Properties and behaviour of irradiated fuel under accident conditions

Transcription

1 IAEA, International Experts Meeting, Vienna, March 2012 Properties and behaviour of irradiated fuel under accident conditions V.V. Rondinella, R.J.M. Konings, J.-P. Glatz, P.D.W. Bottomley, T.A.G. Wiss, D. Papaioannou, O. Benes, J.-Y. Colle, C.T. Walker, S. Bremier, D. Serrano-Purroy, D. Staicu, D. Manara, L. Vlahovic, P.Pöml, Th. Fanghänel European Commission, Joint Research Centre, Institute for Transuranium Elements P.O. Box 2340, Karlsruhe, Germany 1

2 Outline IAEA, International Experts Meeting, Vienna, March Spent Fuel Safety in the Light of the Accident at the Fukushima Daiichi Nuclear Power Plant context: safety of nuclear fuels and cycles at JRC-ITU previous studies on fuel under extreme/accident conditions refocusing activities - source term: high T properties and behaviour - source term: spent fuel corrosion in water - spent fuel: impact load resistance and storage conclusions and perspectives

3 Joint Research Centre (JRC) IAEA, International Experts Meeting, Vienna, March European Atomic Energy Community (EURATOM) JRC: 7 research institutes in 5 EU countries ~2500 staff / 300 M /a budget / 40 M income Nuclear programme within the JRC Nuclear Data, Reference Materials and Measurements Fundamental Properties of Nuclear Materials and Applications Waste Management and Environment Reactors Safety Fuels and Fuel Cycles Safety Safeguards, Non-proliferation & Security

4 Institute for Transuranium Elements IAEA, International Experts Meeting, Vienna, March Reference Centre for policy makers, stakeholders and citizens in the nuclear field NUCLEAR SAFETY and NUCLEAR SECURITY Basic science, Fundamental properties & Applications Safety of nuclear fuel cycle / Nuclear waste / Environment Nuclear safeguards, Non- Proliferation & security Training & Education Exploratory/Discovery Research The mission of JRC-ITU is to provide the scientific foundation for the protection of the European citizen against risks associated with the handling and storage of highly radioactive material

5 Safety of nuclear fuel cycle at ITU IAEA, International Experts Meeting, Vienna, March Conventional, Advanced Nuclear Fuels and Cycles - samples synthesis, materials science studies - PIE: safety during irradiation, (severe) accidents - back-end: storage, disposal, P&T - predictive tools: TRANSURANUS, multi-scale fundamental approach From basic actinide science, to atomistic mechanisms to operational fuel properties LWR fuel experience is basis for studies on advanced fuels

6 Nuclear fuel studies at ITU IAEA, International Experts Meeting, Vienna, March Competences thermodynamics (C p, vapor pressure, melting point) thermal transport fission products, gases, minor actinides: phase distribution, matter transport radiation damage: mechanisms and effects fuel restructuring microstructure macroscopic properties evolution corrosion, creep Scope (fuels) LWR advanced reactors HTR high burnup U, Th MOX non-oxides minor actinides cladding/coating normal/off-normal operation extreme conditions storage multidisciplinary approach analytical/modeling tools Experimental tools samples synthesis (MA lab) optical, acoustic microscopy; SEM EPMA, SIMS; TEM-SEM; XRD th. conductivity: laserflash, POLARIS high T laser-heating (melting, vaporization, conductivity, high-p) high T effusion, revaporization, annealing, KüFA (HTR) non destructive rod examination: profilometry, radiography, outer oxide layer, g-spectrometry clad: H 2 -hydrides, creep, burst (hot) indentation, impact-fracture fission gas release, density chemical analysis, laser ablation separation (aqueous, pyro-) leaching, electrochemistry

7 Fuel under extreme conditions T/K IAEA, International Experts Meeting, Vienna, March Severe accidents programmes/networks: TMI, Phebus, CIT, Coloss, SARNET II irradiated fuel from real and simulated accidents High T behaviour volatiles, fission gas release; vapour pressure up to complete fuel matrix vaporization; thermophysical properties Basic thermodynamic data (T m, phase diagrams) actinides/fuel compounds, corium and other systems Spent fuel rod safety during storage and transport mechanical stability U-Pu oxide system Liq u idso lu tio n So lid so lu tio n PuO 2 (mol%) From conventional to advanced fuels safety UO 2, MOX, Th-MOX, HTR (Küfa), metal alloy, minor actinides, inert matrices, emerging concepts

8 Severe accidents projects (selection) IAEA, International Experts Meeting, Vienna, March Three Mile Island (TMI-2). A real accident; integral test (with incomplete data). OECD-NEA led consortium under the initiative of US-DoE (INL) involving AEA, AECL, ANL, CEA, KIT, JAEA, JRC-ITU, PSI, Studsvik. Phebus FP test. Irradiated fuel bundle degradation and melting ( ). Integral test with good data collection, but still difficulties in interpretation. Led by IRSN (France) and supported by the European Commission. USA, Canada, Japan, Korea and Switzerland also participated. Five integral tests under different conditions. On-line monitoring of bundle degradation, fp release and subsequent behaviour in the simulated primary circuit and containment. Corium Interaction Thermochemistry. EC Framework project, 8 partners ( ) small scale tests of liquid Zry dissolution of (irradiated) UO 2, and modelling Core Loss of geometry. High burnup UO 2, MOX high T interaction with cladding Revaporisation testing. EC Framework project, 3 partners single effect tests of (re)volatilisation of fission products under different atmospheres

9 TMI-2 IAEA, International Experts Meeting, Vienna, March ferrous (Fe, Ni, Cr) U-rich Zr-rich fuel rod remnant C debris samples H upper crust D8- P2,3 Core bore rock G12-P9-B- (1000x, BSE); phase density variations fully molten rock core bore rocks G12-P2-E, G12-P6-E, G12-P9-B, G12-P10-A lower crust N5-P1-E O7-P4 Debris H (40x); white pieces are UO 2 ; long grey piece is zircaloy-uo 2 mix; banded structure is zircaloy interacted with steel, Ni-based alloys

10 Zircon crystal from Chernobyl lava IAEA, International Experts Meeting, Vienna, March Pöml, Burakov et al., 2011 SIMS 235 U enrichment: in zircon 1.08%; in UO x inclusions 0.8% BSE EPMA Zircon: wt.% UO 2 (natural <1.5), Pu traces UO x : wt.% PuO 2, Zr traces Si U Zr

11 PIE of Phebus bundle IAEA, International Experts Meeting, Vienna, March Zone with melt ITU contribution: - sectioning of degraded bundle into 14x2 cm discs - microscopic examination and analysis at selected points of the bundle to establish the principal interactions - examination of PTA samples (filters) FPT2 Disc 2 (lower surface) +51mm Central rod missing melt on north side good correspondence between tomography and sectional macrographs

12 PIE of Phebus bundle-coring IAEA, International Experts Meeting, Vienna, March Degraded fuel rod 7 with cladding broken away Corium Fully oxidised cladding Metallic melt Molten materials with filigree structure Wood s metal Microscopic sample extracted by coring from Disc 2, FPT2, on lower surface at +51.5mm BFC & its position in the disc tomography

13 Outcome IAEA, International Experts Meeting, Vienna, March What type of information comes out of these studies mechanistic: mechanisms, rate thermodynamic: temperature, oxygen potentials thresholds: key materials, specific interactions & transition T (eg. T m ) Corium molten pool forms in a predictable geometry. Composition ~(U,Zr)O 2. Rapid cooldown leaves corium as a single, deformed cubic phase, slower cooling results in formation of separate U-rich & Zr-rich oxides Samples reveal how Ag-Zr and Ni- (or Fe)-Zr interactions can create liquefied cladding already by 1200 C (over 1700 C below UO 2 melting) which can rapidly attack the fuel Irradiated fuel undergoes a more rapid degradation than non-irradiated fuel, because -it is mechanically weak (pre-existing cracks) -fg release & precipitation into bubbles lead to very high porosity: 'foaming' at very high T -increased surface area for attack by corium Cs release <100%, some Cs remains in the overheated fuel and even in the melt pool Cs condenses on cooler surfaces (<700 C), but can easily revolatilise above 500 C in steam (also in inert atmospheres) probably as CsOH - regardless of the deposit composition

14 IAEA, International Experts Meeting, Vienna, March Spent Fuel Safety in the Light of the Accident at the Fukushima Daiichi Nuclear Power Plant characterization of molten fuel/corium extended to cover specific aspects relevant for the Fukushima analysis and remediation refocusing activities - source term: high T properties and behaviour - source term: spent fuel corrosion in water - spent fuel: impact load resistance and storage

15 Normalized Fractional Release Source term studies: thermal release IAEA, International Experts Meeting, Vienna, March Knudsen cell for effusion tests on irradiated fuel mm SAMPLE 14 7 W 1 5 SAMPLE Al2O3 1 TC mm 10 Gas inlet To Q-GAMES (Quantitative Gas Measurement system) TEMP. MS, RANGE AMU 9 steps To PRIMARY VACUUM 11 Ramp 10-50K/min TIME Normalized fractional release of ~70 GWd/t UO 2 central pellet region 4He 86kr 96ZrO 129I 130Te 136Xe 137Cs Gamma, Bata Counts Temperature (K) 138Ba 139La 140Ce 139LaO 140CeO 238UO 88Sr 239PuO 87Rb 153Eu 150Sm

16 Release quantity (kg/s) O/M mass signal (A) O/U Source term: oxidation effects Sample completely vaporised IAEA, International Experts Meeting, Vienna, March Hiernaut et al., E-8 1E-10 1E-11 1E Sr 129 I 130 Te 137 Cs BaO NdO UO 2 (a) effusion behaviour (b) Combined Knudsen cell SEM analysis morphology of ~65 GWd/t UO 2 annealed at 1900 K in vacuum in vacuum 1E-13 T d <U 4 O 9 > T d < -U 3 O 8 > Temperature (K) (a) E-10 1E-11 Cs BaO SrO UO x tot U3 O 8 UO 2 (b) preoxidized preoxidized 1E-12 1E Temperature (K) outer surface fracture surface

17 FRNU Source term: water corrosion IAEA, International Experts Meeting, Vienna, March ,1 0,01 after Fukushima: spent fuel leaching in seawater ongoing Fractional Release Normalized to U; leaching of 35 GWd/t MOX in groundwater Fragments (A, Filtered) Time (days) FRNU>1: Mo, Cs, Rb, Ba, Tc, Np, Sr(Zr)90 FRNU 1: Y, Nd, (Np), Pu ( 1) FRNU<1: Ru, Pd, Zr, Pu ( 1) Rb85 Cs133 Mo98 Zr/Sr90 Np237 Y89 Rh103 Ba138 Nd144 Pu240 U238 Ru102 Tc99 Zr93 Pd105 groundwater leaching tests Secondary phases on leached UO 2 Studtite Schoepite UO 2-10% 238 Pu UO 2-0,1% 238 Pu UO 2

18 Fuel safety out of pile IAEA, International Experts Meeting, Vienna, March Impact load response of a ~74 GWd/t PWR rod (high speed camera sequence) a b c d simulated impact accident during spent fuel rods transportation safety of spent fuel storage/transport corrosion layer removal zone PWR and BWR rods tested: GWd/t fuel release <2 g/break i.e. less than the mass of a single fuel pellet D. Papaioannou et al., 2009 GNS-AREVA collaboration

19 Spent fuel storage IAEA, International Experts Meeting, Vienna, March No direct spent fuel data in the extended range; decay damage saturation? swelling? mechanical integrity? He accumulation during storage exceeds solubility level. Will it all accommodate in defects, fg bubbles? Ongoing work to elucidate conditions and mechanisms relevant for storage: - spent fuel swelling/pressurization; response to long term (low) T history - cladding properties evolution - microstructure alterations at high dose - fuel composition/irradiation history effect He produced per g of fuel, g decay and He production in spent fuel ~100 dpa ~10 dpa ~1 dpa eol Time from discharge, years Approach: ~0.01 dpa fuel 45% Pu He solubility UO 2 40 GWd/tM UO 2 60GWd/tM UO 2 80 GWd/tM UO GWd/tM MOX 25 GWd/tM MOX 45 GWd/tM MOX 60 GWd/tM spent fuel characterization - accelerated -decay, He accumulation - He solubility, thermodynamic equilibrium -decays g -1

20 Conclusions and perspectives IAEA, International Experts Meeting, Vienna, March Significant amount of knowledge on fuel behaviour during severe accidents exists from international projects on fuel from actual or simulated severe accidents New programmes are proposed to extend the experimental basis of data for modeling tools and fill some gaps. This will benefit from advances in experimental characterization tools In JRC-ITU, some R&D activities on fuel safety are refocused to cover specific issues related to the Fukushima accident and to its aftermath, e.g. molten fuel/corium properties, source term assessment for high T release, corrosion effects in seawater/salt, spent fuel behaviour in the pools, storage/treatment of molten fuel, etc. Links/collaboration with Japanese partners (CRIEPI, JAEA) are being developed Integrated approaches are mandatory to optimize use of resources (less money and time than in the past) and to investigate all systems international partnerships/programmes integrated experimental/theoretical

21 IAEA, International Experts Meeting, Vienna, March