38 research outputs found
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Expected radiation effects in plutonium immobilization ceramic
The current formulation of the candidate ceramic for plutonium immobilization consists primarily of pyrochlore, with smaller amounts of hafnium-zirconolite, rutile, and brannerite or perovskite. At a plutonium loading of 10.5 weight %, this ceramic would be made metamict (amorphous) by radiation damage resulting from alpha decay in a time much less than 10,000 years, the actual time depending on the repository temperature as a function of time. Based on previous experimental radiation damage work by others, it seems clear that this process would also result in a bulk volume increase (swelling) of about 6% for ceramic that was mechanically unconfined. For the candidate ceramic, which is made by cold pressing and sintering and has porosity amounting to somewhat more than this amount, it seems likely that this swelling would be accommodated by filling in the porosity, if the material were tightly confined mechanically by the waste package. Some ceramics have been observed to undergo microcracking as a result of radiation-induced anisotropic or differential swelling. It is unlikely that the candidate ceramic will microcrack extensively, for three reasons: (1) its phase composition is dominated by a single matrix mineral phase, pyrochlore, which has a cubic crystal structure and is thus not subject to anisotropic swelling; (2) the proportion of minor phases is small, minimizing potential cracking due to differential swelling; and (3) there is some flexibility in sintering process parameters that will allow limitation of the grain size, which can further limit stresses resulting from either cause
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Progress in evaluating the corrosion of candidate HLW container metals in irradiated air-steam mixtures
The Yucca Mountain Site Characterization Project is evaluating Yucca Mountain in Nye County, Nevada, as a site for a potential high-level nuclear waste repository. Lawrence Livermore National Laboratory is concerned with the development and performance modeling of waste packages for the potential repository. Argonne National Laboratory has performed experimental studies in support of the waste package effort. This effort is currently guided by the Waste Package Plan, which calls for a systems engineering approach to waste package development. Part of this approach involves formulating an approved set of selection criteria to choose the materials to be used in fabricating the waste packages. Technical issues related to the performance of metals in the air/water vapor environment expected in the potential Yucca Mountain repository are discussed. Preliminary experiments, focused on the atmospheric corrosion of copper-based materials, are summarized. These experiments were performed over a broad range of conditions: temperatures between 90 and 150{degrees}C; relative humidities of 0, 15, 40, and 100%; and gamma dose rates between 0.01 and 0.3 Mrad/hr. In irradiated moist air, copper-based materials form cooper oxides and nitrate phases depending on the dose rate, humidity and temperature. The rates of general corrosion increase with temperature, humidity, and dose rate. Chemical intermediates formed by radiolysis of moist air have been clearly associated with observed corrosion. No significant corrosion was observed for Alloy 825. 13 refs., 3 tabs
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Helium generation in copper by 14.8-MeV neutrons
High purity copper foils were irradiated with 14.8-MeV neutrons from the rotating target neutron source facility at LLL. The average energy of the neutrons was 14.75 +- 0.1 MeV, and the average fluence was 7.0 x 10 n/ cm. After irradiation each foil was heated to the melting point and the released helium was measured by a mass spectrometer of special design. Isochronal heating was carried out on several samples to establish the type and temperature of maximum release. Calculated cross sections from the literature for the (eta,) and (eta,eta') nuclear reactions were used, and the predicted amount of helium was consistently about 0.5 of that actually measured. Because there is very little data on helium generation in metals irradiated with high energy neutrons, these results are important and will be related to potential CTR materials. (auth
Progress report on the results of testing advanced conceptual design metal barrier materials under relevant environmental conditions for a tuff repository
This report discusses the performance of candidate metallic materials envisioned for fabricating waste package containers for long-term disposal at a possible geological repository at Yucca Mountain, Nevada. Candidate materials include austenitic iron-base to nickel-base alloy (AISI 304L, AISI 316L, and Alloy 825), high-purity copper (CDA 102), and copper-base alloys (CDA 613 and CDA 715). Possible degradation modes affecting these container materials are identified in the context of anticipated environmental conditions at the repository site. Low-temperature oxidation is the dominant degradation mode over most of the time period of concern (minimum of 300 yr to a maximum of 1000 yr after repository closure), but various forms of aqueous corrosion will occur when water infiltrates into the near-package environment. The results of three years of experimental work in different repository-relevant environments are presented. Much of the work was performed in water taken from Well J-13, located near the repository, and some of the experiments included gamma irradiation of the water or vapor environment. The influence of metallurgical effects on the corrosion and oxidation resistance of the material is reviewed; these effects result from container fabrication, welding, and long-term aging at moderately elevated temperatures in the repository. The report indicates the need for mechanisms to understand the physical/chemical reactions that determine the nature and rate of the different degradation modes, and the subsequent need for models based on these mechanisms for projecting the long-term performance of the container from comparatively short-term laboratory data. 91 refs., 17 figs., 16 tabs
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Metallurgical analysis of a 304L stainless steel canister from the Spent Fuel Test - Climax
Results of a metallurgical examination of a type 304L stainless steel canister that had been used to store spent nuclear fuel in an underground granite formation for about three years are reported. No observable corrosion or cracking were found. The results are applied to waste packages in a potential high level nuclear waste repository in tuff. 10 refs., 9 figs., 2 tabs
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DT fusion neutron radiation strengthening of copper and niobium
The initial results of a comparative study of the radiation strengthening and damage structures produced in Cu and Nb by D-T fusion and fission reactor neutrons are described. The radiation strengthening produced by a given fluence of fusion neutrons above about 10n/cm is equal to that produced by a fluence of fission reactor neutrons (E greater than 0.1 MeV) ten times as great. This difference is about twice as large as would be expected if the strengthening scaled with damage energy or dpa. Initial transmission electron microscopy observations of the damage structures in fusion and fission reactor neutron irradiated copper indicate that the same type of primary structural defects, vacancy and interstitial point defect clusters and small dislocation loops with a/3 (111) and a/2 (110) Burgers vectors, are produced in both cases. The difference in the radiation strengthening produced by fusion and fission reactor neutrons in Cu appears to result from a substantially greater rate of accumulation of damage, in the form of point defect clusters, during irradiation with fusion neutrons than during irradiation with fission reactor neutrons plus a significant difference in the size and spatial distributions of the damage clusters. (auth
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Limitations on scientific prediction and how they could affect repository licensing
The best possibility for gaining an understanding of the likely future behavior of a high level nuclear waste disposal system is to use the scientific method. However, the scientific approach has inherent limitations when it comes to making long-term predictions with confidence. This paper examines some of these limiting factors as well as the criteria for admissibility of scientific evidence in the legal arena, and concludes that the prospects are doubtful for successful licensing of a potential repository under the regulations that are now being reconsidered. Suggestions am made for remedying this situation
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Science and licensing: Let`s get off the collision course
Our best possibility for gaining an understanding of the likely future behavior of a high level nuclear waste disposal system is use of the scientific method. However, science has inherent limitations when it comes to making long-term predictions with confidence. This paper examines these limiting factors as well as the criteria for admissibility of scientific evidence in the legal arena, and concludes that the prospects are doubtful for successful licensing of a potential repository under the regulations that were binding until recently. Suggestions are made for remedying this situation
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Radiation chemical effects in experiments to study the reaction of glass in an environment of gamma-irradiated air, groundwater, and tuff
The results of experiments performed by John K. Bates et al. on the reaction of nuclear waste glass with a gamma-irradiated 90{sup 0}C aqueous solution were analyzed using theory developed from past research in radiation chemistry. The aqueous solution they used is similar to what would be expected in a water-saturated environment in a nuclear waste repository in tuff. The purpose of our study was to develop an understanding of the radiation-chemical processes that occurred in the Bates et al. experiments so the results could be applied to the design and performance analysis of a proposed repository in unsaturated tuff in Nevada. For the Bates et al. experiments at the highest dose (269 Mrad), which originally contained about 16 ml of "equilibrated" water taken from Nevada Test Site Well J-13 and 5.4 ml of air, we predicted that water decomposition to H{sub 2} and O{sub 2} would produce a pressure increase of at least 1.0 MPa at 20{sup 0}C. We also predicted that nitrogen fixation from the air would occur, producing an increase of 1.6 x 10{sup -4} M in total fixed nitrogen concentration in solution. In addition, an equimolar production of H{sup +} would occur, which would be buffered by the HCO{sub 3}{sup -} in the water. The fixed nitrogen in solution was predicted to be present as NO{sub 2}{sup -} and NO{sub 3}{sup -} with the ratio influenced by the presence of materials catalytic to the decomposition of H{sub 2}O{sub 2}. We found reasonable agreement between our predictions and the observations of Bates et al., where comparisons were possible. We apply the results to the proposed Nevada repository to the degree possible, given the different expected conditions
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Radiation doses in granite around emplacement holes in the Spent Fuel Test - Climax. Final report
Final comparisons are made between measured and calculated radiation doses around the holes in which the spent fuel was emplaced in the Spent Fuel Test - Climax. Neutron doses were found to be negligible compared with gamma doses. Good agreement was found between the doses predicted by Monte Carlo calculations and those measured by short-exposure thermoluminescence dosimetry. Poor agreement was found between the calculational results and doses measured by exposure of LiF optical-absorption-type dosimeters for long periods, probably because of an inability to accurately correct for fade resulting from elevated temperature exposure over several months. The maximum dose to the rock occurred at the walls of the emplacement holes, and amounted to 1.6 MGy (1.6 x 10{sup 8} rad) in granite for the emplacement period of nearly 3 years. It is recommended that dose evaluations for future high-level nuclear waste storage facilities also be performed by combining calculations and dosimetry. Passive dosimetry techniques, if used, should involve short exposures, so that laboratory calibrations can be performed with duplicate time, temperature, dose rate, and dose parameters. An attractive alternative would be to use active ionization chambers, inserted only periodically. These could be calibrated under appropriate temperature and pressure conditions, and could be read directly. 23 references, 7 figures, 8 tables