1. Investigation of Feasibility of Incorporation of Hydride in Fuels Advanced 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. 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 only 40% that of UO2
For the same linear power, need ~ 10% enriched in U235
NRC regs?
Black: U
Gray: ZrH1.6
10 mm
(U,Th,Zr)Hx probably a better fuel

5. Hydride fuel+ Liquid-metal Gap: A Better LWR Fuel?
Space-Nuclear-Auxiliary Power (SNAP) Program NASA 1960 –70
TRIGA Research reactors – since 1957
Control rod for fast reactors: U.S. Navy
Metal Hydrides
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
hydride
oxide
characteristic
He bond
LM bond
pellet OD, mm 10
LHR, W/cm
375
fuel centerline, oC
680
555
1505
Tfuel
170
995
Tgap (35 m)
125 1
Tclad 46
Tfluid 39
coolant, oC
300

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

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

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

11. Hydrogen redistribution in temperature gradient
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
At UCB,
We have investigated un-irradiated fuel to some extend. For example we have mapped the mechanical properties of Th-U-Zr hydride with a home made modified AFM and obtained the elastic moduli of the three phases
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
We also invesigated the interaction of Zr with hydride as a function of pressure with and without LM in between and also with Zr coated with thin oxide. The thin oxide prevented the hydride formation of Zr at temperature of 375 C up t one month.
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
fuel LM
clad
~12 cm
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%)-ZrH % U-235

17. Mini Fuel element assembly
Sheath TC
Welded to SS Flange
SS304 CF Mini Flange
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
Thermal conductivity measured by:
Two thermocouples; fuel centerline and rod surface

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

19. Hydride Fuel Irradiation (HYFI) Experiment : location of fuels rods
Center TC
Cover gas line
Flux profile LM
Capsule 3 (Rod 2
Capsule 2 (Rod 3)
FCT-3 and CST-3) Ti
Capsule 1 (Rod 1)
FCT-1 and CST-1
Pellet
MIT Reactor Core
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
Irradiation Position
Irradiation dates
March 20, 2011
May 26, 2011
June 25, 2011
Nov
Jan 20, 2012
Early
2012
Top
Dummy
Fuel rod 2
Fuel rod 4
(w /thermal cond. Probe)
Middle
Fuel rod 3
Fuel rod 5
(W He filled gap)
Bottom
Fuel rod 1
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 pelletTs
TTC
TLMcool Ti
Oxidized call
Neutronic (MCNP) calculations
For 5 MW: H,

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
Maximum
Points 14244
Mean
Median
Std d

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
Phenomenon
Instrumentation
Fuel cracking
OM, SEM
Xe in the plenum gas
Mass Spec.
Gas bubbles in the frozen alloy OM
Hydrogen distribution in fuel
SIMS
Hydride precipitates in cladding
Uranium particles
SEM, AFM
Void around U particles
TEM, AFM
Diametral expansion of fuel and cladding
Micrometer
Second-phase particles in fuel

33. Summary
Irradiation of hydride fuels has started in March 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