investigation of feasibility of incorporation of hydride in fuels advanced test reactor national...
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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, Berkeley2 Massachusetts Institute of Technology
3Idaho National Laboratory4Oak 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.
Outline
o Introduction to LWR Hydride Fuel with liquid metal gap filled Concept
o Laboratory Experiment
o Irradiation Experiment
• Temperature, power measurement
• Thermal conductivity deduction
• Cover gas analysis
o Proposed post Irradiation experiments
o Summary
2
Liquid-metal-bonded fuel rod concept
pellet stack
Cladding
He LMSpring
End plug
He
gap
UO2 U-ZrH1.6
Conventional fuel rod Proposed fuel rod
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 -ZrH1.6 matrix
10 m
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
(U,Th,Zr)Hx probably a better fuel
Hydride fuel+ Liquid-metal Gap: A Better LWR Fuel?
But: can these two technologies be combined in fuel rods for LWRs?
• 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)
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
Thermal conditions in LWR rodshydride hydride oxide
characteristic He bond LM bond He bondpellet OD, mm 10 10 10
LHR, W/cm 375 375 375
fuel centerline, oC 680 555 1505
Tfuel 170 170 995
Tgap (35 m) 125 1 125
Tclad 46 46 46
Tfluid 39 39 39
coolant, oC 300 300 300
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 !!!
Fuel-cladding chemical interaction• Zircaloy is a powerful sink for hydrogen available
from the fuel
At 700oC: pH2(fuel) ~ 10-1 atm; pH2(cladding) ~10-4 atmfuelcladding
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?
11
SEM versus AFM Elastic modulus mapping across the microstructure of
(U4Th2Zr9)H1.5 fuel
100
120
140
160
180
200
220
0 0.5 1 1.5 2 2.5 3 3.5
Position (m)
240
200
150
100 100
120
140
160
180
200
220
0 2 4 6 8
Ela
stic
Mod
ulus
(G
Pa)
Position (μm) Position (μm)
•-U phase modulus is 210 GPa•-ZrH 1.6 phase modulus is 125 GPa
•The elastic modulus of ThZr2H7-x phase is determined 172GPa for the first time12
Modified AFMBackscattering
SEM
Compatibility with Zircaloy cladding• Zircaloy samples pressed against hydride
fuel at various contact pressures immersed in liquid metal at 375 ºC for one month
1 mm
20 μm
Zircaloy
LMHydride
Zircaloy
LM
Hydride
1 mm
Zircaloy
LM
Hydride
Stainless Steel
200 μm
Zircaloy
Hydride
20 µ
m g
ap
120
MPa
con
tact
TERRANI, K., et al., “Liquid Metal as a Gap Filler to Protect Zircaloy Cladding from Hydride Fuel,” Proceedings of Top Fuel 2009, Paris, France
τ
F
Torque Screw
Stainless SteelZircaloy
Pt Wires Hydride Fuel
-
Irradiation Experiment
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– On-line temperature read outs– on-line fission-gas monitoring
~12 cm
fuel
LM
clad
Pellet FabricationTR
IGA
Fuel
Sl
ugDi
amon
d Co
re D
rills
Roug
h Pe
llets
Po
st D
rillin
g
U(30wt%)-ZrH1.6
19.7% U-235
Grinding Wheel
Rubber Wheel
Adjustable Arm
Pellet Position
Stepper Motor
Smoo
th P
elle
ts
Post
Grin
ding
Cent
erle
ss G
rindi
ng
Appa
ratu
s
Mini Fuel element assembly
Thermal conductivity measured by:
•Two thermocouples; fuel centerline and rod surface
Sheath TC Welded to SS Flange
SS304 CF Mini Flange
Zr CF Mini Flange
He PlenumSS302 Spring
Pb-Bi Alloy
Alumina Spacer
Zircaloy-2 Tube
U0.17ZrH1.6 Fuel
Zircaloy-2 End Cap1 cm
Neutron radiography at MITR with resolution of ~ 100 μm
Active Fuel RegionAlumina Spacers
Zirconium FlangeSS 304 Flange
302 SS Spring
Rod 1
Rod 2
Rod 3
Rod 4
Ti
Cover gas line
Center TC
Pellet
LM
Capsule 3 (Rod 2
Capsule 2 (Rod 3)FCT-3 and CST-3)
Capsule 1 (Rod 1)FCT-1 and CST-1
Flux profile
Core StructureCore Tank
Typical Fuel Element
Fine-Control Regulating Rod
Shim Blade Absorbers (6)
Shim Blade Flow Relief Holes (6)
Hexagon Strut(No Absorbers)
Coolant Entrance Channels (6)
Fixed Absorber in Radial Strut
A-2
A-1A-3
B-1
B-2
B-3
B-4B-5
B-6
B-7
B-8 B-9C-1
C-2
C-3
C-4
C-5
C-6C-7C-8
C-10
C-11
C-12
C-13 C-14 C-15
10 cm
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)
Test matrix for the HYFI irradiation of fuel rods
Irradiation Position
Irradiation dates
March 20, 2011
May 26, 2011
June 25, 2011
Nov. 11. 2011
Jan 20, 2012
Early 2012
Top Dummy Dummy Fuel rod 2 Fuel rod 2 Fuel rod 2 Fuel rod 4(w /thermal cond. Probe)
Middle Fuel rod 3 Fuel rod 3 Dummy Dummy Dummy Fuel rod 5(W He filled gap)
Bottom Fuel rod 1 Fuel rod 1 Fuel rod 1 Fuel rod 1 Dummy Fuel rod 2
Abandoned
Temperature and thermal power profile of fuel rods 1, 2 and 3 since March 2011
Note: Unpredicted frequent shot downs and ramp ups
Thermal conductivity calculation through annual fuel pellet
Neutronic (MCNP) calculations
For 5 MW:
300
310
320
330
340
350
Power LHR
Fuel Portion
Pow
er [W
/cm
3 ]
200
210
220
230
240
LHR
[W/cm
]
H,
Rr0
T0
Ts TTC
TLMcool
Ti
Oxidized call
Time dependence of the thermal conductivity
0.155
0.16
0.165
0.17
0.175
0.18
k12k34
Date
k did not change with at the beginning of irradiation
Lower capsule thermal conductivity estimate
0
200
400
600
800
1000
1200
0.16 0.165 0.17 0.175 0.18
k (lower cap), W-cm-1T-1
21ΔT
(W)th
P5.70
Lower k
k12
Minimum 0.15856169Maximum 0.1770716Points 14244Mean 0.1701022Median 0.17003919Std 0.0011548277d
Thermal conductivity variation of three fuel rods during irradiation
0.1
0.12
0.14
0.16
0.18
0.2
0 500 1000 1500 2000 2500 3000 3500
Capsule 1 (rod 1)Capsule 2 (rod 3)Capsule 3 (rod 2)
Time (hr)
• 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
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
Fission products release due to possible leak in capsule 1
Burnup estimate
Neutronic (MCNP) calculations for 5MW:
Upper Solid
Upper Annular
Middle Solid
Middle Annular
Lower Solid
Lower Annular
300
310
320
330
340
350
Power LHR
Fuel Portion
Pow
er [W
/cm
3 ]
200
210
220
230
240
LH
R [W
/cm]H,
Comparison of In-pile Thermal Conductivity of U(30wt%)-ZrH1.6 and
UO2
Structural Damage at MIT NRL
On January 24, 2012Damage to bottom spacer and is identified, some missing alignment pins on capsules are All parts except bottom spacer moved to storage locations
On November 2011Increasing fission gas release from Capsule 1 triggers its removal, is replaced with the dummy from wet storage
Capsule 1
Bottom Spacer
It has been decided to terminate the irradiation at this time and send the three capsules to INL for post irradiation analysis
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?
31
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 OM, SEM
Uranium particles SEM, AFM
Void around U particles TEM, AFM
Diametral expansion of fuel and cladding
Micrometer
Second-phase particles in fuel SEM, AFM
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