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Dissolution characteristics of mixed UO{sub 2} powders in J-13 water under saturated conditions

Description: The Yucca Mountain Project/Spent Fuel program at Argonne National Laboratory is designed to determine radionuclide release rates by exposing high-level waste to repository-relevant groundwater. To gain experience for the tests with spent fuel, a scoping experiment was conducted at room temperature to determine the uranium release rate from an unirradiated UO{sub 2} powder mixture (14.3 wt % enrichment in {sup 235}U) to J-13 water under saturated conditions. Another goal set for the experiment was to develop a method for utilizing isotope dilution techniques to determine whether the dissolution rate of UO{sub 2} matrix is in accordance with an existing kinetic model. Results of these analyses revealed unequal uranium dissolution rates from the enriched and depleted portions of the powder mixture because of undisclosed differences between them. Although the presence of this inhomogeneity has precluded the application of the kinetic model, it also provided an opportunity to elaborate on the utilization of isotope dilution data in recognizing and quantifying such conditions. Detailed listings of uranium release and solution chemistry data are presented. Other problems commonly associated with spent fuel, such as the effectiveness of filtering media, the existence of uranium concentration peaks during early stages of the leach tests, the need for concentration corrections due to water replenishments of sample volumes, and experience derived from isotope dilution data are discussed in the context of the present results. 10 refs., 5 figs., 7 tabs.
Date: March 1, 1991
Creator: Veleckis, E. & Hoh, J.C.
Partner: UNT Libraries Government Documents Department

The release of actinides, cesium, strontium, technetium, and iodine from spent fuel under unsaturated conditions

Description: Drip tests to measure radionuclide release from spent nuclear fuel are being performed at 90{degrees}C at a drip rate of 0.75 mL/3.5 days; the test conditions are designed to simulate the behavior of spent fuel under the unsaturated and oxidizing conditions expected in the potential repository at Yucca Mountain. This paper presents measurements of the actinide, {sup 137}Cs, {sup 90}Sr, {sup 99}Tc, and {sup 129}I contents in the leachates after 581 days of testing at 90{degrees}C. These values provide an estimate of the source term for the long-lived radionuclide release under these test conditions. Comparisons are made between our results and those of other researchers.
Date: December 1995
Creator: Finn, P. A.; Hoh, J. C. & Wolf, S. F.
Partner: UNT Libraries Government Documents Department

Dissolution Characteristics of Mixed UO₂ Powders in J-13 Water Under Saturated Conditions

Description: The Yucca Mountain Project/Spent Fuel program at Argonne National Laboratory is designed to determine radionuclide release rates by exposing high-level waste to repository-relevant groundwater. To gain experience for the tests with spent fuel, a scoping experiment was conducted at room temperature to determine the uranium release rate from an unirradiated uranium dioxide powder mixture (14.3 wt % enrichment in uranium-235) to J-13 water under saturated conditions. Another goal set for the experiment was to develop a method for utilizing isotope dilution techniques to determine whether the dissolution rate of uranium dioxide matrix is in accordance with an existing kinetic model. Results of these analyses revealed unequal uranium dissolution rates from the enriched and depleted portions of the powder mixture because of undisclosed differences between them. Although the presence of this inhomogeneity has precluded the application of the kinetic model, it also provided an opportunity to elaborate on the utilization of isotope dilution data in recognizing and quantifying such conditions. Detailed listings of uranium release and solution chemistry data are presented. Other problems commonly associated with spent fuel, such as the effectiveness of filtering media, the existence of uranium concentration peaks during early stages of the leach tests, the need for concentration corrections due to water replenishments of sample volumes, and experience derived from isotope dilution data are discussed in the context of the present results.
Date: March 1991
Creator: Veleckis, Ewald & Hoh, J. C.
Partner: UNT Libraries Government Documents Department

Spent fuel`s behavior under dynamic drip tests

Description: In the potential repository at Yucca Mountain, failure of the waste package container and the cladding of the spent nuclear fuel would expose the fuel to water under oxidizing conditions. To simulate the release behavior of radionuclides from spent fuel, dynamic drip and vapor tests with spent nuclear fuel have been ongoing for 2.5 years. Rapid alteration of the spent fuel has been noted with concurrent release of radionuclides. Colloidal species containing americium and plutonium have been found in the leachate. This observation suggests that colloidal transport of radionuclides should be included in the performance assessment of a potential repository.
Date: December 1, 1995
Creator: Finn, P.A.; Buck, E.C.; Hoh, J.C. & Bates, J.K.
Partner: UNT Libraries Government Documents Department

Performance of high plutonium-containing glasses for the immobilization of surplus fissile materials

Description: Plutonium from dismantled weapons is being evaluated for geological disposal. While a final waste form has not been chosen, borosilicate glass will be one of the waste forms to be evaluated. The reactivity of the reference blend glass containing the standard amount of Pu ({approximately}0.01 wt %) to be produced by the Defense Waste Processing Facility (DWPF) is compared to that of glasses made from the same nominal frit composition but doped with 2 and 7 wt % Pu, and also equal mole percentages of Gd{sub 2}O{sub 3}. The Gd is added to act as a neutron poison to address criticality concerns. The four different glasses have been reacted using the PCT-B method with a SA/V of 20,000 m{sup {minus}1} and the Argonne Vapor Hydration Test (VHT) method. Both test methods accelerate the reaction of the glass. PCT-B is used to determine the reactivity of the glass by analyzing the solution and reacted test components, while the VHT is used to evaluate the long-term reactivity of the glass and the distribution of Pu to secondary phases that will control the long-term reaction of the glass. The results of the tests with high levels of Pu are compared to those with the nominal levels to be produced in the standard DWPF glass.
Date: July 1, 1995
Creator: Bates, J.K.; Emery, J.W.; Hoh, J.C. & Johnson, T.R.
Partner: UNT Libraries Government Documents Department

Gamma irradiation of nitrate-based salts. [Hitec and Draw Temp. 430 molten salts]

Description: An experiment was devised to determine the radiolytic stability of two commercially available candidate salts - Hitec and Draw Temp 430. The salts were exposed to 0.8 x 10/sup 9/ R of gamma radiation in the /sup 60/Co facility at the Argonne National Laboratory and simultaneously heated to temperatures in excess of 530/sup 0/C. A helium gas stream circulated over the salts was analyzed for decomposition products. It was found that there was no observable thermal or radiolytic decomposition of either salt. Although the exposure was equivalent to only about 1 minute in a controlled thermonuclear reactor, the results were very encouraging and suggest that further experimentation on molten nitrate-based salts is warranted.
Date: March 1, 1980
Creator: Breon, S.R.; Chellew, N.R.; Clemmer, R.G. & Hoh, J.C.
Partner: UNT Libraries Government Documents Department

Glass as a waste form for the immobilization of plutonium

Description: Several alternatives for disposal of surplus plutonium are being considered. One method is incorporating Pu into glass and in this paper we discuss the development and corrosion behavior of an alkali-tin-silicate glass and update results in testing Pu doped Defense Waste Processing Facility (DWPF) reference glasses. The alkali-tin-silicate glass was engineered to accommodate a high Pu loading and to be durable under conditions likely to accelerate glass reaction. The glass dissolves about 7 wt% Pu together with the neutron absorber Gd, and under test conditions expected to accelerate the glass reaction with water, is resistant to corrosion. The Pu and the Gd are released from the glass at nearly the same rate in static corrosion tests in water, and are not segregated into surface alteration phases when the glass is reacted in water vapor. Similar results for the behavior of Pu and Gd are found for the DWPF reference glasses, although the long-term rate of reaction for the reference glasses is more rapid than for the alkali-tin-silicate glass.
Date: December 31, 1995
Creator: Bates, J.K.; Ellison, A.J.G.; Emery, J.W. & Hoh, J.C.
Partner: UNT Libraries Government Documents Department

The release of cesium and the actinides from spent fuel under unsaturated conditions

Description: Tests designed to be similar to the unsaturated and oxidizing conditions expected in the candidate repository at Yucca Mountain are in progress with spent fuel at 90{degree}C. The similarities and the differences in release behavior for {sup 137}Cs during the first 2.6 years and the actinides during the first 1.6 years of testing are presented for tests done with (1) water dripped on the fuel at a rate of 0.075 and 0.75 mL every 3.5 days and (2) in a saturated water vapor environment.
Date: December 31, 1995
Creator: Finn, P.A.; Hoh, J.C.; Wolf, S.F.; Slater, S.A. & Bates, J.K.
Partner: UNT Libraries Government Documents Department

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

Description: 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.
Date: December 31, 1996
Creator: Finn, P.A.; Hoh, J.C. & Wolf, S.F.
Partner: UNT Libraries Government Documents Department

Final report for the Light Water Breeder Reactor proof-of-breeding analytical support project

Description: The technology of breeding /sup 233/U from /sup 232/Th in a light water reactor is being developed and evaluated by the Westinghouse Bettis Atomic Power Laboratory (BAPL) through operation and examination of the Shippingport Light Water Breeder Reactor (LWBR). Bettis is determining the end-of-life (EOL) inventory of fissile uranium in the LWBR core by nondestructive assay of a statistical sample comprising approximately 500 EOL fuel rods. This determination is being made with an irradiated-fuel assay gauge based on neutron interrogation and detection of delayed neutrons from each rod. The EOL fissile inventory will be compared with the beginning-of-life fissile loading of the LWBR to determine the extent of breeding. In support of the BAPL proof-of-breeding (POB) effort, Argonne National Laboratory (ANL) carried out destructive physical, chemical, and radiometric analyses on 17 EOL LWBR fuel rods that were previously assayed with the nondestructive gauge. The ANL work included measurements on the intact rods; shearing of the rods into pre-designated contiguous segments; separate dissolution of each of the more than 150 segments; and analysis of the dissolver solutions to determine each segment's uranium content, uranium isotopic composition, and loading of selected fission products. This report describes the facilities in which this work was carried out, details operations involved in processing each rod, and presents a comprehensive discussion of uncertainties associated with each result of the ANL measurements. Most operations were carried out remotely in shielded cells. Automated equipment and procedures, controlled by a computer system, provided error-free data acquisition and processing, as well as full replication of operations with each rod. Despite difficulties that arose during processing of a few rod segments, the ANL destructive-assay results satisfied the demanding needs of the parent LWBR-POB program.
Date: May 1, 1987
Creator: Graczyk, D.G.; Hoh, J.C.; Martino, F.J.; Nelson, R.E.; Osudar, J. & Levitz, N.M.
Partner: UNT Libraries Government Documents Department

Reactivity of high plutonium-containing glasses for the immobilization of surplus fissile materials

Description: Experiments have been performed on glasses doped with 2 and 7 wt % plutonium to evaluate factors that may be important in the performance of these high-Pu-loaded glasses for repository storage. The high Pu loadings result from the need to dispose of excess Pu from weapons dismantling. The glasses were reacted in water vapor to simulate aging that may occur under unsaturated storage conditions prior to contact with liquid water. They were also reacted with liquid water under standard static leach test conditions. The results were compared with similar tests of a reference glass (202 glass) containing only 0.01 wt % Pu. In vapor hydration testing to date, at 2 wt % loading, the Pu was incorporated into the glass without phase separation, and reaction in water vapor proceeded at a rate comparable with that of the 202 glass. At wt % loading, a Pu phase separated and was not uniformly incorporated into the glass. The vapor reaction of this glass proceeded at a more rapid rate. This phase separation was manifested in the static leach tests, where colloidal phases of Pu-rich material remained suspended in solution, thereby increasing the absolute Pu release when compared to the 202 glass.
Date: June 1, 1995
Creator: Bates, J.K.; Hoh, J.C.; Emery, J.W.; Buck, E.C.; Fortner, J.A.; Wolf, S.F. et al.
Partner: UNT Libraries Government Documents Department

Alteration of spent fuel matrix under unsaturated water conditions

Description: Drip tests which simulate the unsaturated conditions expected in the potential repository at Yucca Mountain are in progress to evaluate the long-term performance of spent fuel. This paper examines the corrosion behavior of the spent fuel matrix under conditions in which water is introduced at a rate of 1.5 mL every 7 days. Our recent results suggest a rapid reaction rate of the spent fuel matrix, the formation of alteration products that are similar to the sequence found in ore deposits in uranium mines, and the presence of colloidal species in the leachate. These results are compared to results from two models developed for a potential repository in an unsaturated zone.
Date: June 1, 1995
Creator: Finn, P.A.; Bates, J.K.; Buck, E.C.; Wronkiewicz, D.J.; Hoh, J.C. & Wolf, S.F.
Partner: UNT Libraries Government Documents Department

Radiation effects in moist-air systems and the influence of radiolytic product formation on nuclear waste glass corrosion

Description: Ionizing radiation may affect the performance of glass in an unsaturated repository site by interacting with air, water vapor, or liquid water to produce a variety of radiolytic products. Tests were conducted to examine the effects of radiolysis under high gas/liquid ratios. Results indicate that nitrate is the predominant radiolytic product produced following both gamma and alpha radiation exposure, with lesser amounts of nitrite and carboxylic acids. The formation of nitrogen acids during exposure to long-lived, alpha-particle-emitting transuranic elements indicates that these acids may play a role in influencing nuclear waste form reactions in a long-term unsaturated disposal scenario. Experiments were also conducted with samples that simulate the composition of Savannah River Plant nuclear waste glasses. Radiolytic product formation in batch tests (340 m{sup {minus}1}, 90 C) resulted in a small increase in the release rates of many glass components, such as alkali and alkaline earth elements, although silicon and uranium release rates were slightly reduced indicating an overall beneficial effect of radiation on waste form stability. The radiolytic acids increased the rate of ion exchange between the glass and the thin film of condensate, resulting in accelerated corrosion rates for the glass. The paragenetic sequence of alteration phases formed on both the irradiated and nonirradiated glass samples reacted in the vapor hydration tests matches closely with those developed during volcanic glass alteration in naturally occurring saline-alkaline lake systems. This correspondence suggests that the high temperatures used in these tests have not changed the underlying glass reaction mechanism relate to that which controls glass reactions under ambient surficial conditions.
Date: July 1, 1997
Creator: Wronkiewicz, D.J.; Bates, J.K.; Buck, E.C.; Hoh, J.C.; Emery, J.W. & Wang, L.M.
Partner: UNT Libraries Government Documents Department

Yucca Mountain Project - Argonne National Laboratory annual progress report, FY 1994

Description: This document reports on the work done by the Nuclear Waste Management Section of the Chemical Technology Division (CMT), Argonne National Laboratory, in the period October 1993-September 1994. Studies have been performed to evaluate the performance of nuclear waste glass and spent fuel samples under unsaturated conditions (low volume water contact) that are likely to exist in the Yucca Mountain environment being considered as a potential site for a high-level waste repository. Tests with simulated waste glasses have been in progress for over eight years and demonstrate that actinides from initially fresh glass surfaces will be released as a result of the spallation of reacted glass layers from the surface, as the small volume of water passes over the waste form. Studies are also underway to evaluate the performance of spent fuel samples and unirradiated UO{sub 2} in projected repository conditions. Tests with UO{sub 2} have been ongoing for nine years and show that the oxidation of UO{sub 2} occurs rapidly, and the resulting paragenetic sequence of secondary phases that form on the sample surface is similar to that observed in natural analogues. The reaction of spent fuel samples under conditions similar to those used with UO{sub 2} have been in progress for nearly two years, and the results suggest that spent fuel follows the same reaction progress as UO{sub 2}. The release of individual fission products and transuranic elements was not congruent, with the release being controlled by the formation of small particles or colloids that are suspended in solution and transported away from the waste form. The reaction progress depends on the composition of the spent fuel samples used and, likely, on the composition of the groundwater that contacts the waste form.
Date: February 1, 1995
Creator: Bates, J.K.; Fortner, J.A.; Finn, P.A.; Wronkiewicz, D.J.; Hoh, J.C.; Emery, J.W. et al.
Partner: UNT Libraries Government Documents Department

Final Report for the Light Water Breeder Reactor Proof-of-Breeding Analytical Support Project

Description: The technology of breeding uranium-233 from thorium-232 in a light water reactor is being developed and evaluated by the Westinghouse Bettis Atomic Power Laboratory (BAPL) through operation and examination of the Shippingport Light Water Breeder Reactor (LWBR). Bettis is determining the end-of-life (EOL) inventory of fissile uranium in the LWBR core by nondestructive assay of a statistical sample comprising approximately 500 EOL fuel rods. This determination is being made with an irradiated-fuel assay gauge based on neutron interrogation and detection of delayed neutrons from each rod. The EOL fissile inventory will be compared with the beginning-of-life fissile loading of the LWBR to determine the extent of breeding. In support of the BAPL proof-of-breeding (POB) effort, Argonne National Laboratory (ANL) carried out destructive physical, chemical, and radiometric analyses on 17 EOL LWBR fuel rods that were previously assayed with the nondestructive gauge. The ANL work included measurements on the intact rods; shearing of the rods into pre-designated contiguous segments; separate dissolution of each of the more than 150 segments; and analysis of the dissolver solutions to determine each segment's uranium content, uranium isotopic composition, and loading of selected fission products. This report describes the facilities in which this work was carried out, details operations involved in processing each rod, and presents a comprehensive discussion of uncertainties associated with each result of the ANL measurements.
Date: May 1987
Creator: Graczyk, D. C.; Hoh, J. C.; Martino, F. J.; Nelson, R. E.; Osudar, John & Levitz, Norman M.
Partner: UNT Libraries Government Documents Department

TRUEX Hot Demonstration

Description: In FY 1987, a program was initiated to demonstrate technology for recovering transuranic (TRU) elements from defense wastes. This hot demonstration was to be carried out with solution from the dissolution of irradiated fuels. This recovery would be accomplished with both PUREX and TRUEX solvent extraction processes. Work planned for this program included preparation of a shielded-cell facility for the receipt and storage of spent fuel from commercial power reactors, dissolution of this fuel, operation of a PUREX process to produce specific feeds for the TRUEX process, operation of a TRUEX process to remove residual actinide elements from PUREX process raffinates, and processing and disposal of waste and product streams. This report documents the work completed in planning and starting up this program. It is meant to serve as a guide for anyone planning similar demonstrations of TRUEX or other solvent extraction processing in a shielded-cell facility.
Date: April 1990
Creator: Chamberlain, D. B.; Leonard, R. A.; Hoh, J. C.; Gay, E. C.; Kalina, D. G. & Vandegrift, G. F.
Partner: UNT Libraries Government Documents Department

Leaching Action of EJ-13 Water on Unirradiated UO₂ Surfaces under Unsaturated Conditions at 90 C : Interim Report

Description: A set of experiments, based on the application of the Unsaturated Test method to the reaction of uranium dioxide with EJ-13 water, has been conducted over a period of 182.5 weeks. One half of the experiments have been terminated, while one half are still ongoing. Solutions that have dripped from uranium dioxide specimens have been analyzed for all experiments, while the reacted uranium dioxide surfaces have been examined for only the terminated experiments. A pulse of uranium release from the uranium dioxide solid, in conjunction with the formation of dehydrated schoepite on the surface of the uranium dioxide, was observed during the 39- to 96-week period. Thereafter, the uranium release decreased and a second set of secondary phases was observed. The latter phases incorporate cations from the EJ-13 water and include boltwoodite, uranophane, sklodowskite, compreignacite, and schoepite. The experiments are being continued to monitor for additional changes in solution composition and secondary phase formation, and have now reached the 319-week period.
Date: July 1991
Creator: Wronkiewicz, D. J.; Bates, John K.; Gerding, Thomas J.; Veleckis, Ewald; Tani, B. & Hoh, J. C.
Partner: UNT Libraries Government Documents Department

Test Plan for Reactions Between Spent Fuel and J-13 Well Water Under Unsaturated Conditions

Description: 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 uranium dioxide 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.
Date: January 1993
Creator: Finn, P. A.; Wronkiewicz, David J.; Hoh, J. C.; Emery, J. W.; Hafenrichter, L. D. & Bates, J. K.
Partner: UNT Libraries Government Documents Department

Elements present in leach solutions from unsaturated spent fuel tests

Description: 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
Date: October 1, 1993
Creator: Finn, P.A.; Bates, J.K.; Hoh, J.C.; Emery, J.W.; Hafenrichter, L.D.; Buck, E.C. et al.
Partner: UNT Libraries Government Documents Department

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

Description: 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.
Date: January 1, 1993
Creator: Finn, P.A.; Wronkiewicz, D.J.; Hoh, J.C.; Emery, J.W.; Hafenrichter, L.D. & Bates, J.K.
Partner: UNT Libraries Government Documents Department

Colloidal products and actinide species in leachate from spent nuclear fuel

Description: Two well-characterized types of spent nuclear fuel (ATM-103 and ATM-106) were subjected to unsaturated leach tests with simulated groundwater at 90{degrees}C. The actinides present in the leachate were determined at the end of two successive periods of {approximately}60 days and after an acid strip done at the end of the second period. Both colloidal and soluble actinide species were detected in the leachates which had pHs ranging from 4 to 7. The uranium phases identified in the colloids were schoepite and soddyite. In addition, the actinide release behavior of the two fuels appeared to be different for both the total amount of material released and the relative amount of each isotope released. This paper will focus on the detection and identification of the colloidal species observed in the leachate that was collected after each of the first two successive testing periods of approximately 60 days each. In addition, preliminary values for the total actinide release for these two periods are reported.
Date: December 31, 1993
Creator: Finn, P. A.; Buck, E. C.; Gong, M.; Hoh, J. C.; Emery, J. W.; Hafenrichter, L. D. et al.
Partner: UNT Libraries Government Documents Department

Development and testing of a glass waste form for the immobilization of plutonium

Description: The United States has declared about 50 metric tons of weapons-grade Pu surplus to national security needs. The President has directed that this Pu be placed in a form that provides a high degree of proliferation resistance in which the surplus Pu is both unattractive and inaccessible for use by others [I]. Three alternatives are being evaluated for the disposal 2048 of this material: (1) use of the Pu as a fuel source for commercial reactors; (2) immobilization, where Pu is fixed in a glass or ceramic matrix that also contains or is surrounded by highly radioactive material; and (3) deep bore hole, where Pu is emplaced at depths of several kilometers. The immobilization alternative is being directed by the staff at Lawrence Livermore National Laboratory (LLNL). The staff at ANL are assisting by developing a glass for the immobilization of Pu and in the corrosion testing of glass and ceramic material prepared both at ANL and at other DOE laboratories. As part of this program, we have developed an ATS glass into which 5-7 wt percent Pu has been dissolved. The ATS glass was engineered to accommodate high Pu loading and to be durable under conditions likely to accelerate glass reactions in the geological environment during long-term storage.
Date: December 31, 1996
Creator: Chamberlain, D.B.; Hanchar, J.M.; Emery, J.W.; Hoh, J.C.; Wolf, S.F.; Finch, R.J. et al.
Partner: UNT Libraries Government Documents Department

YUCCA Mountain Project - Argonne National Laboratory, Annual Progress Report, FY 1997 for activity WP 1221 unsaturated drip condition testing of spent fuel and unsaturated dissolution tests of glass.

Description: This document reports on the work done by the Nuclear Waste Management Section of the Chemical Technology Division of Argonne National Laboratory in the period of October 1996 through September 1997. Studies have been performed to evaluate the behavior of nuclear waste glass and spent fuel samples under the unsaturated conditions (low-volume water contact) that are likely to exist in the Yucca Mountain environment being considered as a potential site for a high-level waste repository. Tests with actinide-doped waste glasses, in progress for over 11 years, indicate that the transuranic element release is dominated by colloids that continuously form and span from the glass surface. The nature of the colloids that form in the glass and spent fuel testing programs is being investigated by dynamic light scattering to determine the size distribution, by autoradiography to determine the chemistry, and by zeta potential to measure the electrical properties of the colloids. Tests with UO{sub 2} have been ongoing for 12 years. They show that the oxidation of UO{sub 2} occurs rapidly, and the resulting paragenetic sequence of secondary phases forming on the sample surface is similar to that observed for uranium found in natural oxidizing environments. The reaction of spent fuel samples in conditions similar to those used with UO{sub 2} have been in progress for over six years, and the results suggest that spent fuel forms many of the same alteration products as UO{sub 2}. With spent fuel, the bulk of the reaction occurs via a through-grain reaction process, although grain boundary attack is sufficient to have reacted all of the grain boundary regions in the samples. New test methods are under development to evaluate the behavior of spent fuel samples with intact cladding: the rate at which alteration and radionuclide release occurs when water penetrates fuel sections and whether ...
Date: September 18, 1998
Creator: Bates, J. K.; Buck, E. C.; Emery, J. W.; Finch, R. J.; Finn, P. A.; Fortner, J. et al.
Partner: UNT Libraries Government Documents Department