Plastic and nonmetallic article shaping or treating: processes – Shaping or treating radioactive material
Reexamination Certificate
1998-09-25
2001-02-20
Derrington, James (Department: 1731)
Plastic and nonmetallic article shaping or treating: processes
Shaping or treating radioactive material
C264S674000, C252S638000
Reexamination Certificate
active
06190582
ABSTRACT:
FIELD OF THE INVENTION
This invention relates to the sintering process and conditions employed in the production of fissionable nuclear fuel comprising an oxide of uranium containing an additive having a silica constituent.
BACKGROUND OF THE INVENTION
Fissionable nuclear fuel for nuclear reactors typically comprise one of two principal chemical forms. One type consists of fissionable elements such as uranium, plutonium and thorium, and mixtures thereof, in metallic, non-oxide form. Specifically this category comprises uranium, plutonium, etc. metal and mixtures of such metals, namely alloys of such metals.
The other principal type of nuclear reactor fuel consists of ceramic or nonmetallic oxides of fissionable and/or fertile elements comprising uranium, plutonium or thorium, and mixtures thereof. This category of ceramic or oxide fuels is disclosed, for example, in U.S. Pat. No. 4,200,492, issued Apr. 29, 1980, and U.S. Pat. No. 4,372,817, issued Feb. 8, 1983. Uranium oxides, especially uranium dioxide, have become the standard form of fissionable fuel in commercial nuclear power plants used for the generation of electrical power. However, minor amounts of other fissionable materials such as plutonium oxide and thorium oxide, and/or neutron absorbers, sometimes referred to as “poisons”, such as gadolinium oxide, are sometimes admixed with the uranium oxide in the fuel product.
Uranium oxide fuel is generally produced by converting uranium hexafluoride or uranium metal to oxides of uranium. The process includes a series of chemical and physical operations, including pressure compacting uranium oxide in particulate form into handlable pellets or physically integrated bodies of suitable size and configuration, then sintering the resultant pellets or bodies of compacted particles. Sintering at high temperature coalesces the compacted particles of each pellet or body into an integrated unit of high density, and produces other desired effects such as manipulating the molecular oxygen content of the material and removal of residual undesirable impurities, e.g. fluorides.
Sintering processes are amply disclosed in the art, for example U.S. Pat. No. 3,375,306, issued Mar. 26, 1968; U.S. Pat. No. 3,872,022, issued Mar. 18, 1975; U.S. Pat. No. 3,883,623, issued May 13, 1975; U.S. Pat. No. 3,923,933, issued Dec. 2, 1975; U.S. Pat. No. 3,930,787, issued Jan. 6, 1976; U.S. Pat. No. 4,052,330, issued Oct. 4, 1977; and U.S. Pat. No. 4,348,339, issued Sep. 7, 1982.
Fissionable nuclear fuel materials for commercial power generating, water cooled and/or moderated reactors, commonly comprising pellets of uranium oxide, are typically enclosed within a sealed container formed of an alloy of zirconium metal, such as zircaloy -2 (U.S. Pat. No. 2,722,964), or possibly stainless steel, to provide a fuel element. The container, sometimes referred to in the nuclear field as “cladding”, generally comprises a tube-like or elongated enclosure housing fuel pellets stacked therein end-on-end to the extent of about ¾ of the length of the containers.
Fissionable fuel is enclosed and sealed in such containers for service in nuclear reactors to isolate it from contact with the coolant and/or liquid moderator. This precludes either any reaction between the fuel or fission products and the coolant or moderator media, or contamination of the coolant or moderator with escaping radioactive matter from the fuel or fission products.
Experience has shown that after extensive exposure to the radiation in the core of an operating nuclear reactor, typical fuel elements consisting of the fissionable fuel sealed within a metal container are susceptible to failures due to breaching of their containers during or following rapid power increases. Fuel container breaching has been determined to be a result of a combination of conditions, namely, stress imposed upon the metal by thermal expansion of the contained fuel, embrittlement of the metal by prolonged exposure to radiation and stress corrosion cracking susceptibility by the presence of accumulated fission products from the fuel enclosed therein.
Studies of this deleterious phenomenon have determined that three conditions contribute to produce such a failure of the metal fuel container, which is commonly referred to in the art as “intergranular stress corrosion cracking”. First, the metal must be susceptible to stress corrosion cracking in the irradiation environment; second, a level of physical stress must be present; and, third, there must be exposure to aggressive corrosive agents. Metal failure due to stress corrosion cracking can be mitigated or even eliminated by alleviating any one or more of these three conditions.
One effective means for deterring such failures in conventional fuel elements comprising zirconium alloy containers housing uranium oxide fuel has been to include a metallurgically bonded barrier liner of unalloyed zirconium metal over the inner surface of the alloy container substrate. The unalloyed zirconium metal of the barrier liner is more resistant to irradiation embrittlement than the alloy substrate whereby it retains its initial relatively soft and plastic characteristics throughout its service life notwithstanding prolonged exposure to irradiations, etc. Localized physical stresses imposed on such a barrier lined fuel container by heat expanding fuel during rapid power increases are moderated by the plastic movement of the relatively soft unalloyed zirconium metal of the liner. Moreover, the unalloyed zirconium metal has been found to be less susceptible than alloys to the effects of corrosive fission products. That is, the unalloyed zirconium has resistance to the propagation of cracks in the presence of corrosive fission products.
The effectiveness of the unalloyed zirconium barrier liners in resisting the deleterious stress corrosion cracking phenomenon due to the interaction between the fuel pellets and the container in the presence of a corrosive environment of irradiation products, is achieved by mitigating the physical stress and stress corrosion crack propagation susceptibility of the zirconium barrier layer. Effective unalloyed zirconium metal barrier linings for nuclear fuel elements comprising fuel pellets enclosed within a container are disclosed in U.S. Pat. No. 4,200,492 and U.S. Pat. No. 4,372,817.
Another approach to this problem of stress corrosion cracking as a cause of failure of fuel elements when subjected to frequent and drastic power increase has been to modify the physical properties of the uranium oxide fuel with the inclusion of additives. For example, aluminum silicates, derived from clays, when dispersed throughout the uranium oxide in amounts as low as a few tenths of one percent, have been demonstrated to be effective in increasing the plasticity of fuel pellets composed thereof, whereby the thermal expansion induced physical stress attributable to the fuel pellets is reduced. The aluminum silicate may also play a role in reducing the effectiveness and availability of the chemically aggressive fission products which promote stress corrosion cracking of the cladding tubes.
Aluminum silicate additives blended with uranium oxide have been found to be effective in eliminating or mitigating two of the three conditions which must be simultaneously present to produce stress corrosion failures in the metal of a fuel container. An aluminum silicate additive substantially increases the creep rate of fuel pellets comprising oxides of uranium and thereby reduces the stress imposed on the container due to thermal expansion of the fuel material. The enhanced plastic deformation and deformation rates attributable to this additive enables the modified fuel to flow into its own void volume or other free space in the fuel rod within the interior of the fuel container, and thereby reduce the stress applied to the cladding. Thus high localized stresses are mitigated by increased distribution of their forces.
Moreover, the aluminum silicate introduced into the fuel material reacts with fission products produced during irradiation.
Derrington James
General Electric Company
Nixon & Vanderhye
LandOfFree
Controlled atmosphere sintering process for urania... does not yet have a rating. At this time, there are no reviews or comments for this patent.
If you have personal experience with Controlled atmosphere sintering process for urania..., we encourage you to share that experience with our LandOfFree.com community. Your opinion is very important and Controlled atmosphere sintering process for urania... will most certainly appreciate the feedback.
Profile ID: LFUS-PAI-O-2604544