Role of microstructure and surface defects on the dissolution kinetics of CeO2, a UO2 fuel analogue.

The release of radionuclides from spent fuel in a geological disposal facility is controlled by the surface mediated dissolution of UO2 in groundwater. In this study we investigate the influence of reactive surface sites on the dissolution of a synthesised CeO2 analogue for UO2 fuel. Dissolution was performed on: CeO2 annealed at high temperature, which eliminated intrinsic surface defects (point defects and dislocations); CeO2-x annealed in inert and reducing atmospheres to induce oxygen vacancy defects; and on crushed CeO2 particles of different size fractions. BET surface area measurements were used as an indicator of reactive surface site concentration. Cerium stoichiometry, determined using X-ray Photoelectron Spectroscopy (XPS) and supported by X-ray Diffraction (XRD) analysis, was used to determine oxygen vacancy concentration. Upon dissolution in nitric acid medium at 90°C, a quantifiable relationship was established between the concentration of high energy surface sites and CeO2 dissolution rate; the greater the proportion of intrinsic defects and oxygen vacancies, the higher the dissolution rate. Dissolution of oxygen vacancy-containing CeO2-x gave rise to rates that were an order of magnitude greater than for CeO2 with fewer oxygen vacancies. While enhanced solubility of Ce3+ influenced the dissolution, it was shown that replacement of vacancy sites by oxygen significantly affected the dissolution mechanism due to changes in the lattice volume and strain upon dissolution and concurrent grain boundary decohesion. These results highlight the significant influence of defect sites and grain boundaries on the dissolution kinetics of UO2 fuel analogues and reduce uncertainty in the long-term performance of spent fuel in geological disposal

Contribution of Energetically Reactive Surface Features to the Dissolution of CeO2 and ThO2 Analogues for Spent Nuclear Fuel Microstructures

In the safety case for the geological disposal of nuclear waste, the release of radioactivity from the repository is controlled by the dissolution of the spent fuel in groundwater. There remain several uncertainties associated with understanding spent fuel dissolution, including the contribution of energetically reactive surface sites to the dissolution rate. In this study, we investigate how surface features influence the dissolution rate of synthetic CeO2 and ThO2, spent nuclear fuel analogues that approximate as closely as possible the microstructure characteristics of fuel-grade UO2 but are not sensitive to changes in oxidation state of the cation. The morphology of grain boundaries (natural features) and surface facets (specimen preparation-induced features) was investigated during dissolution. The effects of surface polishing on dissolution rate were also investigated. We show that preferential dissolution occurs at grain boundaries, resulting in grain boundary decohesion and enhanced dissolution rates. A strong crystallographic control was exerted, with high misorientation angle grain boundaries retreating more rapidly than those with low misorientation angles, which may be due to the accommodation of defects in the grain boundary structure. The data from these simplified analogue systems support the hypothesis that grain boundaries play a role in the so-called “instant release fraction” of spent fuel, and should be carefully considered, in conjunction with other chemical effects, in safety performance assessements for the geological disposal of spent fuel. Surface facets formed during the sample annealing process also exhibited a strong crystallographic control and were found to dissolve rapidly on initial contact with dissolution medium. Defects and strain induced during sample polishing caused an overestimation of the dissolution rate, by up to 3 orders of magnitude

UCRL-89475 PREPRINT Post Emplacement Environment of Waste Packages UCRL--G94 7 POST EMPLACEMENT ENVIRONMENT OF WASTE PACKAGES DE64 C042 8

ABSTRACT Experiments have been conducted as part of the Nevada Nuclear Waste Storage Investigations Project to determine the changes in water chemistry due to reaction of the Topopah Spring tuff with natural groundwater at temperatures up to 150°C. The reaction extent has been investigated as a function of rock-to-water ratio, temperature, reaction time, physical state of the samples, and geographic location of the samples within the tuff unit. Results of these experiments will be used to provide information on the water chemistry to be expected if a high level waste repository were to be constructed in the Topopah Spring tuff

Spent fuel cladding containment credit tests

Preliminary tests are being conducted to evaluate the effectiveness of defected cladding as a barrier to radionuclide release from spent fuel rods stored in a geological repository. The tests are being conducted at the Hanford Engineering Development Laboratory Waste Package Task of the Nevada Nuclear Waste Storage Investigations (NNWSI) tuff repository project. In these tests, spent PWR fuel rod specimens with various artificially induced cladding defects are leach tested in a test matrix which also includes both bare fuel specimens (unclad) and undefected spent fuel rod specimens. Artificial cladding defects are made by laser drilling and sawing to give defect areas in the 10{sup 4} to 10{sup 6} {mu}m{sup 2} range. Periodic samples are taken of the leach solution and fused quartz rods contained in the test vessels. Results for the first 180 days of testing are presented. 5 references, 3 figures, 2 tables

Derivation of a waste package source term for NNWSI from the results of laboratory experiments

Results are performed for the dissolution of Turkey Point pressurized water reactor (PWR) spent fuel in J-13 well water at ambient hot cell temperatures. These results are compared with those previously obtained on Turkey Point fuel in deionized water, on H.B. Robinson PWR fuel in J-13 water, and by other workers using various fuels in dilute bicarbonate groundwaters. A model is presented that represents the conditions under which maximum dissolution of spent fuel could occur in a repository sited at Yucca Mountain, Nevada. Using an experimentally determined upper limit of 5 mg/l for uranium solubility in J-13 water, a fractional release rate of 6.4 x 10{sup -8} per year is obtained by assuming that all water entering the repository carries away the maximum amount of uranium. 14 refs., 3 figs., 3 tabs

Gamma irradiation in a saturated tuff environment

The influence of gamma irradiation on the reaction of actinide doped SRL 165 and PNL 76-68 glasses in a saturated tuff environment has been studied in a series of tests lasting up to 56 days. The reaction, and subsequent actinide release, of both glasses depends on the dynamic interaction between radiolysis effects which cause the solution pH to become more acidic and glass reaction which drives the pH more basic. The use of large gamma irradiation dose rates to accelerate reactions that would occur in an actual repository radiation field may affect this dynamic balance by unduly influencing the mechanism of the glass-water reaction. Comparisons are made between the present results and data obtained by reacting the same or similar glasses using MCC-1 and NNWSI rock cup procedures. 11 references, 3 figures

Spent fuel performance data: An analysis of data relevant to the NNWSI Project

This paper summarizes the physical and chemical properties of spent light water reactor fuel that might influence its performance as a waste form under geologic disposal conditions at Yucca Mountain, Nevada. Results obtained on the dissolution testing of spent fuel conducted by the NNWSI Project are presented and discussed. Work published by other programs, in particular those of Canada and Sweden, are reviewed and compared with the NNWSI testing results. An attempt is made to relate all of the results to a common basis of presentation and to rationalize apparent conflicts between sets of results obtained under different experimental conditions

Petrology and isotope geochemistry of tertiary lavas from the northern flank of the tweed volcano, Southeastern Queensland

The generalized stratigraphic sequence (20-21.8 m.y.) of the northern flank of the Tweed Volcano is: Beechmont Basalt (base)-Rhyolite (composed of two distinct units, the Springbrook and Binna Burra rhyolites)-Hobwee Basalt. In addition, comendite occurs as a postrhyolite intrusive phase. Chemically and mineralogically, the 'basalts' are tholeiitic andesites, which are conveniently divided into olivine-normative and quartz-normative types. Phenocryst mineralogy is olivine and labradorite (microphenocrystic) in the olivine-normative lavas, and plagioclase plus rare augite in the quartz-normative lavas.Rhyolites (which constitute some 7 vol. per cent of the Tweed Shield volume) are of the potassic two-feldspar type; these are characterized by highly fractionated trace element patterns, which are most extreme in the Binna Burra rhyolites. The latter, for example, have low K/Rb

Leaching Savannah River Plant nuclear waste glass in a saturated tuff environment

Samples of SRP glass containing either simulated or actual radioactive waste were leached at 90{sup 0}C under conditions simulating a saturated tuff repository environment. The leach vessels were fabricated of tuff and actual tuff groundwater was used. Thus, the glass was leached only in the presence of those materials (including the Type 304L stainless steel canister material) that would be in the actual repository. Tests were performed for time periods up t 6 months at a SA/V ratio of 100 m{sup -1}. Results with glass containing simulated waste indicated that stainless steel canister material around the glass did not significantly affect the leaching. Based on Li and B (elements not in significant concentrations in the tuff or tuff groundwater), glass containing simulated waste leached identically to glass containing actual radioactive waste. The tuff buffered the pH so that only a slight increase was observed as a result of leaching. Results with glass containing actual radioactive waste indicated that tuff reduced the concentrations of Cs-137, Sr-90, and Pu-238 in the free groundwater in the simulated repository by 10 to 100X. Also, radiolysis of the groundwater by the glass (approximately 1000 rad/h) did not significantly affect the pH in the presence of tuff. Measured normalized mass losses in the presence of tuff for the glass based on Cs-137, Sr-90, and Pu-238 in the free groundwater were extremely low, nominally 0.02, 0.02, and 0.005 g/m{sup 2}, respectively, indicating that the glass-tuff system retained radionuclides well. 9 references, 2 figures, 3 tables