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Eleanore Summers
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Institutt for Energiteknikk OECD Halden Reactor Project Joint International Program - VVER Fuel Presented by: Dr Margaret McGrath Presentation Outline The OECD Halden Reactor Project

1 Institutt for Energiteknikk OECD Halden Reactor Project Joint International Program - VVER Fuel Presented by: Dr Margaret McGrath Presentation Outline The OECD Halden Reactor Project Capabilities for Fuel Testing The Joint International Program - VVER Long-Term Fuel Performance Power Ramping LOCA Tolerable Rod Internal Pressure Cladding Creep Concluding Remarks Page 2 1

The OECD Halden Reactor Project An international co-operative effort Hosted by Institutt for Energiteknikk (IFE) Under auspices of OECD / NEA - Paris Aimed to improve safety at NPPs It is jointly

2 The OECD Halden Reactor Project An international co-operative effort Hosted by Institutt for Energiteknikk (IFE) Under auspices of OECD / NEA - Paris Aimed to improve safety at NPPs It is jointly funded by its Members 17 countries (for ) More than 100 organisations The Members are: Utilities Vendors Licensing authorities & regulators R&D centres The HRP staff ~275 people Page 3 Current Signatories and Associated Parties Belgium (SCK/CEN Mol) Czech Republic (NRI Rez) Denmark (Risø national Lab.) Finland (Stuk, Ministry of T&I) France (Electricité de France) Germany (GRS) Hungary (KFKI) Japan (JAERI) Korea (KAERI) Norway (IFE) Russia (Kurchatov, TVEL) Slovakia (Vuje) Spain (CIEMAT) Sweden (SKI) Switzerland (HSK) United Kingdom (BNFL) United States (USNRC) + GE, EPRI, Westinghouse Page 4 2

Capabilities for Fuel Testing The HBWR test reactor 20 MW, 34 bar, 235 C, UO 2 fuel rods Loops enable simulation of operation conditions of different commercial NPPs ~100 day operation cycles

(temperature, pressure, flux) Capable of measuring a wide range of performance parameters A hot-cell facility Production of fuel for driver elements for cladding related tests Re-fabrication and

3 Capabilities for Fuel Testing The HBWR test reactor 20 MW, 34 bar, 235 C, UO 2 fuel rods Loops enable simulation of operation conditions of different commercial NPPs ~100 day operation cycles Instrumented test rigs (IFAs) Designed to implement different local modes of operation (steady state, load follow, ramping, transients) In-core instrumentation Developed to be robust and reliable (temperature, pressure, flux) Capable of measuring a wide range of performance parameters A hot-cell facility Production of fuel for driver elements for cladding related tests Re-fabrication and instrumentation of commercial fuel and materials Post irradiation examination (PIE) Page 5 The Halden Reactor S61 S53 D06 S30 S35 S47 S S39 S26 16 E06 E03 17 DO4 S65 C04 S37 S60 31 S S49 33 S E01 E02 23 D02 D01 D07 C S23 21 S45 S D10 S S75 S31 S57 D S17 34 S50 35 S EC 602 D05 CS S S S34 E D09 S56 S54 S51 29 S E08 26 E05 S81 S77 S48 S42 S E07 S62 S58 S82 S64 38 S D03 36 S21 S S55 19 S76 S S S46 S78 D08 S79 S80 S73 S83 S59 T1 T2 Hexagonal array of 300 channels within pressure tank Rings 1-6 used for fuel About 30 / 110 positions used for experimental purposes Height of active core is 80 cm Experimental channel Ø: - 71 mm in HBWR moderator mm in pressure flasks Pressure flasks connected to loops for different chemistry & thermal hydraulic conditions 8-10 loops operating Booster fuel rigs to increase local fast flux Page 6 3

An Instrumented Test Rig (IFA) In-core Connector Outlet Coolant Thermocouples Fuel Centre - line Thermocouple Fuel Rod storage position

Transformer (LVDT) Inlet Coolant Thermocouples Inlet Turbine Flowmeter Calibration Valve Page 7 In-Core Fuel Instrumentation Primary

change (monitoring densification/swelling) Clad elongation (monitoring PCMI effects) Clad diameter (creep and stress relaxation) Neutron

4 An Instrumented Test Rig (IFA) In-core Connector Outlet Coolant Thermocouples Fuel Centre - line Thermocouple Fuel Rod storage position Neutron absorber Pressure transducer (PF) Shroud (Ø 73/71mm) He-3 coil Fuel Rod ramp position Neutron Detector (V-type) Differential Transformer (LVDT) Inlet Coolant Thermocouples Inlet Turbine Flowmeter Calibration Valve Page 7 In-Core Fuel Instrumentation Primary Measurements Fuel temperature (at pellet centre) Rod pressure (variations show fuel densification / fission gas release) Fuel stack length change (monitoring densification/swelling) Clad elongation (monitoring PCMI effects) Clad diameter (creep and stress relaxation) Neutron flux / power Secondary Measurements Hydraulic diameter (fuel-clad gap) Gamma spectrometry (fission gas products) Noise analysis from instrument signals Page 8 4

5 Fuel Thermocouple and Rod Pressure Sensor Fuel Segment In-core contact ambient conditions: 325 C and 165 bar Fuel thermocouple Max. Temp:2300 C Pressure sensor: range: bar Page 9 Joint International Program VVER Fuel High Burn-Up Fuel Behaviour - Normal Operation Long-term behaviour from zero burn-up, monitoring fuel densification, swelling, thermal performance and FGR Comparative testing: additive and doped fuels, VVER fuels, inert matrix fuels and MOX fuels Power ramping of re-fabricated fuel rods to study PCMI Separate effects testing: cladding creep and fission-creep of fuel Tolerable internal rod pressure Fuel Response to Transients LOCA test series (PWR, BWR, VVER fuels) Clad Corrosion and Water Chemistry Issues PWR corrosion studies and BWR crud studies Plant Lifetime Assessments IASCC crack initiation & growth studies Stress relaxation Reactor pressure vessel integrity Page 10 5

6 1. Long-Term Fuel Performance Test series to irradiate fresh VVER fuel rods to investigate: Fuel dimensional behaviour at BOL and with burn-up PCMI, thermal behaviour and FGR with burn-up IFA ( ) specifically to compare VVER- 440 fuel with PWR reference fuel (annular versus solid) IFA ( ) extending IFA to include 3 variants of VVER-440 fuel (again with PWR fuel as reference) IFA ( ) to investigate VVER-1000 additive fuel (with large grain size: 26 μm) and Gd-doped VVER fuel (5wt%) with VVER-440 reference fuel Page 11 IFA (in HBWR) Multiple-instrumented rods to study different parameters in same rod (better understanding of a given phenomenon) Temp, pressure, fuel length Temp, fuel length, clad length Page 12 6

performance at BOL Gd-doped fuel temperature will be affected by changing radial power

subsequent (black) and final (blue) ramps Densification in large grain (W+) and reference

7 IFA Power and Fuel Temperature No significant effect of large grains on fuel thermal performance at BOL Gd-doped fuel temperature will be affected by changing radial power profile (BU) Page 13 IFA Fuel Elongation versus Power 1st power-up ramp (red) compared to subsequent (black) and final (blue) ramps Densification in large grain (W+) and reference (Wr) fuel similar More stable dimension behaviour of W+ at power (lower creep) No densification observed in Gd-doped fuel (also observed in Gd PWR fuel) W+ Wr Gd-doped Page 14 7

8 IFA Clad Elongation at BOL Clad elongation (red) and rod ALHR (blue) shown over initial period Early-in-life PCMI indicated in both large grain (W+) and reference (Wr) fuel rods Similar accommodation effect of the two fuel pellets types within the cladding (reduction in clad elongation with time at stable power) no permanent strain W+ Wr Page Power Ramping: IFA Aim is to study thermal performance, FGR and mechanical behaviour of re-fabricated fuel segments from Loviisa NPP (47-48 MWd/kgU) Two ~400 mm segments instrumented with a fuel centreline thermocouple and either a cladding elongation detector or rod internal pressure sensor Rods subjected to successive power ramps up to a maximum achievable power (27-28 kw/m), which was then held for days 900 C (measured) was reached in ~50 C steps but no FGR was observed (1% FGR threshold is ~1075 C) Clad elongation (EC) reacts to PCMI and shows clad relaxation due to fuel creep Rig moved to higher power position for repeat testing in 2008 Page 16 8

3. LOCA Test series to address LOCA issues using refabricated LWR high burn-up fuel segments (PWR, BWR, VVER) VVER fuel test in IFA-650.

9 3. LOCA Test series to address LOCA issues using refabricated LWR high burn-up fuel segments (PWR, BWR, VVER) VVER fuel test in IFA planned to meet the following primary objectives: Maximize the balloon size to promote fuel relocation so as to evaluate effect on the cladding temperature and oxidation Measure the extent of fuel fragment relocation into the ballooned region by gamma scanning Investigate the extent of any secondary transient hydriding on the inner side of the cladding around the burst region Page 17 LOCA Loop Test Rig IFA rod instrumentation: 3 cladding thermocouples 1 rod pressure transducer 1 clad elongation detector (faulty) IFA rig instrumentation: Electrical heater (in flow separator) 3 heater thermocouples 2 coolant inlet thermocouples 2 coolant outlet thermocouples 3 vanadium neutron detectors 2 cobalt neutron detectors He-3 coil (not used) Page 18 9

10 IFA Test Scheme and Execution In 2007 a test of a 4 cycle fuel rod from Loviisa NPP (56 MWd/kgU) re-fabricated and filled with 95 % Ar / 5 % He to 30 bar (RT) was carried out Operational rod power of ~12 W/cm Heater power set to 13.5 W/cm (to simulate neighbouring rods) Blow-down initiated by opening valves to a dump tank - coolant flows out from the bottom of the rig within 2 minutes Cladding heated up with temperature increase rate ~3 to 1 C/s (slowed down as target peak cladding temperature approached) Rod failure at ~800 C (PCT of 830 C achieved) Water spray used from ~800 C (to generate/maintain steam) Test terminated by reactor scram Gamma scanning at Halden carried out to verify cladding deformation and fuel relocation (ballooning indicated) Fuel rod fixed in flask with epoxy resin, PIE to be carried out in 2008 Page Tolerable Rod Internal Pressure: IFA-610.x Outlet thermocouple Gas line Fuel thermocouple Fuel rod Booster rods Gas line Inlet thermocouple Pressure flask To address integral fuel rod behaviour during operation with simulated EOL over-pressure Re-fabricated, commercially irradiated fuel segment in a pressure flask rig with booster fuel rods for fast flux and gas lines for increasing rod internal pressure Gas flow lines also used for in-pile hydraulic diameter measurement Segment instrumented with fuel centreline thermocouple with in-core connector (re-useable) and cladding elongation detector PWR and BWR segments tested, VVER segments planned to be tested in 2009 Page 20 10

Illustration of Test Capability: IFA-610.

Normalised to 15 kw/m Normalised fuel temperature Shows clear response to level of

of thermal feedback occurs at about 138 bar overpressure This represents the lift-off

11 Illustration of Test Capability: IFA (PWR) Rod pressure Increased in steps of ~50 bar maximum 470 bar Fuel Temperature Normalised to 15 kw/m Normalised fuel temperature Shows clear response to level of overpressure and direct effect of pressure step Page 21 Illustration of Test Results: IFA (PWR) The rate of temperature increase is correlated with the overpressure The onset of thermal feedback occurs at about 138 bar overpressure This represents the lift-off threshold for the particular combination of fuel and cladding utilised in the test Below this threshold, any clad creepout is sufficiently compensated, e.g. by fuel swelling, such that no net thermal feedback becomes apparent Page 22 11

rods for fast flux and gas lines for varying rod internal pressure Variable stress history including stress reversals Contact scanning diameter gauge for monitoring clad diameter change (in

12 5. Cladding Creep: IFA To generate in-pile creep data from modern commercial cladding alloys for use with fuel performance modelling Fresh fuelled cladding segments in a pressure flask rig (PWR loop) with booster fuel rods for fast flux and gas lines for varying rod internal pressure Variable stress history including stress reversals Contact scanning diameter gauge for monitoring clad diameter change (in particular to study primary creep) E110 cladding conditions: coolant C, mid-wall ~365 C fast neutron flux ~1.3E+13 n/cm 2 /s hoop stress -75 MPa and -50 MPa OD / ID is 9.50 / 8.35 mm Test started July 2007, planned to end 2010 Page 23 IFA Creep Curve for E110 5,0E-4 0,0E+0 E110-75MPa -50MPa Creep strain -5,0E-4-1,0E-3-1,5E-3-2,0E FPH PRELIMINARY DATA Page 24 12

13 Concluding Remarks Since 1995, VVER fuel testing has been an increasingly important part of the HRP Joint Program Current and planned testing: Fuel performance (IFA-676) continuing to ~60 MWd/kg oxide LOCA (IFA-650) PIE during 2008 plus VVER part of the series Tolerable rod pressure (IFA-610) loading planned for 2009 Power ramping (IFA-700) re-loaded for 2008 Cladding creep (IFA-699) continuing to 2010 (~14,000 fph) HRP Membership operating VVER NPPs consists of: (Bulgaria), Czech Republic, Finland, Hungary, Russia, Slovak Republic Acknowledgement to recent staff working on VVER experiments: Radomir Josek (CZ), Laura Kekkonen (FI), Erik Kolstad, Barbara Oberländer, Barbara Somfai (HU), Terje Tverberg, Boris Volkov (RU) Page 25 Recent HRP contributions to open literature related to VVER fuel testing 6th International Conference on VVER Fuel Performance, Modelling and Experiment Support, Albena, Bulgaria, September, th International Conference on VVER Fuel Performance, Modelling and Experiment Support, Albena, Bulgaria, September 2007 Page 26 13

Fission gas release and temperature data from instrumented high burnup LWR fuel

Fission gas release and temperature data from instrumented high burnup LWR fuel XA0202217 T. Tverberg, W. Wiesenack Institutt for Energiteknikk, OECD Halden Reactor Project, Norway Abstract.The in-pile