P.A. Finn

A new uranyl oxide hydrate phase derived from spent fuel alteration

An alteration phase that formed during the corrosion of commercial oxide spent nuclear fuel has been characterized with analytical transmission electron microscopy (AEM). The phase is a CsBa uranyl molybdate oxide hydrate that has an orthorhombic structure related to the alkaline earth uranyl oxide hydrates of the protasite-group minerals. On the basis of the compositional analysis and a proposed model of the structure, the ideal structural formula is (Cs0.8Ba0.6)(UO2)5(MoO2)O4(OH)6·nH2O (where n is around 6). Low levels of strontium are also present in the phase. The estimated unit cell parameters are a = 0.754 nm, b = 0.654 nm, and c = 3.008 nm. Although many of the phases formed during corrosion of spent oxide fuel are similar to those observed in natural uraninite deposits, such as Pena Blanca in Mexico, there are important differences owing to the presence of fission products in the spent fuel. Thus, accurate determination of corrosion processes in actual radioactive waste forms is important. This study suggests that the natural UMo deposits at Shelby, WY, and Bates Mountain Tuff, NV, may be good analogues for the long-term behavior of UMo phases formed due to spent fuel corrosion.

The trio experiment

The TRIO experiment is a test of in-situ tritium recovery and heat transfer performance of a miniaturized solid breeder blanket assembly. The assembly (capsule) was monitored for temperature and neutron flux profiles during irradiation and a sweep gas flowed through the capsule to an analytical train wherein the amounts of tritium in its various chemical forms were determined. The capsule was designed to operate at different temperatures and sweep gas conditions. At the end of the experiment the amount of tritium retained in the solid was at a concentration of less than 0.1 wppM. More than 99.9% of tritium generated during the experiment was successfully recovered. The results of the experiment showed that the tritium inventories at the beginning and at the end of the experiment follow a relationship which appears to be characteristic of intragranular diffusion.

Retention of Neptunium in Uranyl Alteration Phases Formed During Spent Fuel Corrosion

Uranyl oxide hydrate phases are known to form during contact of oxide spent nuclear fuel with water under oxidizing conditions; however, less is known about the fate of fission and neutron capture products during this alteration. We describe, the first time, evidence that neptunium can become incorporated into the uranyl secondary phase, dehydrated schoepite (UO{sub 3}{lg_bullet}0.8H{sub 2}O). Based on the long-term durability of natural schoepite, the retention of neptunium in this alteration phase may be significant during spent fuel corrosion in an unsaturated geologic repository.

Corrosion mechanisms of spent fuel under oxidizing conditions

The release of {sup 99}Tc can be used as a reliable marker for the extent of spent oxide fuel reaction under unsaturated high drip rate conditions at 90{degrees}C. Evidence from the leachate data and from scanning and transmission electron microscopy (SEM and TEM) examination of reacted fuel samples is presented for radionuclide release, potential reaction pathways, and the formation of alteration products. In the ATM-103 fuel, 0.03 of the total inventory of {sup 99}Tc is release in 3.7 years under unsaturated and oxidizing conditions. Two reaction pathways that have been identified from SEM are (1) through-grain dissolution with subsequent formation of uranyl alteration products, and (2) grain-boundary dissolution. The major alteration product identified by X-ray diffraction (XRD) and SEM, is Na-boltwoodite, Na[(UO{sub 2})(SiO{sub 3}OH)].H{sub 2}O, which is formed from sodium and silicon in the water leachant.

The TRI0-01 experiment: In-situ tritium recovery results

The TRIO-01 experiment is a test of in-situ tritium recovery from ..gamma..-LiAlO/sub 2/ with test conditions chosen to simulate those anticipated in fusion power reactors. A status report is presented which describes qualitatively the results observed during the irradiation phase of the experiment. Both the rate of tritium release and the chemical forms of tritium were measured using a helium sweep gas which flowed past the breeder material to a gas analysis system.

Compatibility study of solid ceramic breeder materials

Abstract A high temperature (873 K) test of the compatibility of ceramic solid breeders (Li 2 O, γ-LiA10 2 and Li 2 SiO 3 ) with candidate structural materials (Type 316 stainless steel, HT-9, Inconel-625, and Ti-6242) was done for 1900 h. Scale formation was greatest for the Li 2 O/alloy couples; Li 5 FeO 4 was observed in the Li 2 O/HT-9 and Li 2 O/316 SS scales; and LiCrO 2 was observed in most scales.

Spent fuel reaction - the behavior of the {epsilon}-phase over 3.1 years

The release fractions of the five elements in the {epsilon}-phase ({sup 99}Tc, {sup 97}Mo, Ru, Rh, and Pd) as well as that of {sup 238}U are reported for the reaction of two oxide fuels (ATM-103 and ATM-106) in unsaturated tests under oxidizing conditions. The {sup 99}Tc release fractions provide a lower limit for the magnitude of the spent fuel reaction. The {sup 99}Tc release fractions indicate that a surface reaction might be the rate controlling mechanism for fuel reaction under unsaturated conditions and the oxidant is possibly H{sub 2}O{sub 2}, a product of alpha radiolysis of water.

Elements Present in Leach Solutions from Unsaturated Spent Fuel Tests

Preliminary results for the composition of the leachate from unsaturated tests at 90{degrees}C with spent fuel for 55--134 days with J-13 groundwater are reported. The pH of the leachate solutions was found to be acidic, ranging from 4 to 7. The actinide concentrations were 10{sup 5} greater than those reported for saturated spent fuel tests in which the leachate pH was 8. We also found that most species in the leachate were present as colloids containing both americium and curium. The presence of actinides in a form not currently included in repository radionuclide transport models provides information that can be used in spent fuel reaction modeling, the performance assessment of the repository and the design of the engineering barrier system. This report was prepared as part of the Yucca Mountain Site Characterization Project

Equation of state of uranium dioxide

Abstract An equation for the free energy of formation of UO 2 (l) is derived. Using this equation, a new calculational method for estimating consistent values for the vapor pressure of a liquid compound has been applied to uranium dioxide. Equations for the partial pressure of all vapor species are calculated. Values of the liquid density to the critical point are presented.

Test plan for reactions between spent fuel and J-13 well water under unsaturated conditions

The Yucca Mountain Site Characterization Project is evaluating the long-term performance of a high-level nuclear waste form, spent fuel from commercial reactors. Permanent disposal of the spent fuel is possible in a potential repository to be located in the volcanic tuff beds near Yucca Mountain, Nevada. During the post-containment period the spent fuel could be exposed to water condensation since of the cladding is assumed to fail during this time. Spent fuel leach (SFL) tests are designed to simulate and monitor the release of radionuclides from the spent fuel under this condition. This Test Plan addresses the anticipated conditions whereby spent fuel is contacted by small amounts of water that trickle through the spent fuel container. Two complentary test plans are presented, one to examine the reaction of spent fuel and J-13 well water under unsaturated conditions and the second to examine the reaction of unirradiated UO{sub 2} pellets and J-13 well water under unsaturated conditions. The former test plan examines the importance of the water content, the oxygen content as affected by radiolysis, the fuel burnup, fuel surface area, and temperature. The latter test plant examines the effect of the non-presence of Teflon in the test vessel.