Abstract Properties of the two-phase hydride U0.3ZrH1.6 pertinent to performance as a nuclear fuel for LWRs are reviewed. Much of the available data come from the Space Nuclear Auxiliary Power (SNAP) program of 4 decades ago and from the more restricted data base prepared for the TRIGA research reactors some 3 decades back. Transport, mechanical, thermal and chemical properties are summarized. A principal difference between oxide and hydride fuels is the high thermal conductivity of the latter. This feature greatly decreases the temperature drop over the fuel during operation, thereby reducing the release of fission gases to the fraction due only to recoil. However, very unusual early swelling due to void formation around the uranium particles has been observed in hydride fuels. Avoidance of this source of swelling limits the maximum fuel temperature to ∼650 °C (the design limit recommended by the fuel developer is 750 °C). To satisfy this temperature limitation, the fuel-cladding gap needs to be bonded with a liquid metal instead of helium. Because the former has a thermal conductivity ∼100 times larger than the latter, there is no restriction on gap thickness as there is in helium-bonded fuel rods. This opens the possibility of initial gap sizes large enough to significantly delay the onset of pellet-cladding mechanical interaction (PCMI). The large fission-product swelling rate of hydride fuel (3× that of oxide fuel) requires an initial radial fuel-cladding gap of ∼300 m if PCMI is to be avoided. The liquid-metal bond permits operation of the fuel at current LWR linear-heat-generation rates without exceeding any design constraint. The behavior of hydrogen in the fuel is the source of phenomena during operation that are absent in oxide fuels. Because of the large heat of transport (thermal diffusivity) of H in ZrHx, redistribution of hydrogen in the temperature gradient in the fuel pellet changes the initial H/Zr ratio of 1.6 to ∼1.45 at the center and ∼1.70 at the periphery. Because the density of the hydride decreases with increasing H/Zr ratio, the result of H redistribution is to subject the interior of the pellet to a tensile stress while the outside of the pellet is placed in compression. The resulting stress at the pellet periphery is sufficient to overcome the tensile stress due to thermal expansion in the temperature gradient and to prevent radial cracking that is a characteristic of oxide fuel. Several mechanisms for reduction of the H/Zr ratio during irradiation are identified. The first is transfer of impurity oxygen in the fuel from Zr to rare-earth oxide fission products. The second is the formation of metal hydrides by these same fission products. The third is by loss to the plenum as H2. The review of the fabrication method for the hydride fuel suggests that its production, even on a large scale, may be significantly higher than the cost of oxide fuel fabrication.
[1]
R. H. Price,et al.
The Diffusion Coefficients for Hydrogen in β‐Zirconium
,
1963
.
[2]
M. Yamawaki,et al.
Estimation of Hydrogen Redistribution in Zirconium Hydride under Temperature Gradient
,
2000
.
[3]
S. Yamanaka,et al.
Thermal properties of zirconium hydride
,
2001
.
[4]
C. Huffine.
CHAPTER 13 – Fabrication of Hydrides
,
1968
.
[5]
C. Shuffler,et al.
Thermal hydraulic analysis for grid supported pressurized water reactor cores
,
2009
.
[6]
D. Olander,et al.
Liquid-Metal Bond for LWR Fuel Rods
,
2007
.
[7]
William F. Lyon,et al.
Capabilities of the FALCON Steady State and Transient Fuel Performance Code
,
2004
.
[8]
M. Yamawaki,et al.
Thermodynamic analysis of chemical states of fission products in uranium–zirconium hydride fuel
,
2001
.
[9]
G. Majer,et al.
The mechanism of hydrogen diffusion in zirconium dihydrides
,
1994
.
[10]
R. V. Houten.
Selected engineering and fabrication aspects of nuclear metal hydrides (Li, Ti, Zr, and Y)☆
,
1974
.
[11]
M. Limbäck,et al.
A Model for Analysis of the Effect of Final Annealing on the In- and Out-of-Reactor Creep Behavior of Zircaloy Cladding
,
1996
.
[12]
P. Roncarolo,et al.
Nuclear Reactor Fuel Elements
,
1962
.
[13]
D. Olander,et al.
Hydride fuel behavior in LWRs
,
2005
.
[14]
M. T. Simnad,et al.
The UZrHx alloy: Its properties and use in TRIGA fuel
,
1981
.
[15]
D. Olander,et al.
Thermodynamics of the Zr-H system
,
1995
.
[16]
A. Claire.
Permeation of hydrogen isotopes in structural alloys
,
1984
.
[17]
J. Abriata,et al.
The H-Zr (hydrogen-zirconium) system
,
1990
.
[18]
Tetsushi Matsuda,et al.
Thermal and mechanical properties of zirconium hydride
,
1999
.
[19]
H. Baker,et al.
Alloy phase diagrams
,
1992
.
[20]
Donald R. Olander,et al.
Re-solution of fission gas : A review: Part. I. Intragranular bubbles
,
2006
.