEMS IMPORTANT TO SAFETY Acronym CEMSIS Contract number FIKS-CT2000-00109 Duration 36 months Proposal number FIS5-1999-00355 Type of action Starting date Total budget Shared cost 1 January 2001 1.948.547 € EC project officer G. Van Goethem EC contribution 775.000 € Co-ordinator Organisation Address Contact person British Energy (Generation) Ltd. Plant Engineering Branch Bernett Way UK-GL4 3RS Barnwood, Gloucester Mr. Paul Tooley Tel: (44-1552) 653503 Fax: (44-1552) 654897 Email p.tooley@ne.british-energy.com Partnership Country B UK F D S S UK Organisations AIB Vinçotte Nuclear British Nuclear Fuels plc (BNFL) Electricité de France (EDF) Framatome ANP GmbH Sycon Energikonsult AB Lund University 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 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations Plant life extension and management Organisation and Management of safety Title: LEARNING ORGANISATIONS FOR NUCLEAR SAFETY Acronym LearnSafe Contract number FIKS-CT2001-00162 Duration 30 months Proposal number FIS5-2001-00066 Type of action Starting date Total budget Shared cost 1 November 2001 1.175.258 € EC project officer S. Casalta EC contribution 500.801 € Co-ordinator Organisation Address Contact person Technical Research Centre of Finland (VTT) Tekniikantie 12 FIN-02044 Espoo Prof. Björn Wahlström Tel: (358-9)4566400 Fax: (358-9)4566752 Email bjorn.wahlstrom@vtt.fi Partnership Country D UK E S E INT FIN S D D UK S S Organisations Technische Universitaet Berlin (TUB) Loughborough University (LBORO) Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT) SwedPower AB Asociacion Espanola de la Industria Electrica (UNESA) World Association of Nuclear Operators (WANO) Teollisuuden Voima Oy (TVO) Forsmarks Krattgrupp AB E. ON Kernkraft GmbH Kernkraftwerk Krümmel GmbH British Nuclear Fuels plc (BNFL) OKG Aktiebolag Ringhals AB 59 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations Plant life extension and management Organisation and Management of safety Title: EVALUATION OF ALTERNATIVE APPROACHES FOR ASSESSMENT OF SAFETY PERFORMANCE INDICATORS FOR NUCLEAR POWER PLANTS Acronym SPI Contract number FIKS-CT2001-20145 Duration 21 months Proposal number FIS5-2001-00041 Type of action Starting date Total budget Concerted action 1 October 2001 337.850 € EC project officer S. Casalta EC contribution 193.150 € Co-ordinator Organisation Address Contact person Gesellschaft für Anlagen - und Reaktorsicherheit (GRS) GmbH Forschungsgelände D-85748 Garching Dr. Klaus Koberlein Tel: (49-89)32004445 Fax: (49-89)32004306 Email koe@grs.de Partnership Country E CH CH F UK CH S HU S SK Organisations Consejo de Seguridad Nuclear (CSN) ERI Consulting, Khatib, Attenhofer & Co. Swiss Federal Nuclear Safety Inspectorate (HSK) Institut de Radioprotection et de Sureté Nucléaire (IRSN) Health & Safety Executive Nordostschweizerische Kraftwerke (NOK) Swedish Nuclear Power Inspectorate (SKI) Institute for Electric Power Research Co.(VEIKI) Swedpower AB 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 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations Plant life extension and management VVER operational safety issues Title: IMPROVED ACCIDENT MANAGEMENT OF VVER NUCLEAR POWER PLANTS Acronym IMPAM-VVER Contract number FIKS-CT2001-00196 Duration 32 months Proposal number FIS5-2001-00117 Type of action Starting date Total budget Shared cost 1 November 2001 1.169.928 € * EC project officer P. Manolatos EC contribution 699.942 € * Co-ordinator Organisation Address Contact person Technical Research Centre of Finland (VTT) Nuclear Energy Techniikantie 4C FIN-02044 Espoo 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 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations Plant life extension and management VVER operational safety issues Title: UNIFIED PROCEDURE FOR LIFETIME ASSESSMENT OF COMPONENTS AND PIPING IN VVER NPPS Acronym VERLIFE Contract number FIKS-CT2001-20198 Duration 24 months Proposal number FIS5-2001-00120 Type of action Starting date Total budget Thematic network 1 October 2001 220.898 € * EC project officer P. Manolatos EC contribution 220.898 € * Co-ordinator Organisation Address Contact person Ustav Jaderného Vyzkumu Rež A.S. Division of Integrity and Technical Engineering Rež 130 CZ-25068 Rež 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 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations - RI Plant life extension and management VVER operational safety issues Title: AMES THEMATIC NETWORK ON AGEING Acronym ATHENA Contract number FIR-CT2001-20170 Duration 36 months Proposal number FIS5-2001-00079 Type of action Starting date Total budget Thematic network 1 November 2001 380.000 € EC project officer P. Manolatos EC contribution 380.000 € Co-ordinator Organisation Address Contact person Tractebel S.A. Energy Engineering - Operation and Maintenance Dept. Avenue Ariane, 4 B-1200 Brussels Mr. Robert Gerard Tel: (32-2)7738363 Fax: (32-2)7738900 Email robert.gerard@tractebel.be Partnership Country UK UK INT CZ FIN Organisations Magnox Electric plc LMD Consultancy European Commission - JRC/IE Nuclear Research Institute Rež (NRI) 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 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations Title: CORE LOSS DURING A SEVERE ACCIDENT Severe accident management Assessment of risks Acronym COLOSS Contract number FIKS-CT1999-00002 Duration 36 months Proposal number FIS5-1999-00013 Type of action Starting date Total budget Shared cost 1 February 2000 3.186.437 € EC project officer A. Zurita EC contribution 1.600.000 € Co-ordinator Organisation Address Contact person Institut de Radioprotection et de Sûreté Nucléaire (IRSN/DRS) CEA CADARACHE F-13108 Saint-Paul-lez-Durance Dr. Bernard Adroguer Tel: 33 4 42 25 23 34 Fax: 33 4 42 25 29 29 Email bernard.adroguer@irsn.fr Partnership Country HU F I F D INT INT CH D CZ E D D FIN Organisations KFKI Atomic Energy Research Institute (AEKI) Electricité de France (EDF) Ente per le Nuove Tecnologie, l'Energia e l'Ambiente (ENEA) Framatome ANP Forschungszentrum Karlsruhe GmbH (FZK) European Commission - JRC/IE European Commission - JRC/ITU Paul Scherrer Institut (PSI) Framatome ANP GmbH SKODA-UJP Praha a.s. Universidad Politécnica de Madrid (UPM) Ruhr-Universität Bochum (RUB) Universität Stuttgart - IKE University Lappeenranta 69 Contents 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-OZr 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 degradationoxidation 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 riskrelevant 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 workpackages, 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 Contents Nuclear Energy Programme Operational safety of existing installations Title: Severe accident management Assessment of risks LIMIT STRAINS FOR SEVERE ACCIDENT CONDITIONS Acronym LISSAC Contract number FIKS-CT1999-00012 Duration 36 months Proposal number FIS5-1999-00075 Type of action Starting date Total budget Shared cost 1 February 2000 2.685.000 € EC project officer A. Zurita EC contribution 950.000 € Co-ordinator Organisation Address Contact person Forschungszentrum Karlsruhe GmbH (FZK) Postfach 3640 D-76021 Karlsruhe Dr. Rudolf Krieg Tel: 49 7247 82 43 56 Fax: 49 7247 82 37 18 Email maeule@irs.fzk.de Partnership Country I E F SI INT D NL CH D EL FIN Organisations Ente per le Nuove Tecnologie, l'Energia e l'Ambiente (ENEA) Equipos Nucleares S.A. Framatome ANP Institut "Josef Stefan" European Commission - JRC/IPSG Staatliche Materialprufungsanstalt (MPA Stuttgart) Nuclear Research and Consultancy Group (NRG) Paul Scherrer Institute (PSI) Framatome ANP GmbH University Aristotle of Thessaloniki 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 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations Title: ASSESSMENT OF REACTOR VESSEL INTEGRITY Severe accident management Assessment of risks Acronym ARVI Contract number FIKS-CT1999-00011 Duration 36 months Proposal number FIS5-1999-00006 Type of action Starting date Total budget Shared cost 1 January 2000 1.042.444 € EC project officer A. Zurita EC contribution 700.000 € Co-ordinator Organisation Address Contact person Kungal Tekniska Högskalan Nucl. Power Safety Div. of Dept. of Energy Technol. Drotning Kristina Vag 33A S-10044 Stockholm Prof. Bal Raj Sehgal Tel: 46 8 790 92 52 Fax: 46 8 790 9197 Email sehgal@ne.kth.se Partnership Country F FIN F CZ D US HU FIN Organisations Commissariat à l' Energie Atomique (CEA/DRN/DTP) Fortum Nuclear Services Ltd Framatome ANP Nuclear Reseach Institute Rež plc (NRI) Universität Stuttgart (IKE) University Regents of California Institute for Electric Power Research Co.(VEIKI) 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 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations Title: Severe accident management Assessment of risks EUROPEAN NUCLEAR THERMODYNAMIC DATABASE (FOR IN- AND EX-VESSEL APPLICATIONS) Acronym ENTHALPY Contract number FIKS-CT1999-00001 Duration 36 months Proposal number FIS5-1999-00001 Type of action Starting date Total budget Shared cost 1 February 2000 1.125.147 € EC project officer A. Zurita EC contribution 599.863 € Co-ordinator Organisation Address Contact person Institut de Radioprotection et de Sûreté Nucléaire (IRSN/DRS) BP 1 F-13108 Saint-Paul-lez-Durance 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 UK HU F F F B D CZ F B B Organisations AEA Technology Plc KFKI Atomic Energy Research Institute (AEKI) Commissarait à l'Energie Atomique (CEA/DRN/DTP) Commissarait à l'Energie Atomique (CEA/DTA/CEREM) Electricité de France (EDF) Belgian Nuclear Research Centre (SCK-CEN) Framatome ANP GmbH SKODA - UJP Praha a.s. Thermodata Université Catholique de Louvain (UCL) 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 Contents deduced from the phase diagrams of the elements and systems present in the in- and exvessel 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)-LaSrRu 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 exvessel 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 Contents Nuclear Energy Programme Operational safety of existing installations Title: Severe accident management Assessment of risks EX-VESSEL CORE MELT STABILISATION RESEARCH Acronym ECOSTAR Contract number FIKS-CT1999-00003 Duration 48 months Proposal number FIS5-1999-00016/70 Type of action Starting date Total budget* Shared cost 1 January 2000 4.439.000 € EC project officer A. Zurita EC contribution* 2.399.000 € Co-ordinator Organisation Address Contact person Forschungszentrum Karlsruhe GmbH (FZK) Postfach 3640 D-76021 Karlsruhe Mr. Werner Scholtyssek Tel: 49 7247 82 55 25 Fax: 49 7247 82 55 08 Email werner.scholtyssek@psf.fzk.de Partnership Country F F I D D CZ D D D F D S D Organisations Commissariat à l' Energie Atomique (CEA/DEN) Electricité de France (EDF) Ente per le Nuove Tecnologie, l'Energia e l'Ambiente (ENEA) Forschungszentrum Rossendorf e.V. (FZR) Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH Nuclear Reseach Institute Rež plc (NRI) Framatome ANP GmbH Ruhr-Universität Bochum (RUB) Universität Stuttgart (IKE) Université de Provence Universität Aachen Royal Institute of Technology (KTH/RIT) 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 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations Title: Severe accident management Assessment of risks INTEGRAL LARGE SCALE EXPERIMENTS ON HYDROGEN COMBUSTION FOR SEVERE ACCIDENT CODE VALIDATION Acronym HYCOM Contract number FIKS-CT1999-00004 Duration 36 months Proposal number FIS5-1999-00017 Type of action Starting date Total budget Shared cost 1 February 2000 1.420.521 € EC project officer G. Van Goethem EC contribution 700.000 € Co-ordinator Organisation Address Contact person Forschungszentrum Karlsruhe GmbH (FZK) Postfach 3640 D-76021 Karlsruhe Mr. Werner Scholtyssek Tel: 49 7247 82 55 25 Fax: 49 7247 82 55 08 Email werner.scholtyssek@psf.fzk.de Partnership Country D F INT RU D Organisations Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH Institut de Radioprotection et de Sûreté Nucléaire (IRSN) European Commission - JRC/IE Petten Kurchatov Institute 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 Contents 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 crosssection 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 Contents Nuclear Energy Programme Operational safety of existing installations Title: Severe accident management Assessment of risks EUROPEAN VALIDATION OF THE INTEGRAL CODE ASTEC Acronym EVITA Contract number FIKS-CT1999-00010 Duration 36 months Proposal number FIS5-1999-00062 Type of action Starting date Total budget Shared cost 1 February 2000 2.279.380 € EC project officer G. Van Goethem EC contribution 1.199.847 € Co-ordinator Organisation Address Contact person Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH Postfach 10 15 64 / Schwertnergasse 1 D-50667 Köln Dr. Hans-Josef Allelein Tel: 49.221 206.86.68 Fax: 49.221 206.88.88 Email all@grs.de Partnership Country A D F E I F D F SK INT CZ D SK D D HU SK F Organisations Austrian Research Centre Seibersdorf (ARCS) Becker Technologies GmbH Electricité de France (EDF) Empresarios Agrupados (EA) Ente per le Nuove Tecnologie, l'Energia e l'Ambiente (ENEA) Framatome ANP Forschungszentrum Karlsruhe GmbH (FZK) Institut de Radioprotection et de Sûreté Nucléaire (IRSN) Inzinierska Vypoctova Spolocnoast (IVS) Trnava European Commission - JRC/IE Nuclear Reseach Institute Rež plc (NRI) Framatome ANP GmbH Urad Jadroveho dozoru Slovenskej republiky (UJD SR) Ruhr-Universität Bochum (RUB) Universität Stuttgart (IKE) Institute for Electric Power Research Co.(VEIKI) Nuclear Power Plant Research Institute (VUJE) Trnava Commissariat à l'Energie Atomique (CEA) 81 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations Title: LATE PHASE SOURCE TERM PHENOMENA Severe accident management Assessment of risks Acronym LPP Contract number FIKS-CT1999-00005 Duration 36 months Proposal number FIS5-1999-00023 Type of action Starting date Total budget Shared cost 1 February 2000 1.341.519 € EC project officer A. Zurita EC contribution 749.998 € Co-ordinator Organisation Address Contact person AEA Technology PLC Winfrith Technology Centre UK-DT2 8DH Dorchester, Dorset Dr Christopher Benson Tel: 44.13.05. 20 2751 Fax: 44.13.05.20.26.63 Email christopher.benson@aeat.co.uk Partnership Country F RU CZ D D Organisations Institut de Radioprotection et de Sûreté Nucléaire (IRSN) Leningrad Special Integrated Plant "Radon" Nuclear Reseach Institute Rež plc (NRI) Framatome ANP GmbH 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 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations Title: Severe accident management Assessment of risks VALIDATION OF SEVERE ACCIDENT CODES AGAINST PHEBUS FP FOR PLANT APPLICATIONS Acronym PHEBEN 2 Contract number FIKS-CT1999-00009 Duration 48 months Proposal number FIS5-1999-00057 Type of action Starting date Total budget Shared cost 1 March 2000 1.271.048 € * EC project officer A. Zurita EC contribution 619.078 € * Co-ordinator Organisation Address Contact person European Commission - Joint Research Centre - IE Postbus 2 NL-1755 ZG Petten Dr Alan Victor Jones Tel: 39.0332.78.96.29 Fax: 39.0332.78.58.15 Email alan.jones@jrc.it Partnership Organisations AEA Technology Plc Commissariat à l' Energie Atomique (CEA/DRN/DTP) 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). Country UK F E 85 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations Title: Severe accident management Assessment of risks ARCHIVE MODELS FOR SOURCE TERM INFORMATION AND SYSTEM MODELS Acronym ASTERISM II Contract number FIKS-CT1999-20001 Duration 18 months Proposal number FIS5-1999-00019 Type of action Starting date Total budget Concerted action 1 February 2000 299.178 € EC project officer A. Zurita EC contribution 299.178 € Co-ordinator Organisation Address Contact person AEA Technology PLC Winfrith Technology Centre UK-DT2 8DH Dorchester, Dorset Dr. Ann Tuson Tel: 44.1305.20.21.95 Fax: 44.1305.20.25.08 Email ann.tuson@aeat.co.uk Partnership Country INT UK Organisations European Commission - JRC/IPSC 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 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations Title: Severe accident management Assessment of risks EUROPEAN EXPERT NETWORK FOR THE REDUCTION OF UNCERTAINTIES IN SEVERE ACCIDENT SAFETY ISSUES Acronym EURSAFE Contract number FIKS-CT2001-20147 Duration 24 months Proposal number FIS5-2001-00044/46 Type of action Starting date Total budget Thematic network 1 December 2001 540.046 € EC project officer A. Zurita EC contribution 400.000 € Co-ordinator Organisation Address Contact person Institut de Radioprotection et de Sûreté Nucléaire (IRSN) Bât. 219 F-13108 Saint-Paul-lez-Durance Dr. Daniel Magallon Tel: (33-4)42254920 Fax: (33-4)42256465 Email magallon@drncad.cea.fr Partnership Country F F FIN UK CZ D D D D S HU E E INT HU UK CH US Organisations Commissariat à l'Energie Atomique (CEA) Electricité de France (EDF) Teollisuuden Voima Oy (TVO) AEA Technology Plc Nuclear Reseach Institute Rež plc (NRI) Forschungszentrum Karlsruhe GmbH (FZK) Universität Stuttgart Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH Framatome ANP GmbH Royal Institute of Technology (KTH/RIT) KFKI Atomic Energy Research Institute (AEKI) Consejo de Seguridad Nuclear (CSN) Universidad Politécnica de Madrid (UPM) European Commission - JRC/IE Institute for Electric Power Research (VEIKI) Health and Safety Executive (HSE) Paul Scherrer Institut (PSI) Nuclear Regulatory Commission 89 Contents 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 (InVessel 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 Contents Nuclear Energy Programme Operational safety of existing installations Title: Severe accident management Assessment of risks THEMATIC NETWORK FOR A PHEBUS FPT-1 INTERNATIONAL STANDARD PROBLEM Acronym THENPHEBISP Contract number FIKS-CT2001-20151 Duration 24 months Proposal number FIS5-2001-00048 Type of action Starting date Total budget Thematic network 1 December 2001 240.899 € * EC project officer A. Zurita EC contribution 240.899 € * Co-ordinator Organisation Address Contact person Institut de Radioprotection et de Sûreté Nucléaire (IRSN) Dept. de Recherches en Sûreté (DRS) Centre d'Etudes Nucléaires Bat. 702 F-13108 Saint-Paul-lez-Durance Dr. Bernard Clément Tel: (33-4)42257646 Fax: (33-4)42252929 Email bernard.clement@irsn.fr Partnership Country INT UK HU CZ I I E D D CH B F SI BG ____________________ Organisations European Commission - JRC/IE AEA Technology Plc. KFKI Atomic Energy Research Institute (AEKI) Nuclear Reseach Institute Rež plc (NRI) Ente per le Nuove Tecnologie, l'Energia e l'Ambiente (ENEA) University of Pisa Universidad Politecnica de Madrid (UPM) Forschungszentrum Karlsruhe GmbH (FZK) Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH Paul Scherrer Institute (PSI) Belgian Nuclear Research Centre (SCK-CEN) Electricité de France (EDF) Jozef Stefan Institute (*) Enpro Consulting (*) * Budget expanded following inclusion of new partners from Newly Associated States (NAS). 91 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations - RI Title: SCALING OF CONTAINMENT EXPERIMENTS Severe accident management Assessment of risks Acronym SCACEX Contract number FIR1-CT2001-20127 Duration 12 months Proposal number FIS5-2001-00017 Type of action Starting date Total budget Thematic network 1 January 2002 372.812 € EC project officer A. Zurita EC contribution 372.812 € Co-ordinator Organisation Address Contact person Becker Technologies GmbH Koelnerstrasse 6 D-65760 Eschborn Dr. Karsten Fischer Tel: (49-6196)936116 Fax: (49-6196)936100 Email fischer@becker-technologies.com Partnership Country D D UK F I I CZ D SK E F Organisations Framatome ANP GmbH Forschungszentrum Karlsruhe GmbH (FZK) The Victoria University of Manchester Institut de Radioprotection et de Sûreté Nucléaire (IRSN) Università di Pisa Università di Roma Nuclear Research and Consultancy Group (NRG) Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH Nuclear Power Plant Research Institute (VUJE) Trnava Inc Empresarios Agrupados Internacional SA 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 Contents - 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 Contents Nuclear Energy Programme Operational safety of existing installations - RI Title: Severe accident management Assessment of risks PLATFORM FOR IMPROVEMENTS IN NUCLEAR INDUSTRY AND UTILITY SAFETY Acronym PLINIUS Contract number FIR1-CT2001-40152 Duration 36 months Proposal number FIS5-2001-00049 Type of action Starting date Total budget Research infrastructure 1 December 2001 699.300 € EC project officer A. Zurita EC contribution 699.300 € Co-ordinator Organisation Commissariat à l'Energie Atomique (CEA) - Direction de l'Energie Nucléaire Dept. Thermohydraulique et Physique BP1 F-13108 Saint-Paul-lez-Durance Mr. Christophe Journeau Tel: (33-4)42254121 Fax: (33-4)42256465 Email cjourneau@cea.fr Address Contact person 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 50100 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 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations - RI Title: Severe Accident Management Assessment of risks LARGE SCALE EXPERIMENTS ON CORE DEGRADATION, MELT RETENTION AND COOLABILITY Acronym LACOMERA Contract number FIR1-CT2002-40158 Duration 36 months Proposal number FIS5-2002-00007 Type of action Starting date Total budget Research Infrastructures 1 September 2002 1.250.000 € EC project officer A. Zurita EC contribution 500.000 € Co-ordinator Organisation Address Contact person Forschungszentrum Karlsruhe, GmbH, Institut fuer Nukleare Entsorgungstechnik Hermann-von-Helmholtz-Platz, 1; PO Box 3640 D-76021 Karlsruhe 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 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations Title: Severe accident management Severe accident management EUROPEAN GROUP FOR ANALYSIS OF CORIUM RECOVERY CONCEPTS Acronym EUROCORE Contract number FIKS-CT1999-20003 Duration 24 months Proposal number FIS5-1999-00050 Type of action Starting date Total budget Thematic network 1 March 2000 385.404 € EC project officer A. Zurita EC contribution 385.404 € Co-ordinator Organisation Address Contact person Commissariat a l'Energie Atomique (CEA) DRN / DTP 17 rue des Martyrs F-38054 Grenoble Cedex 9 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 UK E F FIN F D D D S Organisations Serco Assurance Análisis y Soluciones Tecnológicas (AST) Electricité de France (EDF) Fortum Nuclear Services Ltd Framatome ANP Forschungszentrum Karlsruhe GmbH (FZK) Framatome ANP GmbH Universität Stuttgart (IKE) 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 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations Title: STEAM GENERATOR TUBE RUPTURE SCENARIOS Severe accident management Severe accident management Acronym SGTR Contract number FIKS-CT1999-00007 Duration 36 months Proposal number FIS5-1999-00031 Type of action Starting date Total budget Shared cost 1 January 2000 2.260.000 € EC project officer A. Zurita EC contribution 800.000 € Co-ordinator Organisation Address Contact person Technical Research Centre of Finland (VTT) P.O. Box 1401 / Biologinkuja 7 FIN-02044 Espoo VTT Dr. Jorma Jokiniemi Tel: 358 9 456.61.58 Fax: 358 9 456.70.21 Email jorma.jokiniemi@vtt.fi Partnership Country E FIN NL CZ CH Organisations Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT) Fortum Nuclear Services Ltd Nuclear Research and Consultancy Group (NRG) Nuclear Reseach Institute Rež plc (NRI) 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 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations Title: IODINE CHEMISTRY AND MITIGATION METHODS Severe accident management Severe accident management Acronym ICHEMM Contract number FIKS-CT1999-00008 Duration 36 months Proposal number FIS5-1999-00038 Type of action Starting date Total budget* Shared cost 1 February 2000 1.244.696 € EC project officer A. Zurita EC contribution* 548.707 € Co-ordinator Organisation Address Contact person AEA Technology plc A32, Winfrith Technology Centre UK-DT2 8DH Dorchester, Dorset Dr. Shirley Dickinson Tel: 44-1305-20 28 55 Fax: 44-1305-20 26 63 Email shirley.dickinson@aeat.co.uk Partnership Country F INT CZ CH D S Organisations Institut de Radioprotection et de Sûreté Nucléaire (IRSN) European Commission - JRC/IE Nuclear Reseach Institute Rež plc (NRI) Paul Scherrer Institut (PSI) Framatome ANP GmbH 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. * Total eligible costs and EC contribution reduced respectively to 1.187.856 € and 521.879 € following the changes in the technical annex of the project. 103 Contents 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 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 Contents Nuclear Energy Programme Operational safety of existing installations Title: Severe accident management Severe accident management HYDROGEN REMOVAL FROM LWR CONTAINMENTS BY CATALYTIC COATED THERMAL INSULATION ELEMENTS Acronym THINCAT Contract number FIKS-CT1999-00006 Duration 28,5 months Proposal number FIS5-1999-00029 Type of action Starting date Total budget* Shared cost 1 January 2000 1.200.000 € EC project officer A. Zurita EC contribution* 600.000 € Co-ordinator Organisation Address Contact person Forschungszentrum Jülich (FZJ) ISR - 2 D-52425 Jülich Dr. Ernst-Arndt Reinecke Tel: 49.2461.615530 Fax: 49.2461.614059 Email e.-a.reinecke@fz-juelich.de Partnership Country E E D S Organisations Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT) Empresarios Agrupados Kaefer Isoliertechnik GmbH & Co 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 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations Title: Severe accident management Severe accident management HYDROGEN HAZARD - PASSIVE AUTOCATALYTIC RECOMBINERS STATE-OF-THEART Acronym PARSOAR Contract number FIKS-CT1999-20002 Duration 24 months Proposal number FIS5-1999-00030 Type of action Starting date Total budget Thematic network 1 February 2000 150.000 € EC project officer G. Van Goethem EC contribution 150.000 € Co-ordinator Organisation Address Technicatome BP 34000 / 1100, Avenue Jean-René Guillibert Gautier De La Lauzière F-13791 Aix-en-Provence Cedex 03 Mr. François Arnould Tel: 33 4 42 60 28 50 Fax: 33 442 60 25 11 Email arnouldf@ta-aix.tecatom.fr Contact person Partnership Country B CA F CH D Organisations Association Vinçotte Nuclear Atomic Energy of Canada Limited Commissariat à l'Energie Atomique (CEA/DRN/DTP) Electrowatt Engineering Ltd 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 Contents − 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 Contents Nuclear Energy Programme Operational safety of existing installations Title: Severe accident management Severe accident management OPTIMISATION OF SEVERE ACCIDENT MANAGEMENT STRATEGIES FOR THE CONTROL OF RADIOLOGICAL RELEASES Acronym OPTSAM Contract number FIKS-CT1999-00013 Duration 24 months Proposal number FIS5-1999-00074 Type of action Starting date Total budget Shared cost 1 June 2000 1.011.260 € EC project officer A. Zurita EC contribution 499.843 € Co-ordinator Organisation Address Contact person National Nuclear Corporation Limited (NNC) Booths Hall, Chelford Road UK-WA16 8QZ Knutsford, Cheshire 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 F FIN D E CZ S S B HU Organisations Electricité de France (EDF) Fortum Nuclear Services Ltd Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH Iberdrola Nuclear Reseach Institute Rež plc (NRI) Swedpower AB Sycon Energikonsult Tractebel S.A. 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 Contents 2. 3. To establish a comprehensive technical basis for the definition of source terms for operating reactors across Europe. 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 Contents Nuclear Energy Programme Operational safety of existing installations Title: A PERSPECTIVE ON OPERATOR SUPPORT COMPUTERIZED Severe accident management Severe accident management SEVERE ACCIDENT MANAGEMENT Acronym SAMOS Contract number FIKS-CT2001-20189 Duration 18 months Proposal number FIS5-2001-00104 Type of action Starting date Total budget Thematic network 1 December 2001 177.044 € EC project officer A. Zurita EC contribution 107.160 € Co-ordinator Organisation Address Contact person Nuclear Service Corporation Netherlands (NSC) Akenwerf 35 NL-2317DK Leiden Dr. George Vayssier Tel: (31-71)523245 Fax: (31-71)5232341 Email vayssier@hetnet.nl Partnership Country B SK E SI Organisations Westinghouse Electric Europe Nuclear Regulatory Authority of the Slovak Republic Tecnatom 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 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations Title: Severe accident management Severe accident management CONCERTED UTILITY REVIEW OF VVER-440 SAFETY RESEARCH NEEDS Acronym VERSAFE Contract number FIKS-CT2000-20044 Duration 24 months Proposal number FIS5-1999-00181 Type of action Starting date Total budget Concerted action 1 September 2000 186.000 € EC project officer P. Manolatos EC contribution 186.000 € Co-ordinator Organisation Address Contact person Fortum Nuclear Services Ltd Nuclear Power Rajatorpantie 8 FIN-00048 Vantaa Dr. Harri Tuomisto Tel: (358-1045)32464 Fax: (358-1045)92464 Email harri.tuomisto@fortum.com Partnership Country HU CZ SK Organisations PAKS Nuclear Power Plant Ltd. Czech Energy Company, Inc 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 Contents 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 plantspecific 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 Contents Nuclear Energy Programme Operational safety of existing installations Title: Evolutionary concepts Evolutionary safety concepts ADVANCED THREE-DIMENSIONAL TWO-PHASE FLOW SIMULATION TOOL FOR APPLICATION TO REACTOR SAFETY Acronym ASTAR Contract number FIKS-CT2000-00050 Duration 36 months Proposal number FIS5-1999-00204 Type of action Starting date Total budget Shared cost 1 September 2000 1.888.691 € EC project officer G. Van Goethem EC contribution 799.723 € Co-ordinator Organisation Address Contact person Commissariat à l'Energie Atomique (CEA) Department of Mechanics and Technology CEA Saclay, DMT / Sysco F-91191 Gif-sur-Yvette Cedex Dr. Henri Paillere Tel: (33-1) 69088409 Fax: (33-1) 69089696 Email henri.paillere@cea.fr Partnership Country F INT D UK B CH Organisations Electricité de France (EDF) European Commission - JRC/IE Petten Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH Manchester Metropolitan University Von Karman Institute for Fluid Dynamics (VKI) Paul Scherrer Institute (PSI) Project Summary The objectives of the ASTAR project are to substantially enhance the three-dimensional twophase 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 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations Title: Evolutionary concepts Evolutionary safety concepts TESTING AND ENHANCED MODELLING OF PASSIVE EVOLUTIONARY SYSTEMS TECHNOLOGY (FOR CONTAINMENT COOLING) Acronym TEMPEST Contract number FIKS-CT2000-00095 Duration 36 months Proposal number FIS5-1999-00273/317 Type of action Starting date Total budget Shared cost 1 December 2000 3.238.785 € EC project officer G. Van Goethem EC contribution 1.000.015 € Co-ordinator Organisation Address Contact person Nuclear Research and Consultancy Group (NRG) Plant, Performance and Technology Westerduinweg 3 NL-1755 ZG Petten Mr. Victor A. Wichers Tel: (31-224)564656 Fax: (31-224)563490 Email wichers@nrg-nl.com Partnership Country F CH FIN D D D US US Organisations Commissariat à l'Energie Atomique (CEA) Paul Scherrer Institute (PSI) Technical Research Centre of Finland (VTT) Gesellschaft für Anlagen - und Reaktorsicherheit (GRS) mbH Forschungszentrum Karlsruhe GmbH (FZK) Framatome ANP GmbH University of California, Berkeley 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 Contents 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 steamhydrogen 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 (CFXWAVCO). 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 Contents Nuclear Energy Programme Operational safety of existing installations Title: Evolutionary concepts Evolutionary safety concepts EVALUATION OF COMPUTATIONAL FLUID DYNAMIC METHODS FOR REACTOR SAFETY ANALYSES Acronym ECORA Contract number FIKS-CT2001-00154 Duration 36 months Proposal number FIS5-2001-00051/67 Type of action Starting date Total budget Shared cost 1 October 2001 1.623.803 € EC project officer G. Van Goethem EC contribution 753.480 € Co-ordinator Organisation Address Gesellschaft für Anlagen - und Reaktorsicherheit (GRS) GmbH Thermalhydraulics and Process Engineering PO Box 1328 Forschungsgelände D-85739 Garching Mrs. Martina Scheuerer Tel: (49-89)32004401 Fax: (49-89)32004599 Email bam@grs.de Contact person Partnership Country DE HU F F D NL CZ CH S FIN UK Organisations AEA Technology GmbH KFKI Atomic Energy Research Institute (AEKI) Commissariat à l'Energie Atomique (CEA) Electricité de France (EDF) Forschungszentrum Rossendorf e.V. (FZR) Nuclear Research and Consultancy Group (NRG) Nuclear Research Institute Rež plc (NRI) Paul Scherrer Institute (PSI) Vattenfall Utveckling AB Technical Research Centre of Finland (VTT) Serco Assurance/Serco Ltd 119 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations Title: Evolutionary concepts Evolutionary safety concepts EUROPEAN GROUP FOR FUTURE ADVANCES IN SCIENCES AND TECHNOLOGY FOR NUCLEAR ENGINEERING THERMALHYDRAULICS Acronym EUROFASTNET Contract number FIKS-CT2000-20100 Duration 18 months Proposal number FIS5-1999-00324 Type of action Starting date Total budget Concerted action 1 September 2000 318.725 € EC project officer G. Van Goethem EC contribution 199.988 € Co-ordinator Organisation Address Contact person Commissariat à l'Énergie Atomique (CEA) Departement de Thermohydraulique et Physique 17 Rue des Martyrs F-38054 Grenoble Cedex 9 Dr. Dominique Bestion Tel: (33-4)76883645 Fax: (33-4)76885179 Email dominique.bestion@cea.fr Partnership Country CH I D F FIN SI UK UK INT CZ I F F Organisations Paul Scherrer Institute (PSI) Università degli Studi di Pisa Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH Electricité de France (EDF) Lappeenranta University of Technology Institut "Jozef Stefan" AEA Technology The Imperial College of Science, Technology and Medicine (ICSTM) European Commission - JRC/IE Petten Nuclear Research Institute (NRI) Rež plc Ente per le Nuove Tecnologie, l'Energia e l'Ambiente (ENEA) Framatome ANP S.A. Institut de Radioprotection et de Sûreté Nucléaire (IRSN) 121 Contents 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 thermalhydraulic, 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 Contents Nuclear Energy Programme Operational safety of existing installations Title: Evolutionary concepts Evolutionary safety concepts REVISITING CRITICAL ISSUES IN NUCLEAR REACTOR DESIGN / SAFETY BY USING 3-D NEUTRONICS / THERMALHYDRAULICS MODELS: STATE-OF-THE-ART Acronym CRISSUE-S Contract number FIKS-CT2001-00185 Duration 24 months Proposal number FIS5-2001-00099 Type of action Starting date Total budget Shared cost 1 January 2002 288.489 € EC project officer G. Van Goethem EC contribution 150.000 € Co-ordinator Organisation Address Contact person University of Pisa Dept. of Mechanical, Nuclear & Production Engineering Via Diotisalvi I-56100 Pisa Prof. Francesco d'Auria Tel: (39-050)836653 Fax: (39-050)836665 Email dauria@ing.unipi.it Partnership Country E S S E E CZ Organisations Asociacion Nuclear Asco Vandellos Studsvik Eco & Safety AB Swedish Nuclear Power Inspectorate (SKI) Universidad Politécnica de Madrid (UPM) Universidad Politécnica de Valencia Nuclear Research Institute Rež plc (NRI) Project Summary The CRISSUE-S project deals with the techniques of applying coupled 3-D neutronicsthermalhydraulics 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 Contents 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 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations Title: Evolutionary concepts Evolutionary safety concepts VALIDATION OF COUPLED NEUTRONICS/THERMAL HYDRAULICS CODES FOR VVER REACTORS Acronym VALCO Contract number FIKS-CT2001-00166 Duration 24 months Proposal number FIS5-2001-00070 Type of action Starting date Total budget Shared cost 1 January 2002 1.092.045 € EC project officer G. Van Goethem EC contribution 672.902 € Co-ordinator Organisation Address Contact person Forschungszentrum Rossendorf E.V. (FZR) Institute of Safety Research D-01314 Dresden Prof. Frank-Peter Weiss Tel: (49-351)2603480 Fax: (49-351)2603440 Email F.P.Weiss@fz-rossendorf.de Partnership Country D FIN HU CZ SK BG UA SK SK RU Organisations Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH Technical Research Centre of Finland (VTT) KFKI Atomic Energy Research Institute (AEKI) Nuclear Research Institute Rež plc (NRI) Nuclear Power Plant Research Institute (VUJE) Trnava Inc Institute of Nuclear Research and Nuclear Energy State Scientific and Technical Centre on Nuclear and Radiation Safety/KIEV Slovenske Elektrarne, a.s. (EBO), BOHUNICE Slovenske Elektrarne, a.s. (EBO), MOHOVCE 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 Contents 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, KIKO3DATHLET, 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 Contents Nuclear Energy Programme Operational safety of existing installations Title: Evolutionary concepts Evolutionary safety concepts RELIABILITY METHODS FOR PASSIVE SAFETY FUNCTIONS Acronym RMPS Contract number FIKS-CT2000-00073 Duration 36 months Proposal number FIS5-1999-00250 Type of action Starting date Total budget Shared cost 1 February 2001 944.581 € * EC project officer S. Casalta EC contribution 550.000 € * Co-ordinator Organisation Address Contact person Commissariat à l'Énergie Atomique (CEA) DRN/DER/SSAE/LAER C.E. Cadarache - Building 238 F-13108 Saint-Paul-lez-Durance Mr. Flavio De Magistris Tel: (33)442256336 Fax: (33)442252408 Email flavio.magistris@cea.fr Partnership Country I Organisations 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 Contents 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 TH 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 Contents Nuclear Energy Programme Operational safety of existing installations Title: Evolutionary concepts Evolutionary safety concepts NATURAL CIRCULATION AND STABILITY PERFORMANCE OF BWRS Acronym NACUSP Contract number FIKS-CT2000-00041 Duration 48 months Proposal number FIS5-1999-00175 Type of action Starting date Total budget Shared cost 1 December 2000 2.973.033 € EC project officer S. Casalta EC contribution 1.197.508 € Co-ordinator Organisation Address Contact person Nuclear Research and Consultancy Group (NRG) Plant Performance & Technology Westerduinweg 3 NL-1755 ZG Petten Dr. Kees Ketelaar Tel: (31-224) 564342 Fax: (31-224) 563490 Email ketelaar@nrg-nl.com Partnership Country F NL CH S D E CH E Organisations Commissariat à l'Energie Atomique (CEA) Delft University of Technology Swiss Federal Institute of Technology Zuerich Forsmarks Kraftgrupp AB Forschungszentrum Rossendorf e.V. (FZR) Iberdrola, S.A. Paul Scherrer Institut (PSI) 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 Contents 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 naturalcirculation 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 lowpower 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 naturalcirculation 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 Contents Nuclear Energy Programme Operational safety of existing installations Title: Evolutionary concepts Evolutionary safety concepts DESIGN AND DEVELOPMENT OF A STEAM GENERATOR EMERGENCY FEEDWATER PASSIVE SYSTEM FOR EXISTING AND FUTURE PWR'S USING ADVANCED STEAM INJECTORS Acronym DEEPSSI Contract number FIKS-CT2000-00113 Duration 36 months Proposal number FIS5-1999-00262 Type of action Starting date Total budget Shared cost 1 December 2000 1.568.840 € EC project officer S. Casalta EC contribution 700.000 € Co-ordinator Organisation Address Contact person Commissariat à l'Énergie Atomique (CEA) DRN / DER / SERSI CEA Cadarache - Building 212 F-13108 Saint-Paul-lez-Durance Cedex Dr. Patrick Dumaz Tel: (33) 4 42 25 40 98 Fax: (33) 4 42 25 40 46 Email patrick.dumaz@cea.fr Partnership Country CZ I I I PL Organisations Nuclear Research Institute Rež plc (NRI) Centro Electrotecnico Sperimentales Italiano (CESI) Ente per le Nuove Tecnologie, l'Energia e l'Ambiente (ENEA) Società Italiana Esperienze Termoidrauliche (SIET) S.p.A. 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 convergingdiverging 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 Contents 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 twophase 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 Contents Nuclear Energy Programme Operational safety of existing installations Title: FAST-ACTING BORON INJECTION SYSTEM Evolutionary concepts Evolutionary safety concepts Acronym FABIS Contract number FIKS-CT2001-00195 Duration 24 months Proposal number FIS5-2001-00116 Type of action Starting date Total budget Shared cost 1 September 2001 602.891 € EC project officer S. Casalta EC contribution 348.962 € Co-ordinator Organisation Address Contact person Technical Research Center of Finland (VTT) Tekniikantie 4C FIN-02044 VTT Espoo Dr. Jari Tuunanen Tel: (358-9)4565081 Fax: (358-9)4565000 Email jari.tuunanen@vtt.fi Partnership Country FIN D Organisations Lappeenranta University of Technology 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 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations - RI Title: Evolutionary concepts Evolutionary safety concepts EUROPEAN NETWORK FOR THE CONSOLIDATION OF THE INTEGRAL SYSTEM EXPERIMENTAL DATA BASES FOR REACTOR THERMAL-HYDRAULIC SAFETY ANALYSIS Acronym CERTA Contract number FIR1-CT2000-20052 Duration 36 months Proposal number FIS5-1999-00213 Type of action Starting date Total budget Thematic network 1 October 2000 297.800 € EC project officer G. Van Goethem EC contribution 250.000 € Co-ordinator Organisation Address Contact person European Commission (EC) JRC/IE Via Enrico Fermi 1 I-21020 Ispra (VA) Mr. Carmelo Addabbo Tel: (39-0332) 789812 Fax: (39-0332) 785584 Email carmelo.addabbo@jrc.it Partnership Country I D D HU FIN I F S CH Organisations Universita' degli Studi di Pisa Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH Framatome ANP GmbH KFKI Atomic Energy Research Institute (AEKI) Technical Research Centre of Finland (VTT) Società Italiana Esperienze Termoidrauliche (SIET S.p.A.) Commissariat à l'Energie Atomique (CEA) Studsvik Eco & Safety AB 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 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations - RI Title: Evolutionary concepts Evolutionary safety concepts IMPROVEMENT OF TECHNIQUES FOR MULTISCALE MODELLING OF IRRADIATED MATERIALS Acronym ITEM Contract number FIR1-CT2001-20136 Duration 48 months Proposal number FIS5-2001-00031 Type of action Starting date Total budget Thematic network 1 November 2001 442.933 € EC project officer G. Van Goethem EC contribution 397.733 € Co-ordinator Organisation Address Contact person Electricité de France (EDF) Departement Etude des Matériaux EDF Site des Renardières F-77818 Moret sur Loing Prof. Jean-Claude Van Duysen Tel: (33-1)60736813 Fax: (33-1)60737369 Email jean-claude.van-duysen@edf.fr Partnership Country F UK E F I B F E UK FIN D RU RU S F FIN Organisations Commissariat à l'Energie Atomique (CEA) University of Liverpool Universidad Complutense de Madrid Centre National de la Recherche Scientifique/ONERA Ente per le Nuove Tecnologie, l'Energia e l'Ambiente (ENEA) Belgian Nuclear Research Centre (SCK-CEN) Université de technologie de Compiègne Universidad Politecnica de Catalunya University of Edinburgh University of Helsinki Max-Planck-Institut für Metallforschung Russian Research Centre "Kurchatov Institute" Ioffe Institut Kungliga tekniska högskolan Laboratoire de Recherches sur la Réactivité des Solides GPM2 Unité Mixte de Recherche 138 Contents F F CZ RU I NL F F UK E B E CH CH DK E RU D D Ecole Nationale Supérieure des Mines de St-Etienne LTPCM Unité Mixte de Recherche Charles University in Prague St-Petersburg State Technical University Instituto Nazionale per la Fisica della materia University of Groningen Ecole Centrale de Paris Centre National de la Recherche Scientifique Kings College London Universidad Politecnica de Madrid (UPM) Université Libre de Bruxelles (ULB) Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT) Paul Scherrer Institut (PSI) CRPP Fusion Technology - Materials Risoe National Laboratory University of Seville State Research Center of Russian Federation - Troitsk Institute Hahn-Meitner-Institut Berlin GmbH 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 singlecrystal 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 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations Title: Evolutionary concepts High burn-up and MOX fuel THE INFLUENCE OF MICROSTRUCTURE OF MOX FUEL ON ITS IRRADIATION BEHAVIOUR UNDER TRANSIENT CONDITIONS Acronym MICROMOX Contract number FIKS-CT2000-00030 Duration 48 months Proposal number FIS5-1999-00149 Type of action Starting date Total budget Shared cost 1 October 2000 2.346.011 € EC project officer A. Zurita EC contribution 800.027 € Co-ordinator Organisation Address Contact person Belgonucléaire S.A. Engineering Avenue Ariane, 4 B-1200 Brussels Mr. Marc Lippens Tel: (32-2) 7740625 Fax: (32-2) 7740547 Email m.lippens@belgonucleaire.be Partnership Country UK INT INT NL CH Organisations British Nuclear Fuels plc (BNFL) European Commission - JRC/IE European Commission - JRC/ITU Nuclear Research and Consultancy Group (NRG) 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 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations Title: Evolutionary concepts High burn-up and MOX fuel OXIDE FUELS: MICROSTRUCTURE AND COMPOSITION VARIATIONS Acronym OMICO Contract number FIKS-CT2001-00141 Duration 36 months Proposal number FIS5-2001-00037 Type of action Starting date Total budget Shared cost 1 October 2001 2.047.945 € EC project officer A. Zurita EC contribution 1.023.972 € Co-ordinator Organisation Address Contact person Belgian Nuclear Research Centre (SCK-CEN) Reactor Materials Research Boeretang 200 B-2400 Mol Dr. Marc Verwerft Tel: (32-14)333048 Fax: (32-14)321216 Email mverwerf@sckcen.be Partnership Country INT F Organisations European Commission - JRC/ITU 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 subassembly 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 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations Title: Evolutionary concepts High burn-up and MOX fuel VALIDATION OF HIGH BURNUP MOX FUELS CALCULATIONS Acronym VALMOX Contract number FIKS-CT2001-00191 Duration 30 months Proposal number FIS5-2001-00107 Type of action Starting date Total budget Shared cost 1 October 2001 1.000.000 € EC project officer A. Zurita EC contribution 500.000 € Co-ordinator Organisation Address Contact person Belgonucléaire S.A. Avenue Ariane, 4 B-1200 Brussels Mr. Servais Pilate Tel: (32-2)7740569 Fax: (32-2)7740547 Email s.pilate@belgonucleaire.be Partnership Country B F NL Organisations Belgian Nuclear Research Centre (SCK-CEN) Commissariat à l'Energie Atomique (CEA) 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 Contents 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 Contents Nuclear Energy Programme Operational safety of existing installations Title: Evolutionary concepts High burn-up and MOX fuel SIMULATION OF RADIATION EFFECTS IN ZR-NB ALLOYS: APPLICATION TO STRESS CORROSION CRACKING BEHAVIOUR IN IODINE-RICH ENVIRONMENT Acronym SIRENA Contract number FIKS-CT2001-00137 Duration 36 months Proposal number FIS5-2001-00032 Type of action Starting date Total budget Shared cost 1 January 2002 641.787 € EC project officer G. Van Goethem EC contribution 398.617 € Co-ordinator Organisation Address Contact person Electricité de France (EDF) Dept. Etude des Matériaux Site des Renardières F-77818 Moret sur Loing Ms. Stéphanie Jumel Tel: (33-1)60736174 Fax: (33-1)60737369 Email stephanie.jumel@edf.fr Partnership Country E UK B F S E F Organisations Universidad Politecnica de Madrid (UPM) The University of Liverpool Université Libre de Bruxelles (ULB) Commissariat à l'Energie Atomique (CEA) Westinghouse Atom AB Universidad Politecnica de Madrid (UPM) 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 Contents the range of 1 million euros). It is therefore interesting to complement it by using state-ofthe-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. ZrNb-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° FIR1CT2001-20136) 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 Contents Nuclear Energy Programme Operational safety of existing installations Title: Evolutionary concepts High burn-up and MOX fuel EXTENSION OF TRANSURANUS CODE APPLICABILITY WITH NB CONTAINING CLADDING MODELS Acronym EXTRA Contract number FIKS-CT2001-00173 Duration 24 months Proposal number FIS5-2001-00083 Type of action Starting date Total budget Shared cost 1 December 2001 438.999 € EC project officer A. Zurita EC contribution 219.499 € Co-ordinator Organisation Address Contact person Atomic Energy Research Institute (KFKI) Fuel & Reactor Materials Department Konkoly Thege str. 29-33 HU-1121 Budapest Dr. Csaba Gyori Tel: (36-1)3922294 Fax: (36-1)3959293 Email gyori@sunserv.kfki.hu Partnership Country SK INT Organisations Nuclear Power Plant Research Institute (VUJE) Trnava Inc 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 Contents 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 150