STATE AND INVESTIGATIONS OF NUCLEAR SAFETY IN EU

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1 STATE AND INVESTIGATIONS OF NUCLEAR SAFETY IN EU LONG--TERM RESEARCH IN THE FIELD OF NUCLEAR SAFETY FOR SUSTAINABLE DEVELOPMENT IN EU D.POPOV member of BgNS and ENS HSC Kozloduy NPP 1

2 WHAT IS SAFETY? SAFETY MARGINS SAFETY LIMITS UNCERTAINTY 2

3 SAFETY Safety of Fuel mining, fabrication, operation and reprocessing Operational Safety Radiation Safety Aging Safety (PLIM) Decommissioning Safety 3

4 EC Sustainable Nuclear Energy Technological Platform (SNE-TP) To remain competitive, European industry needs to specialize more in high-technology areas. Investment in research must be increased, coordination across Europe enhanced, and the technological content of industrial activity raised. Technology platforms address these challenges through: - a shared vision of stakeholders; - positive impact on a wide range of policies; - reduced fragmentation of research and development efforts; -mobilisation of public and private funding sources As the biggest provider of low-ghg-emitting energy in Europe, and one of the least expensive, nuclear fission has a key role to play in the future energy policy. The research in Europe is still fragmented and suffers from a lack of funding, at national and industrial levels and at the level of the Euratom Framework Programme. Action was therefore needed now to enable Europe to retain its leading technological and industrial position in the field of civil nuclear technology. 4

5 Sustainable Nuclear Energy Technological Platform (SNE-TP) Strategic research Agenda (SRA) It establish a strategic research agenda (SRA) for developing technologies, taking into account users requirements as well as safety considerations. The proposed Sustainable Nuclear Energy Technology Platform (SNE-TP) fulfills the following tasks: establish a strategic research agenda (SRA) and a deployment strategy (DS) to ensure that nuclear fission energy is generated in a manner that meets the criteria for sustainable development in strict compliance with the safety requirements; coordinate techno-economic studies to monitor the advances of nuclear technologies and EU needs, and the role of nuclear energy in the EU energy mix; foster joint initiatives between researchers, industry, utilities, Member States and the EU, such as joint undertakings; provide expert advice and recommendations to the Commission and national governments for strengthening the European scientific base, integrating research teams and tools, optimising the use of existing research infrastructures, and creating new infrastructures; foster joint projects between Member States; suggest topics for coordination or funding at European level, e.g. via the Euratom Framework Programme; provide timely information about advances in nuclear energy to the general public. 5 disseminate the results of the above activities to appropriate policy-making and stakeholder bodies to ensure a common European vision; Interact with other EU Technological platforms, such as Geological Disposal Technology Platform

6 Energy needs of EU Community EU25 Growth of import dependency U3O8 spot prices since 1990 (Source: T.Abram, Integration of niternational research in innovative GenIV system.., FISA 06) Source: J. Mišák, F. Pazdera, Future prospects for European cooperation in nuclear safety research,fisa 06 6

7 EURATOM ROADMAP KEEPING THE NUCLEAR OPTION OPEN; NEED FOR DRASTIC DECREASE OF THE CARBON INTENSITY IN THE ECONOMY; ADDRESSING THE CHALLENGES OF GLOBAL WARMING; EMPHASIS ON THE SAFETY OF INSTALLATIONS AND OF DISTRIBUTION NETWORKS AND SECURITY OF SUPPLY DEVELOPMENT OF FAST NEUTRON REACTORS DEVELOPMENT OF PASSIVE DESIGNS Source: J. Mišák, F. Pazdera, Future prospects for European cooperation in nuclear safety research,fisa 06 7

8 EURATOM FP-7 PROJECTS Generation-IV Innovative Concepts 6 REACTOR SYSTEMS SELECTED BY GIF FOR FURTHER DEVELOPMENT (~2030): Water-Cooled: 1. Super-Critical Water Reactor (SCWR) Gas-cooled: 2.Very-High Temperature Reactor (VHTR) 3.Gas-cooled Fast Reactor (GCFR) Liquid Metal-Cooled 4. Sodium cooled Fast Reactor (SFR) 5. Lead Fast Reactor (LFR) /Pb-Bi cooled Non-Classical 6. Molten Salt cooled Reactor (MSR) Features of Gen IV: 1. High Safety Level 2. Good Economy 2.1. High efficiency (η~ up to 50%) 2.2.Possibility to develop H2 industry 3. Proliferation resistance (Minor Actinides) 8

9 EURATOM FP-7 PROJECTS Generation-IV Innovative Concepts Gas Cooled Fast Reactor FEATURES Lead-cooled by natural convection Outlet temperature of the helium 850 C Possible to: deliver Electricity deliver Hydrogen From breeder to burner and recycling minor actinides. Closed fuel cycle for efficient conversion of fertile uranium and management of actinides Reference reactor MWth/288-MWe using a direct Brayton cycle gas turbine for high thermal efficiency. The GCFR reference has an integrated, on-site spent fuel treatment and refabrication plant. Through the combination of a fast spectrum and full recycle of actinides, the GCFR minimizes the production of longlived radioactive waste. 9

10 EURATOM FP-7 PROJECTS Generation-IV Innovative Concepts Gas Cooled Fast Reactor cont d Fuel type: CERCER plate fuel - hold the potential to operate at very high temperatures and to ensure an excellent retention of fission products: composite ceramic fuel, advanced fuel particles, or ceramic clad elements of actinide compounds. Core configurations may be based on prismatic blocks, pin- or plate-based assemblies. CERCER plate fuel High power density ~ 100 MW/m^3 PARTICIPANTS National Nuclear Corporation Ltd Nexia Solutions Commissariat à l Energie Atomique Empresarios Agrupados Internacional S.A. Framatome ANP SAS Joint Research Centre ITU and IE Nuclear Research and Consultancy Group Paul Scherrer Institut Delft University of Technology InterUniversities Consortium for Nuclear Technological Research University of Pisa NNC NEXIA CEA EA FANP SAS JRC NRG PSI TUD CIRTEN- UNIPI The GFR system is top-ranked in sustainability because of its closed fuel cycle and excellent performance in actinide management. It is rated good in safety, economics, and in proliferation resistance and physical protection 10

11 155,00 144, , EURATOM FP-7 PROJECTS Generation-IV Innovative Concepts Gas Cooled Fast Reactor cont d Development Strategy Commercial electricity production plant GFR 2400 MWth (modular 600 MWth) by ~2030 ETDR (Experimental Technology Demonstrator Reactor) MWth, first GCFR at Cadarache, France Separate project from 2008, decision to build 2012 Contrasts earlier GCFR projects (direct to prototype/demo) Synergies with HTR Helium-gas cooled High temperature materials and components Benefit from FP6 RAPHAEL-IP/Gen IV VHTR 9080, , ,25 750, , ,00 990, ,00 11

12 EURATOM FP-7 PROJECTS Generation-IV Innovative Concepts LCR - Lead Cooled (Fast) Reactor System FEATURES Lead-cooled by natural convection, fastneutron spectrum. Fuel = MOX Outlet temperature of the helium 550 C up to 850 C depending on the advanced materials to be used. Possible to: deliver Electricity deliver Hydrogen deliver potable Water Full actinide recycle fuel cycle with central or regional fuel cycle facilities Options include a range of plant ratings, including a battery of MWe that features a very long refueling interval (15-20 yr) with cassette core or replaceable reactor module, a modular system rated at MWe, and a large monolithic plant option at 1200 MWe. 12

13 EURATOM FP-7 PROJECTS Generation-IV Innovative Concepts LCR - Lead Cooled (Fast) Reactor System cont d Sustainability - Resource utilization. Because lead is a coolant with very low neutron absorption and moderation, it makes possible an efficient utilization of excess neutrons and reduction of specific uranium consumption. Reactor designs can readily achieve a breeding ratio of about 1, and long core life and a high fuel burnup can be achieved. -Waste minimization and management. A fast neutron flux significantly reduces waste generation, Pu recycling in a closed cycle being the condition recognized by GEN IV for waste minimization. The capability of the LFR systems to safely burn recycled minor actinides within the fuel will add to the attractiveness of the LFR. Economics. -Life cycle cost. The cost advantage features of the LFR must include low capital cost, short construction duration and low fuel and low production cost. The economic utilization of MOX fuel in a fast spectrum has been already demonstrated in the case of the SFR, and no significantly different conclusion can be expected for the LFR except from improvement due to the harder spectrum. 13

14 Because of the favorable characteristics of molten lead, it will be possible to significantly simplify the LFR systems in comparison with the well known designs of the SFRs, and hence to reduce its overnight capital cost, which is a major cost factor for the competitive generation of nuclear electricity. A simple plant will be the basis for reduced capital and operating cost. A pool-type, low-pressure primary system configuration offers great potential for plant simplification. 14

15 EURATOM FP-7 PROJECTS Generation-IV Innovative Concepts SCR - Sodium-Cooled (Fast) Reactor System SFR Sodium-Cooled Fast Reactor System The Sodium-Cooled Fast Reactor (SFR) system features fast-neutron spectrum and a closed fuel cycle for efficient conversion of fertile uranium and management of actinides. A full actinide recycle fuel cycle is envisioned with two major options: One is an intermediate size (150 to 500 MWe) sodium-cooled reactor with uranium-plutonium-minor-actinidezirconium metal alloy fuel, supported by a fuel cycle based on pyrometallurgical processing in collocated facilities. The second is a medium to large (500 to 1500 MWe) sodium-cooled fast reactor with mixed uranium-plutonium oxide fuel, supported by a fuel cycle based upon advanced aqueous processing at a central location serving a number of reactors. The outlet temperature is approximately 550 C for both. The primary focus of the R&D is on the recycle technology, economics of the overall system, assurance of passive safety, and accommodation of bounding events. The SFR system is top-ranked in sustainability because of its closed fuel cycle and excellent potential for actinide management, including resource extension. It is rated good in safety, economics, and proliferation resistance and physical protection. It is primarily envisioned for missions in electricity production and actinide management. The SFR system is the nearest term actinide management system. Based on the experience with oxide fuel, this option is estimated to be deployable by

16 EURATOM FP-7 PROJECTS Generation-IV Innovative Concepts SCWR Supercritical-Water-Cooled Reactor System The Supercritical-Water-Cooled Reactor (SCWR )system features two fuel cycle options: the first is an open cycle with a thermal neutron spectrum reactor; the second is a closed cycle with a fast-neutron spectrum reactor and full actinide recycle. Both options use a high-temperature, high-pressure, water-cooled reactor that operates above the thermodynamic critical point of water (22.1 MPa, 374 C) to achieve a thermal efficiency approaching 44%. The fuel cycle for the thermal option is a once-through uranium cycle. The fast-spectrum option uses central fuel cycle facilities based on advanced aqueous processing for actinide recycle. The fast-spectrum option depends upon the materials R&D success to support a fast-spectrum reactor. In either option, the reference plant has a 1700-MWe power level, an operating pressure of 25 MPa, and reactor outlet temperature of 550 C. Passive safety features similar to those of the simplified boiling water reactor are incorporated. Owing to the low density of supercritical water, additional moderator is added to thermalize the core in the thermal option. Note that the balance-of-plant is considerably simplified because the coolant does not change phase in the reactor. The SCWR system is highly ranked in economics because of the high thermal efficiency and plant simplification. If the fast-spectrum option can be developed, the SCWR system will also be highly ranked in sustainability. The SCWR is rated good in safety, and in proliferation resistance and physical protection. The SCWR system is primarily envisioned for missions in electricity production, with an option for actinide management. Given its R&D needs in materials compatibility, the SCWR system is estimated to be 16deployable by 2025.

17 EURATOM FP-7 PROJECTS Generation-IV Innovative Concepts VHTR Very-High-Temperature Reactor System The Very-High-Temperature Reactor (VHTR) system uses a thermal neutron spectrum and a once-through uranium cycle. The VHTR system is primarily aimed at relatively faster deployment of a system for high temperature process heat applications, such as coal gasification and thermo-chemical hydrogen production, with superior efficiency. The reference reactor concept has a 600-MWth helium cooled core based on either the prismatic block fuel of the Gas Turbine Modular Helium Reactor (GT-MHR) or the pebble fuel of the Pebble Bed Modular Reactor (PBMR). The primary circuit is connected to a steam reformer/steam generator to deliver process heat. The VHTR system has coolant outlet temperatures above 1000 C. It is intended to be a high-efficiency system that can supply process heat to a broad spectrum of high temperature and energy-intensive, nonelectric processes. The system may incorporate electricity generation equipment to meet cogeneration needs. The system also has the flexibility to adopt U/Pu fuel cycles and offer enhanced waste minimization. The VHTR requires significant advances in fuel performance and high temperature materials, but could benefit from many of the developments proposed for earlier prismatic or pebble bed gas-cooled reactors. Additional technology R&D for the VHTR includes high-temperature alloys, fiber-reinforced ceramics or composite materials, and zirconium-carbide fuel coatings. The VHTR system is highly ranked in economics because of its high hydrogen production efficiency, and in safety and reliability because of the inherent safety features of the fuel and reactor. It is rated good in proliferation resistance and physical protection, and neutral in sustainability because of its open fuel cycle. It is primarily envisioned for missions in hydrogen production and other process-heat applications, although it could produce electricity as well. The VHTR system is the nearest-term hydrogen production system, estimated to be deployable by

18 EURATOM FP-7 PROJECTS Generation-IV Innovative Concepts MSR- Molten Salt Reactor Sytem The Molten Salt Reactor (MSR) system produces fission power in a circulating molten salt fuel mixture with an epithermal-spectrum reactor and a full actinide recycle fuel cycle. In the MSR system, the fuel is a circulating liquid mixture of sodium, zirconium, and uranium fluorides. The molten salt fuel flows through graphite core channels, producing an epithermal spectrum. The heat generated in the molten salt is transferred to a secondary coolant system through an intermediate heat exchanger, and then through a tertiary heat exchanger to the power conversion system. The reference plant has a power level of 1,000 MWe. The system has a coolant outlet temperature of 700 degrees Celsius, possibly ranging up to 800 degrees Celsius, affording improved thermal efficiency. The closed fuel cycle can be tailored to the efficient burn up of plutonium and minor actinides. The MSR's liquid fuel allows addition of actinides such as plutonium and avoids the need for fuel fabrication. Actinides - and most fission products - form fluorinides in the liquid coolant. Molten fluoride salts have excellent heat transfer characteristics and a very low vapor pressure, which reduce stresses on the vessel and piping. 18 The MSR system is top-ranked in sustainability because of its closed fuel cycle and excellent performance in waste burndown. It is rated good in safety, and in proliferation resistance and physical protection, and it is rated neutral in economics because of its large number of subsystems. It is primarily envisioned for missions in electricity production and waste burndown. Given its R&D needs for system development, the MSR is estimated to be deployable by 2025.

19 SAFETY OF EXISTING INSTALLATIONS - AGING (FP7-NULIFE) 19

20 SAFETY OF EXISTING INSTALLATIONS - AGING VERLIFE Unified Procedure for lifetime assessment for VVER Main content: 1. Introduction 2. General statements, definitions and abbreviations 3. General requirements for calculations of residual lifetime 4. Procedure for assessment of residual lifetime of the component 5. Assessment of component resistance against fast fracture 6. Residual lifetime of the component from the point of view of resistance against fatigue damage 7. Residual lifetime of the component from the point of view of resistance against corrosion-mechanical damage 8. Assessment of acceptability of flaws found during in-service inspection 9. Complex (total) assessment of residual lifetime 20

21 SAFETY OF EXISTING INSTALLATIONS - AGING VERLIFE Unified Procedure for lifetime assessment for VVER Verlife-2008 IAEA Project NoRER 4030/9004 Kick-off meeting March 2009, Řež, Czech Republic New appendices: Appendix A: LBB assessment procedure of WWER piping during operation Appendix B: No-break-zone assessment procedure of WWER piping during operation Appendix C: Integrity and lifetime assessment procedure of RPV internals in WWER NPPs during operation Appendix D: Procedure for risk informed approach for ISI of piping Appendix E: Recommended procedure for NDE qualification Appendix F: Component and Piping Supports 21

22 EURATOM FP-7 PROJECTS SAFETY OF EXISTING INSTALLATIONS NURECIM -NURISP European Platform for Nuclear Reactor Simulation NURISP SCOPE AND OBJECTIVES 18 Organizations 13 Countries 22

23 EURATOM FP-7 PROJECTS SAFETY OF EXISTING INSTALLATIONS NURISP European Platform for Nuclear Reactor Simulation cont d NURISP PLATFORM 23

24 EURATOM FP-7 PROJECTS SAFETY OF EXISTING INSTALLATIONS NURISP European Platform for Nuclear Reactor Simulation cont d SP1: CORE PHYSICS Structure & Tasks WP1.1 Advanced Monte Carlo Methods TRIPOLI4 (CEA) + Adjoin methods (TUD+KTH) WP1.2 Advanced Deterministic Diffusion and Transport Methods APOLLO2 + CRONOS => DESCARTES (CEA) ANDES nodal solver + COBAYA3 cell-nodal (UPM) WP1.3 Advanced Neutron Kinetics Methods DESCARTES + COBAYA3 + DYN3D (FZR) WP1.4 Coupled Calculations and Transient Benchmarks (MSLB, CRE for PWR,VVER) 24

25 EURATOM FP-7 PROJECTS SAFETY OF EXISTING INSTALLATIONS NURISP European Platform for Nuclear Reactor Simulation cont d Structure & Tasks cont d SP2: THERMALHYDRAULICS (with CFD) WP 2.1: Pressurized Thermal Shock & Direct Contact Condensation WP 2.2: Critical Heat Flux SCIENTIFIC CHALLENGES ASSOCIATED TO PTS and CHF Condensation on the jet Bubble Entrainment by jet Turbulence production below jet Turbulence production at free surface Turbulence effects on condensation Friction at free surface Review of existing data Identification of experimental needs Implementation of available modules Development of new physical models Benchmarking and assessment Interactions between waves, turbulence & condensation Effects of T stratification Flow Separation Wall to fluid in CL & downcomer 25

26 EURATOM FP-7 PROJECTS SAFETY OF EXISTING INSTALLATIONS NURISP European Platform for Nuclear Reactor Simulation cont d Structure & Tasks cont d SP3: MULTIPHYSICS Review and specification, within the NURISP platform, of coupling schemes for core analysis based on existing CP and TH core codes At the nodal level (fuel assembly) At the sub-node level (pin) Development and integration within the NURESIM platform of core parameters interpolation and averaging schemes and data transfer Application to PWR (MSLB) and BWR A multi-scale analysis of accidental transients with 3D simulation 26

27 EURATOM FP-7 PROJECTS SAFETY OF EXISTING INSTALLATIONS NURISP European Platform for Nuclear Reactor Simulation cont d Structure & Tasks cont d SP4: SENSITIVITY AND UNCERTAINTY Sensitivity and uncertainty analysis of multiphysics modules - State of the Art report on deterministic and statistical methods - Evaluation of critical points of model responses Implementation within the NURISP platform of procedures for propagation of uncertainties SP5: INTEGRATION Specific training courses on the SALOME platform Assistance in integrating codes Adaptation of the SALOME platform Ensuring consistency and non-regression 27

28 EURATOM FP-7 PROJECTS SAFETY OF EXISTING INSTALLATIONS NURISP European Platform for Nuclear Reactor Simulation cont d NURISP PROVIDES THE BASIS FOR A LONG TERM STRATEGY TOWARDS A EUROPEAN SOFTWARE PLATFORM FOR NUCLEAR SAFETY ANALYSES STRONG POSSIBLE BREAKTHROUGHS IN PHYSICAL MODELLING, NUMERICAL METHODS AND COMPUTER SCIENCE AN ANSWER TO MEET THE NEEDS OF THE EUROPEAN NUCLEAR INDUSTRY, TO MAINTAIN ITS EFFICIENCY AND COMPETITIVENESS 28

29 EUROPEAN NETWORK FOR DECOMMISSIONING Working Areas Strategies, Policies and Funding Site Characterisation, Remediation and Reuse Recycle and Reuse of Materials Public Perception, Public Relation Institutional, Legal, Regulatory Aspects, Licensing and Decommissioning Plan Radiological Protection and Industrial Safety Environmental and Socio-economic Aspects Project Management and Planning Material Management and Characterisation Techniques Cost Aspects Dismantling Techniques Decontamination Techniques Dissemination of Best Practice, Experience, Know-how Co-operation/Relations with International Organisations Compendium on Decommissioning and Decontamination 29 Participants Agency for Radwaste Management (Slovenia) AllDeco s.r.o. (Slovakia) AREVA/COGEMA (France) AWE plc. (United Kingdom) Babcock Noell Nuclear GmbH (Germany) Belgoprocess N.V. (Belgium) Brenk Systemplanung GmbH (Germany) British Nuclear Group (United Kingdom) Centrale Organisatie voor Radioactief Afval (COVRA) N.V. (Netherlands) CEZ a.s. (Czech Republic) Colenco Power Engineering Ltd (Switzerland) Commissariat а l'energie Atomique (CEA) (France) Consejo de Seguridad Nuclear (Spain) CORE2 Consult GmbH (Germany) DDR Consult (Belgium) DECOM Slovakia Ltd (Slovakia) EC-CND.NET (Germany) Electricitй de France (France) Empresa Nacional de Residuos Radiactivos S.A. (ENRESA) (Spain) Energiewerke Nord GmbH (EWN) (Germany) ENPRO Consult Ltd. (Bulgaria) Equipos Nucleares S.A. (ENSA) (Spain) European Commission (Belgium) Federal Office for Radiation Protection (BfS) (Germany) Forschungszentrum Karlsruhe GmbH (Germany) HMS SULTAN (United Kingdom) International Atomic Energy Agency (IAEA) (Austria) Ministry of Economy, State Fund for Decommissioning (Slovakia) NIRAS/ONDRAF (Belgium) NIS Ingenieurgesellschaft mbh (Germany) Nuclear Engineering Seibersdorf GmbH (Austria) Nuclear Research and consultancy Group (NRG) (Netherlands) Nuclear Research Institute Rez plc (Czech Republic) NUKEM Limited (United Kingdom) Nuklearna elektrarna Krљko (Slovenia) Oxford Biosphere Ltd (United Kingdom) Public Agency for Radioactive Waste Management (PURAM) (Hungary) Rolls-Royce Nuclear Engineering Services Ltd. (United Kingdom) Siempelkamp Nukleartechnik GmbH (Germany) SOGIN SpA (Italy) Stй Delattre Entrepose (France) STEAG encotec GmbH (Germany) Swedish Nuclear Fuel and Waste Management Company (SKB) (Sweden) Tecnubel S.A. (Belgium) The North Highland College (United Kingdom) TЬV Nord e.v. (Germany) TЬV NORD SysTec GmbH & Co. KG (Germany) United Kingdom Atomic Energy Authority (UKAEA) (United Kingdom)

30 EURATOM FP-7 PROJECTS SAFETY DURING SEVERE ACCIDENT SARNET Severe Accident Research Network DISCO-H DISCO-H facility with internal structures 30

31 EROBAROMETER HOW EU CITIZEN ACCEPT THE NUCLEAR ENERGY 31

32 EROBAROMETER HOW EU CITIZEN ACCEPT THE NUCLEAR ENERGY 32

33 ENS HSC ACTIVITY AND GOALS According to New Action 1/7 from the Protocol of ENS Board Meeting held in Brussels on November 30th 2007, Each Board Member should nominate a representative to construct and populate the ENS High Scientific Council The main conception has been that as a federation of learned societies, the influence of ENS worldwide could be increased if it could have a generic scientific production. The High Scientific Council is an instrument for this production 33

34 During the first meeting of HSC on April 16th 2008 in Brussels it has been underlined that the role and main tasks of HSC will focus on the following: 1. To critically review the current situation of the science and technology in the fields of interest for ENS. 2. To advise the ENS Programme Committee on ENSsponsored Conferences 3. To initiate discussion on certain issues of interest. 4. To produce a highlight paper after each topical meeting organized by ENS 5. To produce a position papers on critical topics in Nuclear 6. To produce expert analyses 34

35 In the beginning of 2008 HSC has been constructed as follows: Prof. Phil Beeley Director, Nuclear Department Defence Academy United Kingdom Nuclear Engineering, Education and Training Prof. Dr. Frans Corstens Medical Faculty, University of Nijmegen NETHERLANDS Nuclear Medicine Dr. Ingeborg Hagenlocher Former president of the Swiss Nuclear Society Heidelberg, GERMANY Radioactive waste disposal, Knowledge management in nuclear technology Dr Bernard Bonin Direction Scientifique Direction de l'énergie Nucléaire CEA, France Nuclear Fuel Pierre Goldschmidt Carnegie Endowment for International Peace Brussels,BELGIUM Non-proliferation, Safeguards Juraj Klepac Division of Nuclear Safety VUJE Trnava Inc SLOVAKIA Assessment operation safety of NPP 35

36 MEMBERS OF ENS HSC (cont d) Prof. Vladimir Slugen SNUS President FEI STU KJFT Bratislava, SLOVAKIA José Emeterio Gutiérrez Technical Service Director Westinghouse SPAIN Nuclear fuel Prof. Ernest Mund BELGIUM Nuclear reactor calculations Jerôme Pamela EFDA (European Fusion Development Agreement) Garching, Germany Prof. Riitta Kyrki-Rajamäki Lappeenranta University of Technology Finland Nuclear safety&technology Prof. Dr. Oldrich Matal ENERGOVYZKUM Brno, CZECH REPUBLIC Innovative Reactors, Advanced Technology 36

37 MEMBERS OF ENS HSC (cont d) Kosta Stoychev Dimitar Popov Prof. DScTroyo Troev Training Centre, Engineering Support Divis INRNE,BAS Kozloduy NPP, Plc. Kozloduy NPP, Plc. BULGARIA BULGARIA BULGARIA Dr. Bernard Bonin (CEA, France) was elected as Chairman of HSC 37

38 SAFETY OF RESEARCH REACTORS First highlights paper of new HSC devoted to RRFM 2008 Conference (Key points) 1. The biggest issue at the conference was the progress made in the conversion of the cores of research reactors from highly-enriched uranium (HEU) to lowenriched uranium (LEU). The Program of core conversion (initiated in 1978 under the auspices of the US Department of Energy) supports the minimization and the elimination of the use of HEU in civilian nuclear applications. 2. As of 2008, a total of 207 research reactors were involved in the project worldwide. 56 have already been converted, 78 are beyond scope, and 46 are planned for conversion with existing LEU fuel. The remaining 28 are high performance reactors, also planned for conversion but these will need fuel of a new type to comply with core conversion without losing too much in performance. 3. The permanent challenge of research reactors devoted to testing or irradiation is to produce high neutron fluxes with limited amounts of fissile material. Conversion of research reactor cores to LEU has made the need for dense fuel all the more urgent. 38

39 SAFETY OF RESEARCH REACTORS (Key points cont d) 4. The intermetallic compound U3Si2 is presently the reference fuel, with a well mastered production process on the industrial scale and a good behaviour profile under irradiation. But its density is only 4.8 gu.cm-3, and this is clearly not sufficient for the conversion of some of the more demanding research reactors. 5. Higher densities can be reached by switching to UMo alloy, where the 7-10% Mo additive has been chosen for its capacity to stabilize the gamma phase of uranium. Monolithic UMo alloy has a density as high as 16 gu.cm-3; UMo can also be made of powder, sandwiched between two co-laminated plates of Al. 6. The behaviour of this type of fuel plates has been tested under irradiation in various laboratories, with as yet not entirely satisfactory results. The general finding is that the Al matrix interacts with the UMo alloy to form an interaction layer where the gamma phase of the uranium crystal lattice is locally destroyed, with negative consequences on the behaviour of the fuel under irradiation - the swelling and pillowing of the fuel plate can modify the cooling of the fuel and cause its buckling; the fission gas release can cause blistering of the plate and its ultimate rupture. 39

40 SAFETY OF RESEARCH REACTORS (Key points cont d) 7. The addition of 2-5% of Si either in the Al matrix or in the UMo itself seems to limit both the development of this undesirable, mainly amorphous interaction region, and the resulting swelling. Reports from all laboratories confirm the positive role of Si on the fuel behaviour under irradiation. The phenomenology of the role of silicon is being better mastered, as silicated phases located at the interface between UMo and Al play the role of a diffusion barrier, which limits the development of the amorphous interaction layer. 8. Cumulated fission rates as high as 5.E21 fissions.cm-3 in the fuel grains, corresponding to burn-ups of 10 %, have been achieved with UMo fuels in powder form. Alternatives to the Aluminium cladding have been researched (stainless steel, zirconium alloy), with promising results so far. It is hoped that the promising additon of Si will ultimately result in a well-mastered fabrication process, with satisfactory fuel performance under irradiation. But progress is slow. 9. The US National Nuclear Security Administration recently issued a request for information, or RFI, on the nuclear industry's capability to fabricate very-highdensity low-enriched UMo fuel for research and test reactors. According to RFI s very ambitious schedule, the qualification of monolithic fuel for use in US reactors by the US Nuclear Regulatory Commission is anticipated for

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