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Effects of alpha, gamma, and alpha-recoil radiation on borosilicate glass containing Savannah River Plant defense high-level nuclear waste. [Lead ions-250 keV; xenon ions-160 keV]

Description: At the Savannah River Plant, the reference process for the immobilization of defense high-level waste (DHLW) for geologic storage is vitrification into borosilicate glass. During geologic storage for 10/sup 6/ y, the glass would be exposed to approx. 3 x 10/sup 10/ rad of ..beta.. radiation, approx. 10/sup 10/ rad of ..gamma.. radiation, and 10/sup 18/ particles/g glass for both ..cap alpha.. and ..cap alpha..-recoil radiation. This paper discusses tests of the effect of these radiations on the leachability and density of the glass. Even though the doses were large, no effect of the radiations was detected that reduced the effectiveness of the glass for long-term storage of DHLW even at doses corresponding to 10/sup 6/ years storage for the actual glass. For the tests, glass containing simulated DHLW was prepared from frit of the reference composition. Three methods were used to irradiate the glass: external irradiations with beams of approx. 200 keV Xe or Pb ions, internal irradiations with Cm-244 doped glass, and external irradiations with Co-60 ..gamma.. rays. Results with both Xe and Pb ions indicate that a dose of 3 x 10/sup 13/ ions/cm/sup 2/ (simulating > 10/sup 6/ years storage) does not significantly increase the leachability of the glass in deionized water. Tests with Cm-244 doped glass show no increase in leach rate in water or brine up to a dose of 10/sup 18/ ..cap alpha.. and ..cap alpha..-recoils/g glass. Results of larger doses are being examined. The density of the Cm-244 doped glass has decreased by 1% at a dose of 10/sup 18/ particles/g glass. With ..gamma..-radiation, the density has changed by < 0.05% at a dose of 8.5 x 10/sup 10/ rad. Results of leach tests in deionized water and brine indicated that this very large dose of ..gamma..-radiation increased the leach rate by ...
Date: January 1, 1981
Creator: Bibler, N.E.
Partner: UNT Libraries Government Documents Department

Radiolytic gas production from concrete containing Savannah River Plant waste

Description: To determine the extent of gas production from radiolysis of concrete containing radioactive Savannah River Plant waste, samples of concrete and simulated waste were irradiated by /sup 60/Co gamma rays and /sup 244/Cm alpha particles. Gamma radiolysis simulated radiolysis by beta particles from fission products in the waste. Alpha radiolysis indicated the effect of alpha particles from transuranic isotopes in the waste. With gamma radiolysis, hydrogen was the only significant product; hydrogen reached a steady-state pressure that increased with increasing radiation intensity. Hydrogen was produced faster, and a higher steady-state pressure resulted when an organic set retarder was present. Oxygen that was sealed with the wastes was depleted. Gamma radiolysis also produced nitrous oxide gas when nitrate or nitrite was present in the concrete. With alpha radiolysis, hydrogen and oxygen were produced. Hydrogen did not reach a steady-state pressure at <140 psi. From these results, estimates of pressure in conceptual containers (cylinders 2 feet ID by 10 feet tall, 90% full) of SRP waste concrete were made. During the first 300 years of storage when radiolysis will mainly be from beta-gamma radiation (from /sup 137/Cs and /sup 90/Sr), hydrogen will reach a steady-state pressure of 8 to 28 psi, and oxygen will be partially consumed. These predictions were confirmed by measurement of gas produced over a short time in a container of concrete and actual SRP waste. The tests with simulated waste also indicated that nitrous oxide may form, but because of the low nitrate or nitrite content of the waste, the maximum pressure of nitrous oxide after 300 years will be <60 psi. After decay of these fission products, alpha radiolysis from /sup 238/Pu and /sup 239/Pu will predominate; the hydrogen and oxygen pressures will increase to >200 psi.
Date: January 1, 1978
Creator: Bibler, N.E.
Partner: UNT Libraries Government Documents Department

Radiolytic gas generation in concrete made with incinerator ash containing transuranium nuclides

Description: The effects of various factors on H/sub 2/ generation by alpha radiolysis of concrete containing TRU incinerator ash were studied. Methods for reducing H/sub 2/ generation were investigated. Samples of Portland and high-alumina cement containing up to 30% calcined ash (dry basis) were doped with /sup 238/PuO/sub 2/. Gas pressures were measured as a function of radiation dose; gas compositions were determined. Gas yields were calculated in terms of G values (molecules produced per 100 eV of alpha energy absorbed). These yields were used to estimate pressures in containers of radioactive concrete waste during storage. 4 figures.
Date: January 1, 1979
Creator: Bibler, N.E.
Partner: UNT Libraries Government Documents Department

Characterization of Three Samples Taken from the Off Gas System of DWPF Melter One

Description: The primary objective of this characterization was to determine if there was any evidence for the accumulation of fissile material relative to the element Fe in the deposits. Secondary objectives were to determine their chemical and crystalline compositions and determine what species could be leached from the deposits and appear in condensate water going to the SRS Tank Farm system.
Date: December 17, 2003
Creator: Bibler, N.E.
Partner: UNT Libraries Government Documents Department

Results of Scoping Studies for Determining Radiolytic Hydrogen Production from Moist CST and CST Slurries

Description: In support of the Salt Disposition team, scoping studies have been performed on the radiolysis of moist and aqueous slurries of Crystalline Silicotitanate(CST). If CST is used for removal of Cs-137 from SRS salt solutions, radiolysis of the water by Cs-137 on the CST will produce H2. Also it has been shown that the presence of a solid in the system can enhance the production of H2 by transferring absorbed energy from the solid to the water (1). As indicated in the test plan (2) for this scoping study, it is the intent of this study to determine if CST enhances the radiolytic production of H2 and to estimate the radiolytic hydrogen generation rate from an aqueous CST slurry in a column at the maximum expected Cs-137 loading on the CST.Initially several CST slurry systems were irradiated with Co-60 gamma rays and the radiolytic yield of H2 measured in terms of its G value (molecules of H2 produced per 100 eV of energy absorbed). Based on the results of these tests it was determined that CST did not enhance the radiolytic production of H2 by transferring energy to the water and causing it to decompose.Calculations were then performed to estimate the rate of H2 production from a process column 16ft. long by 5ft. in diameter containing CST that was fully loaded with Cs-137. The maximum rate of H2 production based on the G values measured in this study was one liter per minute at STP (0.04 cfm). This was for a 63 percent water/CST slurry with a G value of 0.2 molecules/100eV for H2 production and a loading of 1 gram of Cs-137 per 100 grams of resin. The present work also indicates that for a column containing salt solution and CST rather than water and CST, the rate would ...
Date: December 11, 1998
Creator: Bibler, N.E.
Partner: UNT Libraries Government Documents Department

Effect of Internal Alpha Radiation on Borosilicate Glass Containing Simulated Radioactive Waste

Description: To evaluate borosilicate glass as a matrix for long-term storage of radioactive waste, samples containing 45 wt. percent simulated waste along with 0.5 wt. percent 244Cm or 1 wt. percent 238Pu as alpha particle emitters were synthesized. A glass containing 238Pu without simulated waste was also made. Effects of internal alpha radiolysis from 244Cm and 238Pu on physical stability, leachability, and dilatation of the glasses were examined. Results confirm that glass may be a desirable matrix for fixing SRP radioactive waste for long-term storage. Internal alpha radiolysis and helium accumulation in the small samples did not significantly damage the glass. Actual values for helium solubility and permeability would be necessary, however, to determine whether helium accumulation might eventually damage larger glass monoliths during long-term storage.
Date: March 14, 2001
Creator: Bibler, N.E.
Partner: UNT Libraries Government Documents Department

Leaching fully radioactive SRP nuclear waste glass in tuff groundwater in stainless steel vessels

Description: SRP glass containing actual radioactive waste was leached in static tests at 90{sup 0}C in a tuffaceous groundwater (J-13 water at pH {similar_to}7.4) at a SA/V ratio of 100m{sup -1} in 316 stainless steel vessels. Tests were performed for time periods up to 134 days. Normalized mass losses were calculated for {sup 137}Cs, {sup 90}Sr, and {sup 238}Pu. The {sup 137}Cs in the leachate appeared to reach a steady value of {similar_to}3 g/m{sup 2}, corresponding to a steady-state concentration of only 1.0 ppB for total cesium. The mass losses based on {sup 90}Sr and {sup 238}Pu appearing in solution were low (<0.3 and <0.01, respectively) because of their low solubilities. However, significant amounts of these radionuclides had deposited on the steel vessel while the amount of deposited {sup 137}Cs was negligible. During the leach tests, the pH changed <0.4 unit and the only significant effect of radiolysis was reduction of NO{sub 3}{sup -} ions in solution to NO{sub 2}{sup -}. When compared to earlier tests, the results confirm that leach rates in the earlier tests with radioactive glass in Teflon vessels were high due to radiolysis of the Teflon. The results also indicate that radioactive and nonradioactive glasses of comparable composition and surface finish leach essentially identically. 12 refs., 3 figs., 4 tabs.
Date: January 1, 1986
Creator: Bibler, N.E.
Partner: UNT Libraries Government Documents Department

Leaching Tc-99 from SRP glass in simulated tuff and salt groundwaters

Description: Results of leach tests with Tc-99 doped SRP borosilicate waste glass are presented. The glass was prepared by melting a mixture of SRP 165 powdered frit doped with a carrier free solution of Tc-99 at 1150{sup 0}C. Dissolution of portions of the resulting glass indicated that the Tc-99 was distributed homogeneously throughout the glass. Static leach tests up to 90 days were performed at 90{sup 0}C in J-13 tuff groundwater or WIPP brine A at a SA/V of 100m{sup -1}. Normalized mass losses were calculated for Tc-99 as well as all the major elements in the glass. Results indicated that under ambient oxidizing conditions Tc-99 leached no faster than the glass-forming elements of the glass. In J-13 water, Tc-99 leached congruently with B. In WIPP brine A, it leached congruently with Si. Leach rates for Li were higher in both groundwaters, probably due to a contribution from an ion exchange mechanism. Leach tests were performed under reducing conditions in J-13 water by adding Zn/Hg amalgam to the leachate. In these tests the pH increased significantly, probably because of the reaction of the amalgam with the water. In a 21-day test, the pH increased to 13 and leach rates for the glass were very high. Even though there was signifcant dissolution of the glass, the normalized mass loss based on Tc-99 was only 0.02g/m{sup 2}. This result and the fact that reducing conditions at normal pH values do not significantly affect the dissolution of the glass, indicate that the low concentrations for Tc-99 obtained under reducing conditions are due to is solubility and not due to an increased durability of the glass. 14 refs., 2 figs., 5 tabs.
Date: January 1, 1987
Creator: Bibler, N E & Jurgensen, A R
Partner: UNT Libraries Government Documents Department

Durabilities and Microstructures of Radioactive Glasses to Immobilize Excess Actinides and Reprocessing Wastes at SRS

Description: This paper presents results of an investigation of the microstructures and durabilities of glasses for immobilization of excess Pu, Am, and Cm, and of the reprocessing wastes at Savannah River Site (SRS). The reprocessing wastes will be vitrified in the Defense Waste Processing Facility (DWPF) at SRS. Another facility at SRS will be used for the Pu, Am, and Cm glasses. In this paper results are presented for a DWPF radioactive glass containing the actual fission product-actinide waste from one the million gallon storage tanks at SRS. This waste is the first radioactive sludge that will be processed in DWPF. The actinide glasses investigated had compositions based on a commercial borosilicate glass composition. All the glasses were so radioactive that they had to be prepared remotely in shielded cells and the analyses had to be performed in gloveboxes or radiobenches. Durabilities were measured using the ASTM C-1285 standard leach test. Results for four glasses are presented. The glasses are a DWPF type glass containing Tank 51 radioactive waste, two glasses containing 15 and 13 wt.percent Pu, respectively, and a glass containing Am and Cm. The radioactive DWPF glass contained 28 wtpercent waste from SRS Tank 51 and was homogeneous. The 15 wt percent Pu contained dissolved PuO2 and as well PuO2 crystals that were not dissolved but were trapped in the glass. The 13 wt percent Pu glass was homogenous. The Am/Cm glass contained <1 wt percent actinides and was homogenous. The PCT test indicated that B, Li, and Na were leaching congruently from the glass. Release rates for Tc-99 and Np-237 were also congruent while Cs-133, Th-232, U-238, and Pu-239 were slower. The two Pu glasses were 25 to 50 times more durable than the DWPF glass. B and Ba were leached congruently while Sm and Pu were lower. ...
Date: December 4, 1995
Creator: Bibler, N.E.
Partner: UNT Libraries Government Documents Department

Determination of Reportable Radionuclides for DWPF Sludge Batch 2 (Macro Batch 3)

Description: The Waste Acceptance Product Specifications (WAPS) 1.2 require that ''The Producer shall report the inventory of radionuclides (in Curies) that have half-lives longer than 10 years and that are, or will be, present in concentrations greater than 0.05 percent of the total inventory for each waste type indexed to the years 2015 and 3115''. As part of the strategy to meet WAPS 1.2, the Defense Waste Processing Facility (DWPF) will report for each waste type, all radionuclides (with half-lives greater than 10 years) that have concentrations greater than 0.01 percent of the total inventory from time of production through the 1100 year period from 2015 through 3115. The initial listing of radionuclides to be included is based on the design-basis glass as identified in the Waste Form Compliance Plan (WCP) and Waste Form Qualification Report (WQR). However, it is required that this list be expanded if other radionuclides with half-lives greater than 10 years are identified that meet the greater than 0.01 percent criterion for Curie content.
Date: December 18, 2002
Creator: Bibler, N.E.
Partner: UNT Libraries Government Documents Department

Role of groundwater oxidation potential and radiolysis on waste glass performance in crystalline repository environments

Description: Laboratory experiments have shown that groundwater conditions in a Stripa granite repository will be as reducing as those in a basalt repository. The final oxidation potential (Eh) at 70/sup 0/C for Stripa groundwater deaerated and equilibrated with crystalline granite was -0.45V. In contrast, the oxidation potential at 60/sup 0/C for Grande Ronde groundwater equilibrated with basalt was -0.40V. The reducing groundwater conditions were found to slightly decrease the time-dependent release of soluble components from the waste glass. Spectrophotometric analysis of the equilibrated groundwaters indicated the presence of Fe/sup 2 +/ confirming that the Fe/sup 2 +//Fe/sup 3 +/ couple is controlling the oxidation potential. It was also shown that in the alkaline pH regime of these groundwaters the iron species are primarily associated with x-ray amorphous precipitates in the groundwater. Gamma radiolysis in the absence of waste glass and in the absence of oxygen further reduces the oxidation potential of both granitic and basaltic groundwaters. The effect is more pronounced in the basaltic groundwater. The mechanism for this decrease is under investigation but appears related to the reactive amorphous precipitate. The results of these tests suggest that H/sub 2/ may not escape from the repository system as postulated and that radiolysis may not cause the groundwaters to become oxidizing in a crystalline repository when abundant Fe/sup 2 +/ species are present. 23 refs., 3 figs., 3 tabs.
Date: January 1, 1985
Creator: Jantzen, C M & Bibler, N E
Partner: UNT Libraries Government Documents Department

Radiation effects on separations materials and processes

Description: This paper briefly summarizes published information on the effects of ionizing radiation on separation processes and materials. Special emphasis is given those processes, solvent extraction, ion exchange, and precipitation, that may have application in removing radioactivity from nuclear waste solutions. The separation and eventual isolation of any radionuclide requires a knowledge of the effect of radiation on the separations process itself and on the materials used in the process. The higher the radiation dose rate, i.e. the more concentrated the radionuclides being processed, the more important is this knowledge. In some cases, such as the separation of intense alpha emitters or the treatment of concentrated solutions of fission products, consideration of the effects of the radiation is a critical factor in the design of the separations materials and in the implementation of the process.
Date: January 1, 1991
Creator: Bibler, N.E.
Partner: UNT Libraries Government Documents Department

Effect of internal alpha radiation on borosilicate glass containing Savannah River Plant waste

Description: Effects of internal alpha radiation on borosilicate glass, a perspective matrix for long-term storage of Savannah River Plant (SRP) radioactive waste, were evaluated in samples containing 45 wt % simulated waste (Fe(OH)/sub 3/--MnO/sub 2/) and either 0.5 wt % /sup 244/Cm or 1 wt % /sup 238/Pu. A glass containing /sup 238/Pu without waste was also studied for comparison. The glasses were examined for changes in physical stability, leachability, and dilatation. Alpha dose rates in the test glasses ranged from 4.5 x 10/sup 14/ to 1.3 x 10/sup 15/ alpha dis/(g-day). After 420 days, microcracks had formed; however, no macrostructural damage to the glasses was observed. Leachabilities for /sup 244/Cm and /sup 238/Pu were <7 x 10/sup -8/ g/(cm/sup 2/-day) and were not affected by the radiation. Continuous leaching by water for 5 days removed <10/sup -5/% of the isotopes. Alpha radiolysis caused expansion of the simulated-waste glasses in proportion to dose. Application of these results to glass containing radioactive Savannah River Plant waste indicated that internal alpha radiolysis will not cause detrimental effects during long-term storage (>10/sup 6/ years) of the waste glass.
Date: May 1, 1978
Creator: Bibler, N.E. & Kelley, J.A.
Partner: UNT Libraries Government Documents Department

Radiation effects on separations materials and processes

Description: This paper briefly summarizes published information on the effects of ionizing radiation on separation processes and materials. Special emphasis is given those processes, solvent extraction, ion exchange, and precipitation, that may have application in removing radioactivity from nuclear waste solutions. The separation and eventual isolation of any radionuclide requires a knowledge of the effect of radiation on the separations process itself and on the materials used in the process. The higher the radiation dose rate, i.e. the more concentrated the radionuclides being processed, the more important is this knowledge. In some cases, such as the separation of intense alpha emitters or the treatment of concentrated solutions of fission products, consideration of the effects of the radiation is a critical factor in the design of the separations materials and in the implementation of the process.
Date: December 31, 1991
Creator: Bibler, N. E.
Partner: UNT Libraries Government Documents Department

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

Description: 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.
Date: November 1984
Creator: Bibler, N. E.; Wicks, G. G. & Oversby, V. M.
Partner: UNT Libraries Government Documents Department

Rheology of Savannah River site tank 42 and tank 51 HLW radioactive sludges

Description: Knowledge of the rheology of the radioactive sludge slurries at the Savannah River Site (SRS) is necessary in order to ensure that they can be retrieved from waste tanks and processed for final disposal. The high activity radioactive wastes stored as caustic slurries at SRS result from the neutralization of acid waste generated from production of nuclear defense materials. During storage, the wastes separate into a supernate layer and a sludge layer. In the Defense Waste Processing Facility (DWPF) at SRS, the radionuclides from the sludge and supernate will be immobilized into borosilicate glass for long term storage and eventual disposal. Before transferring the waste from a storage tank to the DWPF, a portion of the aluminum in the waste sludge will be dissolved and the sludge will be extensively washed to remove sodium. Tank 51 and Tank 42 radioactive sludges represent the first batch of HLW sludge to be processed in the DWPF. This paper presents results of rheology measurements of Tank 51 and Tank 42 at various solids concentrations. The rheologies of Tank 51 and Tank 42 radioactive slurries were measured remotely in the Shielded Cells Operations (SCO) at the Savannah River Technology Center (SRTC) using a modified Haake Rotovisco RV-12 with an M150 measuring drive unit and TI sensor system. Rheological properties of the Tank 51 and Tank 42 radioactive sludges were measured as a function of weight percent solids. The weight percent solids of Tank 42 sludge was 27, as received. Tank 51 sludge had already been washed. The weight percent solids were adjusted by dilution with water or by concentration through drying. At 12, 15, and 18 weight percent solids, the yield stresses of Tank 51 sludge were 5, 11, and 14 dynes/cm2, respectively. The apparent viscosities were 6, 10, and 12 centipoises at 300 ...
Date: January 19, 1996
Creator: Ha, B.C. & Bibler, N.E.
Partner: UNT Libraries Government Documents Department

Rheology of Savannah River Site Tank 42 radioactive sludges. Revision 1

Description: Knowledge of the rheology of the radioactive sludge slurries at the Savannah River Site (SRS) is necessary in order to ensure that they can be retrieved from waste tanks and processed for final disposal. At Savannah River Site (SRS), Tank 42 sludge represents one of the first HLW radioactive sludges to be vitrified in the Defense Waste Processing Facility (DWPF). The rheological properties of unwashed Tank 42 sludge slurries at various solids concentrations were measured remotely in the Shielded Cells at the Savannah River Technology Center (SRTC) using a modified Haake Rotovisco viscometer. Rheological properties of Tank 42 radioactive sludge were measured as a function of weight percent total solids to ensure that the first DWPF radioactive sludge batch can be pumped and processed in the DWPF with the current design bases. The yield stress and consistency of the sludge slurries were determined by assuming a Bingham plastic fluid model.
Date: December 31, 1995
Creator: Ha, B.C. & Bibler, N.E.
Partner: UNT Libraries Government Documents Department

The product consistency test for the DWPF wasteform

Description: The preliminary specifications on the glass wasteform to be produced by the Defense Waste Processing Facility (DWPF) require extensive characterization of the glass product both before actual production begins and then during production. To aid in this characterization, a leach test was needed that was easily reproducible, could be performed remotely on highly radioactive samples, and could yield results rapidly. Several standard leach tests were examined with a variety of test configurations. Using existing tests as a starting point, the DWPF Product Consistency Test (PCT) was developed in which crushed glass samples are exposed to 90{degrees}C deionized water for seven days. Based on extensive testing, including a seven-laboratory round robin and confirmatory testing with radioactive samples, the PCT is very reproducible, yields reliable results rapidly, and can be performed in shielded cell facilities with radioactive samples.
Date: January 1, 1990
Creator: Jantzen, C.M. & Bibler, N.E.
Partner: UNT Libraries Government Documents Department

Assessment of Savannah River borosilicate glass in the repository environment

Description: Since 1973, borosilicate glass has been studied as a matrix for the immobilization of high-level radioactive waste generated at the Savannah River Plant (SRP). In 1977, efforts began to develop and test the large-scale equipment necessary to convert the alkaline waste slurries at SRP into a durable borosilicate glass. A process has now been developed for the proposed Defense Waste Processing Facility (DWPF) which will annually produce approximately 500 canisters of SRP waste glass which will be stored on an interim basis on the Savannah River site. Current national policy calls for the permanent disposal of high-level waste in deep geologic repositories. In the repository environment, SRP waste glass will eventually be exposed to such stresses as lithostatic or hydrostatic pressures, radiation fields, and self-heating due to radioactive decay. In addition, producing and handling each canister of glass will also expose the glass to thermal and mechanical stresses. An important objective of the extensive glass characterization and testing programs of the Savannah River Laboratory (SRL) has been to determine how these stresses affect the performance of SRP waste glass. The results of these programs indicate that: these stresses will not significantly affect the performance of borosilicate glass containing SRP waste; and SRP waste glass will effectively immobilize hazardous radionuclides in the repository environment.
Date: April 1, 1982
Creator: Plodinec, M.J.; Wicks, G.G. & Bibler, N.E.
Partner: UNT Libraries Government Documents Department

Response of the Asiatic clam, Corbicula manilensis, to gamma radiation

Description: When heat exchangers for reactors were plugged by the Asiatic clam, acute gamma radiation was considered as a possible control. Clams were collected and sorted by size; during irradiation the clams were submerged in natural water. Clams of both sizes survived large doses with no radiation damage evident in 30 days. Mortality rose steeply at doses of 2.4 x 10/sup 4/ Rad and above; smaller clams showed a greater resistance than large ones. The feasibility of using periodic exposure to gamma radiation as a means for controlling corbicula infestations is discussed. (HLW)
Date: January 1, 1977
Creator: Tilly, L.J.; Corey, J.C. & Bibler, N.E.
Partner: UNT Libraries Government Documents Department

Near-surface leaching studies of Pb-implanted Savannah River waste glass

Description: The present experiments with Savannah River Plant simulated nuclear waste glass implanted with Pb ions, used Rutherford backscattering spectrometry and elastic recoil detection to follow in detail the changes in composition which occur in the near-surface region upon leaching in deionized water at 90/sup 0/C. Analyses of the leach solutions were made in an attempt to correlate the actual leach rates with the observed near-surface compositional changes. These experiments show that radiation damage can cause changes in the composition of the near-surface of the leached glass. We also find that a critical fluence is reached where abrupt changes of the surface elemental composition occur as a result of leaching. This fluence is near the value observed by both Dran, et al. and Primak. Solution analyses were not made for all the leaching experiments. However, those analyses which were made indicate that the amount of material actually leaving the glass is not significantly increased as a result of the radiation damage.
Date: January 1, 1982
Creator: Arnold, G.W.; Northrup, C.J.M. & Bibler, N.E.
Partner: UNT Libraries Government Documents Department

Effects of temperature and radiation on the nuclear waste glass product consistency leach test

Description: Previous leach studies carried out with monolithic glass samples have shown that glass dissolution rates increase with increasing temperature and may or may not increase on exposure to external gamma radiolysis. In this study we have investigated the effects of temperature (70--1200[degrees]C) and radiation on the dissolution of simulated radioactive waste glasses using the Product Consistency Test (PCT). The PCT is a seven day, crushed glass leach test in deionized water that is carried out at 9OO[degrees]C. To date our results indicate no significant effect of external Co--60 gamma radiation when testing various simulated waste glasses at 90[degrees]C in a wellinsulated compartment within a Gammacell 220 irradiation unit. The temperature dependence for glass dissolution clearly exhibits Arrheniustype behavior for two of the three glasses tested. For the third glass the dissolution decreases at the higher temperatures, probably due to saturation effects. Actual radioactive waste glasses will be investigated later as part of this study.
Date: January 1, 1993
Creator: Crawford, C.L. & Bibler, N.E.
Partner: UNT Libraries Government Documents Department