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Investigation of Feasibility of Incorporation of Hydride in Fuels Advanced Test Reactor National Scientific User Facility (ATR NSUF). Donald Olander 1 , Kurt Terrani 1,4 , Tom Newton 2 , Gordon Kohse 2 , Lin-wen Hu 2 , David Carpenter 2 Mitch Meyer 3 , Jim Cole 3 , Joy Rempe 3

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  1. Investigation of Feasibility of Incorporation of Hydride in FuelsAdvanced Test Reactor National Scientific User Facility (ATR NSUF) Donald Olander1, Kurt Terrani1,4, Tom Newton2, Gordon Kohse2, Lin-wen Hu2, David Carpenter2 Mitch Meyer3 , Jim Cole3 , Joy Rempe3 1 University of California, Berkeley 2 Massachusetts Institute of Technology 3Idaho National Laboratory 4Oak Ridge National Laboratory Work supported by the U. S. Department of Energy, Office of Nuclear Energy under DOE Idaho Operations Office Contract DE-AC07-051D14517, as part of an ATR National Scientific User Facility experiment.

  2. Outline • Introduction to LWR Hydride Fuel with liquid metal gap filled Concept • Laboratory Experiment • Irradiation Experiment • Temperature, power measurement • Thermal conductivity deduction • Cover gas analysis • Proposed post Irradiation experiments • Summary

  3. Liquid-metal-bonded fuel rod concept Conventional fuel rod Proposed fuel rod He LM End plug Spring He gap Cladding pellet stack UO2 U-ZrH1.6

  4. Black: U Gray: ZrH1.6 What is a hydride fuel? TRIGA fuel: uranium metal + zirconium hydride - (U,Zr)H1.6 up to 45 wt % or 21 vol% a-U dispersed in d-ZrH1.6 matrix Limiting U content Why so much U? U density of hydride only40% that of UO2 For the same linear power, need ~ 10% enriched in U235 NRC regs? 10 mm (U,Th,Zr)Hx probably a better fuel

  5. Hydride fuel+ Liquid-metal Gap: A Better LWR Fuel? Metal Hydrides • Space-Nuclear-Auxiliary Power (SNAP) Program NASA 1960 –70 • TRIGA Research reactors – since 1957 • Control rod for fast reactors: U.S. Navy Liquid Metals • Sodium-cooled fast reactor EBR II, Phenix, JOYO • Lead-cooled fast reactor (Gen IV) But: can these two technologies be combined in fuel rods for LWRs?

  6. Why Hydride in Place of Oxide? • Moderator (H) is in the fuel => reduce volume of water => smaller core, pressure vessel –or, higher power from same core • Higher LHR possible – higher burnup (enrichment limited) • Improved safety: faster negative feedback than oxide fuel (TRIGA) • Lower fuel temp. khyd~ 6 x kox (Tmax < 650oC) • - reduced FP release • - reduced stored energy

  7. Thermal conditions in LWR rods

  8. Fission-gas release from U-ZrH1.6 Note: data are old (1960 – 1980) and poorly documented

  9. What causes this? Fission- product swelling 3x that of UO2 !!! Swelling of U-ZrH1.6 SNAP program (1965) data

  10. Fuel-cladding chemical interaction • Zircaloy is a powerful sink for hydrogen available from the fuel cladding fuel At 700oC: pH2(fuel) ~ 10-1 atm; pH2(cladding) ~10-4 atm

  11. Fuel cracking? • Thermal stress – tensile at periphery • Hydrogen redistribution – transports H from the center to the surface; generates compression at the surface Hydrogen redistribution in temperature gradient Total stress is compressive at periphery – prevents cracking?

  12. SEM versus AFM Elastic modulus mapping across the microstructure of (U4Th2Zr9)H1.5 fuel Backscattering SEM Modified AFM • a-U phase modulus is 210 GPa • d-ZrH 1.6 phase modulus is 125 GPa • The elastic modulus of ThZr2H7-x phase is determined 172GPa for the first time

  13. Compatibility with Zircaloy cladding • Zircaloy samples pressed against hydride fuel at various contact pressures immersed in liquid metal at 375 ºC for one month 20 µm gap 120 MPa contact - TERRANI, K., et al., “Liquid Metal as a Gap Filler to Protect Zircaloy Cladding from Hydride Fuel,” Proceedings of Top Fuel 2009, Paris, France

  14. Irradiation Experiment

  15. Objectives • Design, construct and irradiate a mini-fuel element under realistic LWR conditions • At stack midplane: • maintain Tmax ~600oC • >10% U-235 burnup • gap closure • At ends of stack • Tmax ~ 500oC • gap remains open clad fuel ~12 cm LM • On-line temperature read outs • on-line fission-gas monitoring

  16. Pellet Fabrication TRIGA Fuel Slug Centerless Grinding Apparatus Diamond Core Drills Rough Pellets Post Drilling Smooth Pellets Post Grinding U(30wt%)-ZrH1.6 19.7% U-235

  17. Mini Fuel element assembly Sheath TC Welded to SS Flange SS304 CF Mini Flange • Thermal conductivity measured by: • Two thermocouples; fuel centerline and rod surface Zr CF Mini Flange He Plenum SS302 Spring Pb-Bi Alloy Alumina Spacer Zircaloy-2 Tube U0.17ZrH1.6 Fuel Zircaloy-2 End Cap 1 cm

  18. Neutron radiography at MITR with resolution of ~ 100 μm Active Fuel Region Alumina Spacers Rod 1 Zirconium Flange SS 304 Flange Rod 2 Rod 3 302 SS Spring Rod 4

  19. Hydride Fuel Irradiation (HYFI) Experiment :location of fuels rods Center TC Cover gas line Flux profile LM Capsule 3 (Rod 2 MIT Reactor Core Capsule 2 (Rod 3) FCT-3 and CST-3) Ti Capsule 1 (Rod 1) FCT-1 and CST-1 Pellet • Assemblies simultaneously irradiated at each time • One assembly removed every 4 months (Burnup dependent data) • Longest assembly to remain within the core for 1 year (0.30% Fission of initial metal atom, FIMA)

  20. Test matrix for the HYFI irradiation of fuel rods Abandoned

  21. Temperature and thermal power profile of fuel rods 1, 2 and 3 since March 2011 Note: Unpredicted frequent shot downs and ramp ups

  22. Thermal conductivity calculation through annual fuel pellet Oxidized call Neutronic (MCNP) calculations For 5 MW: R r0 T0 Ti TLMcool H, TTC Ts

  23. Time dependence of the thermal conductivity k did not change with at the beginning of irradiation

  24. Lower capsule thermal conductivity estimate k12 Minimum 0.15856169 Maximum 0.1770716 Points 14244 Mean 0.1701022 Median 0.17003919 Std 0.0011548277d

  25. Thermal conductivity variation of three fuel rods during irradiation • The initial rise is attributed to the lag time of thermal power with respect to TC readouts • What is the reduction of deduced thermal conductivity due ?: • Large initial swelling • Good retention of fission gas products • Hydrogen redistribution in the fuel • Or Oxidation of clad, formation of bubbles in LM, configurations changes with time within the duct holding capsules

  26. Thermal conductivity change on Capsule 3 by Jan. 2012 The drastic reduction in deduced-conductivity is of concern. Needs to be verified by post irradiation analysis

  27. Fission products release due to possible leak in capsule 1

  28. Burnup estimate Neutronic (MCNP) calculations for 5MW: H,

  29. Comparison of In-pile Thermal Conductivity of U(30wt%)-ZrH1.6 and UO2

  30. Structural Damage at MIT NRL • On November 2011 • Increasing fission gas release from Capsule 1 triggers its removal, is replaced with the dummy from wet storage • On January 24, 2012 • Damage to bottom spacer and is identified, some missing alignment pins on capsules are • All parts except bottom spacer moved to storage locations Bottom Spacer Capsule 1 It has been decided to terminate the irradiation at this time and send the three capsules to INL for post irradiation analysis

  31. PIE to determine irradiation effects on fuel & cladding • Fission-gas behavior • Bubbles in LM bond - state in fuel • Fission-product swelling • Verify SNAP data • Fuel-cladding chemical interaction (hydriding) • Can LM protect Zircaloy cladding from attack by fuel? • Fuel cracking • Do pellet chips in gap stress cladding? - Do cracked wedges close gap?

  32. Post-irradiation examination

  33. Summary • Irradiation of hydride fuels has started in March 2011 and terminated January 2012 • Clad and centerline temperature, reactor power as well as fission products release were monitored in time • Out of 5 mini rods prepared, three were actually irradiated • Post irradiation analysis will be accomplished at INL • References • K. Terrani, J. Seifried, D. Olander, “Transient Hydride Fuel Behavior in LWRs,”J. Nuc. Mat., 392, (2009) 192. • D.R. Olander, E. Greenspan, H.D. Garkisch, B. Petrovic, “Uranium-zirconium hydride fuel properties,”Nucl. Eng. Design, 239, (2009) 1406. • Kurt A. Terrani, Mehdi Balooch, Gordon Kohse, David Carpenter, Lin-wen Hu, Mitchell K. Meyer, Donald Olander “In-Pile Thermal Conductivity Measurement of Uranium-Zirconium Hydride Fuel” Nuclear Fuels and Structural Materials for the Next Generation Nuclear Reactors (NFSM 2012) June 24–28, 2012 , Chicago, IL

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