ACTIVITIES in NUCLEAR FUEL BEHAVIOUR

Transcription

1 ACTIVITIES in NUCLEAR FUEL BEHAVIOUR Nuclear Science Committee Status: October 2002 Presented by Wolfgang Wiesenack R&D Needs for Current and Future Nuclear Systems, Nov

2 Outline Introduction - Historical Background Important Issues for Fuel Performance Modelling Identification of Available Experimental Data IFPE Database International Workshops Co-ordination with CSNI on Safety Related activities Conclusions / Needs / Actions R&D Needs for Current and Future Nuclear Systems, Nov

3 Introduction - historical background (1/2) Need for improved knowledge and international co-ordination of important scientific issues related to fuel behaviour NSC Task Force: end 1993 identify areas of high priority on basic underlying phenomena of fuel behaviour normal / off-normal operating conditions benefit for international co-ordination R&D Needs for Current and Future Nuclear Systems, Nov

4 Introduction - historical background (2/2) Advice on developments needed regarding data, models and experiments Better understanding of fuel behaviour Improved predictive models Reports: Scientific Issues in Fuel Behaviour(1995) Review of Nuclear Fuel Experimental Data - Fuel Behaviour Data Available from IFE/OECD Halden Project for Development and Validation of Computer Codes (1995) R&D Needs for Current and Future Nuclear Systems, Nov

5 Important issues for fuel performance modelling (1/3) Thermal Performance - the calculation of fuel temperatures, the effect of design parameters and the effect of irradiation Fission Gas Release - the release under different operating regimes and the effect of high burn-up Fission Product Swelling - the distinction between inexorable solid fission product swelling which is predominantly a function of burn-up and fission product accumulation, and gaseous swelling which is also dependent on high power operation R&D Needs for Current and Future Nuclear Systems, Nov

6 Important issues for fuel performance modelling (2/3) Stress Corrosion Cracking - the conditions of stress, time and fission product release on the propensity for clad failure from this mechanism Water Chemistry - current practice and monitoring, its effect along with Zr composition and final treatment on clad corrosion Hydrogen in Cladding - its distribution and measurement Failed Fuel - detection and modelling degradation processes R&D Needs for Current and Future Nuclear Systems, Nov

7 Important issues for fuel performance modelling (3/3) Spent Fuel - long term storage under wet and dry conditions Quality Assurance - as applied to materials, experimental tests and data production The Task force reviewed the state of understanding at that time and identified priority areas where further work was required R&D Needs for Current and Future Nuclear Systems, Nov

8 Conclusions made in the first report Although fuel in nuclear reactors had proven to be highly reliable in its performance and safety, reactor operation was often supported by rather generous bounding operating conditions. In the future the requirements would be more onerous. The economic requirement to increase efficiency meant that fuel performance must be calculated accurately on a best estimate basis. To do this requires good fuel performance computer codes validated by high quality data R&D Needs for Current and Future Nuclear Systems, Nov

9 Recommendations made in the first report (1/2) special research efforts are needed to reduce uncertainties in modelling specific aspects of fuel behaviour: thermal performance and calculating fuel temperatures fission gas release fission product swelling and creep of UO 2 thermo-mechanical behaviour high burn-up fuel behaviour in transient conditions R&D Needs for Current and Future Nuclear Systems, Nov

10 Recommendations made in the first report (2/2) A review should be made of existing data A public domain database should be assembled of well-qualified experiments and data which can be used for model development and code validation. This database should be organised and maintained by the NEA Data Bank International topical meetings covering high priority issues should be organised by the NEA in co-ordination with the IAEA R&D Needs for Current and Future Nuclear Systems, Nov

11 Identification of Available Experimental Data Review of existing data that could be made available to fulfil the aim of improving code performance. The report concentrated primarily on data produced in the Joint Programme carried out by the Halden Reactor (HR) Project Conditions in the HR are particularly well suited to studies of fuel performance. The boiling conditions ensure a constant coolant temperature, and hence a well-defined boundary condition from which to assess thermal performance from measurements of centreline fuel temperatures. Where prototypic conditions are required, dedicated in-pile loops are available to simulate the thermal hydraulic conditions of temperature and pressure as well as neutron flux spectrum for PWR, BWR and most recently, Advanced Gas Cooled Reactors R&D Needs for Current and Future Nuclear Systems, Nov

12 categories depending on the use Data useful for model development and validation Data of direct relevance to licensing requirements Data for fuel development and optimisation Phenomenological aspects: Radial Flux Depression Thermal Performance Fuel Densification and Swelling UO 2 Grain Growth Fission Product Release Clad Properties PCMI - Pellet Clad Mechanical Interaction Integral Behaviour High Burn-up Effects. R&D Needs for Current and Future Nuclear Systems, Nov

13 Review of Experimental Data - initial data sets - Halden experiments addressing single effects integral behaviour certain Studsvik ramp tests three Risø fission gas release tests R&D Needs for Current and Future Nuclear Systems, Nov

14 Review of Experimental Data - further requirements - Data to cover the extremes of burn-up and power expected in commercial reactors (>70 MWd/kgUO2) High burnup structure ('rim effect ) Fuel thermal conductivity Clad corrosion and hydriding Clad mechanical properties R&D Needs for Current and Future Nuclear Systems, Nov

15 Recommendation Assemble an International Fuel Performance Experiment (IFPE) database, including of experimental data identified in the report Such a database was then set up R&D Needs for Current and Future Nuclear Systems, Nov

16 IFPE Database - Aims Make available in one place a set of data and documentation generated in a variety of experiments/examinations Create a data set which is homogeneous in terms of validation requirements, review criteria and completeness of information Make the Database accessible in a straight forward manner Encourage the use of the Database for code validation purposes Provide a safe haven for data at risk R&D Needs for Current and Future Nuclear Systems, Nov

17 IFPE Database - Content Index file Summary file, data set description/scope Pre-characterisation data Irradiation history, in-pile data Post-irradiation examination data Quality assurance report Full archive of original reports After compilation and reviews, the data are stored in an archive and distributed on CD-ROM R&D Needs for Current and Future Nuclear Systems, Nov

18 IFPE Database Types of data Instrumented tests providing on-line data on fuel behaviour Post irradiation examination data Steady state, long term operation Power ramps (single, multiple ramps) Test reactor data and data after irradiations in commercial reactors Failed fuel data Cover BWR, CAGR, PHWR, PWR and WWER irradiation conditions R&D Needs for Current and Future Nuclear Systems, Nov

19 IFPE Database Data Currently Available (1/2) Halden irradiated IFA rods Halden irradiated IFA rods Halden irradiated IFA rods Halden irradiated IFA rod Halden irradiated IFA &.6 4 rods The Third Risø Fission Gas Release Project 16 rods The Risø Transient Fission Gas Release Project 15 rods The SOFIT WWER fuel Irradiation Programme 12 rods The High Burn-up Effects Programme 81 rods WWER rods from Kola-3 32 rods Studsvik INTER-RAMP BWR Project 20 rods Studsvik OVER-RAMP PWR Project 39 rods Studsvik SUPER-RAMP PWR Sub-Programme 28 rods Studsvik SUPER-RAMP BWR Sub-Programme 16 rods Studsvik DEMO- RAMP I BWR 5 rods Studsvik DEMO- RAMP II BWR 8 rods R&D Needs for Current and Future Nuclear Systems, Nov

20 IFPE Database Data Currently Available (2/2) Rods from the TRIBULATION programme 19 rods CEA/EDF/FRAMATOME Contact 1 & 2 3 rods AEAT-IMC NFB 8 and samples CEA/EDF/FRAMATOME PWR and OSIRIS ramped fuel rods 4 rods CENG defect fuel experiments 8 rods AECL-CANDU elements irradiated in NRU 36 rods Siemens PWR rods irradiated in GINNA 17 rodlets CNEA six power ramp irradiations with (PHWR) MOX fuels 5 rods CEA failed PWR rods irradiated in SILOE: EDITH-MOX 01 1 rod INR Pitesti - RO-89 and RO-51 CANDU fuel type in-reactor measurement of internal gas pressure 2 rods Belgonucleaire GAIN programme - fabrication, irradiation and PIE of 4 ramped UO 2 -Gd 2 O 3 fuel rods with Zr4 cladding, irradiated in BR3 PWR (50 GWd/tU for peak pellet) 4 rods R&D Needs for Current and Future Nuclear Systems, Nov

21 Recent Additions and Updates to IFPE NEW ADDITIONS Belgonucleaire GAIN programme - fabrication, irradiation and PIE of 4 ramped UO 2 -Gd 2 O 3 fuel rods with Zry-4 cladding, irradiated in BR3 PWR (50 GWd/tU for peak pellet): 4 rods INR Pitesti - RO-89 and RO-51 CANDU fuel type inreactor measurement of internal gas pressure: 2 rods REVISIONS CEA/EDF/FRAMATOME PWR and OSIRIS ramped fuel rods: 4 rods Rods from the TRIBULATION programme : 19 rods R&D Needs for Current and Future Nuclear Systems, Nov

22 Data Released or being processed Studsvik/SKI data from TRANS-RAMP I, II and IV Zaporoshye VVER1000 fuel behaviour data (4-8 cycles, Bu 50 MWd/kgUO 2 ) BR-3 High Burnup Fuel Rod Hot Cell Program (DOE/ET , Vol. 1 & 2) RISØ-I experiment IFE/OECD/HRP FUMEX 1-6 R&D Needs for Current and Future Nuclear Systems, Nov

23 IFPE data release in progress or requested HRP IFA rods 7 and 6 (cladding elongation, fuel temperature, FGR at Bu 60 MWd/kgUO 2 (for FUMEX-II) HRP IFA rods 18 and 19 ( EOL FGR and pressure, grains size of 22 and 8.5 micrometers and Bu 52 MWd/kgUO 2 (for FUMEX-II) HRP IFA-507 TF3 and TF5 (transient temperature during power increase) IFA-508 and IFA-515 conducted by JAERI at HRP - PCMI behavior data on different cladding thickness by means of diameter rig IMC (UK) swelling data from ramping CAGR UO 2 fuel in the Halden Reactor HRP He/Ar/Xe gas flow, Nb doped fuel, IFA-504 R&D Needs for Current and Future Nuclear Systems, Nov

24 IFPE data requested KOLA-3 + MIR test (temperature during ramp, FGR and pressure at EOL, Bu 55 MWd/kgUO 2 (FUMEX-II) NUPEC owned irradiation experiments VNIINM ramp data from VVER-440 up to 4 cycles and VVER-1000 up to 3 cycles CEA sweep gas experiment - HATAC CEA failed PWR rods irradiated in SILOE (EDITH-3, EDITH- MOX 02) Data from laser flash measurements Data from high temperature transients RIA data (FUMEX-II) LOCA data (FUMEX-II) R&D Needs for Current and Future Nuclear Systems, Nov

25 Contributing Organisations AEA Technology Atomic Energy of Canada Ltd. (AECL) Battelle Pacific North-West Laboratory (PNL) Belgonucleaire British Nuclear Fuels Ltd. (BNFL) Comision Nacional d'energia Atomica (CNEA) Commissariat a l'energie Atomique (CEA) EC Institute for Transuranium (ITU) Electricite de France (EdF) Framatome Halden Reactor Project (HRP) Imatran Voima Oy (IVO) Institute of Nuclear Research Pitesti (INR-Pitesti) Institute for Nuclear Research and Nuclear Energy, Sofia (INRNE Sofia) Kurchatov Institute Moscow (INR RCC) Research Institute of Inorganic Materials (VNIINM) Risoe National Laboratory Siemens Power Corporation Studsvik AB Swedish Nuclear Power Inspectorate (SKI) TVEL - Joint Stock Company, Moscow UK Health and Safety Executive UK Industry Management Committee (IMC) US Department of Energy(DOE) R&D Needs for Current and Future Nuclear Systems, Nov

26 IFPE Users / Projects 5 editions (since 1996) requested by 59 different establishments in 24 countries + 1 International organisation Argentina, Belgium, Bulgaria, Canada, Czech R., Denmark, Finland, France, Germany, Hungary, Italy, Japan, Korea, Norway, Poland, Romania, Russia, Slovak R., Spain, Sweden, Turkey, Ukraine, UK, USA, IAEA. feedback (PECO-1 project, ITU, INRNE, BN) FRAPCON upgrade IAEA 4/012 programme, FUMEX-II (IAEA) R&D Needs for Current and Future Nuclear Systems, Nov

27 IFPE Database Summary of activities and future perspectives operation and maintenance of IFPE (currently 422 cases) facilitate end users, encourage use and feedback handling of feedback from its utilisation improve database through re-analysis identification of needs for further data to be integrated acquisition of new experimental data be selective on data in term of quality, reliability, applicability link to concrete issues, e.g., fuel performance at high burn-up establishing an IFPE user forum cooperation with IAEA on FUMEX-II (FUel Modelling EXercise) R&D Needs for Current and Future Nuclear Systems, Nov

28 3 Seminars on Basic Phenomena (Journées de Cadarache) CEA, Cogema, EDF, Framatome, IAEA, OECD/NEA Thermal Performance in High Burnup LWR Fuels 3-6 March participants from 19 countries and 4 International Organisations Industry, Research labs, Utilities, Vendors Proceedings published by OECD/NEA Fission Gas Behaviour in Water Reactor Fuels September participants / 21 countries and 4 International Organisations ~ 40 papers have been accepted held in conjunction with meeting of Expert Group on scientific issues in fuel behaviour and ANS Group 5.4 Proceedings published by OECD/NEA Pellet Clad Mechanical Interactions (20-22(?) October 2003) R&D Needs for Current and Future Nuclear Systems, Nov

29 Expert Group on Reactor-based Pu disposition Separate project - WG-Pu (RF & USA) Disposition of t of surplus WG-Pu Issues addressed: Reactor Physics Fuel Cycle Fuel Behaviour --> Experience from the use of civil MOX fuels Contribution to speeding up disposition process R&D Needs for Current and Future Nuclear Systems, Nov

30 Fuel Performance Exercises and Benchmarks PWR & VVER Two sets of experimental data concerned with MOX fuel behaviour will be available to the project: Solid and Hollow pellet performance from HRP (blind benchmark) The solid and hollow pellets blind benchmark comparison against experiment has been adopted as the first exercise. Both US NRC and K.I. have released the data for use within the TFRPD benchmark exercise PRIMO experiment from Belgonucléaire / SCK-CEN R&D Needs for Current and Future Nuclear Systems, Nov

31 Co-ordination with CSNI on Safety Related Activities (1/3) Special Expert Group on Fuel Safety Margins 3rd meeting, Cadarache May 2002 Fuel related safety parameters Phenomena & associated safety limits and criteria Review - new/modern fuel characteristics MOX, new claddings Transients & Accident Conditions High-burnup Technical rather than regulatory safety rather than general fuel performance R&D Needs for Current and Future Nuclear Systems, Nov

32 Co-ordination with CSNI on Safety Related Activities (2/3) Oxidation of zirconium alloys at high pressure (small or intermediate breaks, transients) "Best estimate" calculation of LOCA scenario with high burnup fuel Updating information on nuclear safety research facilities at risk (follow-up to the SESAR/FAP report) The Phebus reactor future in the frame of LOCA programs IAEA working group assessment of differences and common features between PWR and WWER fuel safety criteria R&D Needs for Current and Future Nuclear Systems, Nov

33 Co-ordination with CSNI on Safety Related Activities (3/3) Follow - up to the LOCA Topical Meeting: Proposal to develop "standard" experimental methods to derive LOCA Safety Limits Proposal on development of a diffusion model for high temperature oxidation of Zr-Nb-O tubing Follow-up to the RIA Topical Meeting (establishing a TF)/Presentation to CNRA/CSNI Proposal of a benchmark exercise of computer codes modelling high burnup fuel behaviour in accident conditions for licensing purposes R&D Needs for Current and Future Nuclear Systems, Nov

34 RIA TOPICAL MEETING "Best estimate" core calculations for RIA energy deposition in high burnup fuel Current and new RIA safety criteria, the technical background Ongoing RIA Experimental Programmes CABRI SEMINAR R&D Needs for Current and Future Nuclear Systems, Nov

35 The CABRI Project (1/2) IRSN proposed project under OECD/CSNI umbrella (3 years) behaviour of high-burnup fuel under prototypical conditions of fast reactivity transients develop suitable safety criteria and limits installation of pressurised water loop in CABRI modification of reactor control and instrumentation Tests: Ultra-high burnup parametric tests (ramp rate, fuel grain size, &c.) MOX fuels possibly LOCA tests R&D Needs for Current and Future Nuclear Systems, Nov

36 The CABRI Project (2/2) Pressurised water loop installation in 2002 tests during 2-3 years complemented by analytical development programme and separate effects tests -> special phenomena & material properties for codes a number of OECD countries expressed interest legal agreement is being finalised technical committee is being formed R&D Needs for Current and Future Nuclear Systems, Nov

37 Conclusions / Needs / Actions Tasks assigned to the NSC Expert Group on Scientific Issues of Fuel Behaviour are not yet complete. Need for extending the Database to include a more comprehensive set of experiments for model development and improvement, especially in view of higher enrichments and burnup. The inclusion of experimental data related to gadolinia doped UO 2, Er-fuel, Zr(1%Nb) clad performance data and MOX fuels should also be expanded, plus cases of higher discharge burn-up Need for restructuring to facilitate and expand its use. Use in the FUMEX-II exercise will be beneficial Use in further model developments / improvements / verification Organisation of PCMI seminar in October 2003 Next meetings 16 December 2002 / 20(?) October 2003 R&D Needs for Current and Future Nuclear Systems, Nov

38 Web Page Updated information and reports on the IFPE are accessible in (11 July 2002) IFPE Internet user forum R&D Needs for Current and Future Nuclear Systems, Nov