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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

Measuring and Predicting Fission Product Noble Metals in SRS HLW Sludges

Description: The noble metals Ru, Rh, Pd, and Ag were produced in the Savannah River Site (SRS) reactors as products of the fission of U-235. Consequently they are in the High Level Waste (HLW) sludges that are currently being immobilized into a borosilicate glass in the Defense Waste Processing Facility (DWPF). The noble metals are a concern in the DWPF because they catalyze the decomposition of formic acid used in the process to produce the flammable gas hydrogen. As the concentration of these noble metals in the sludge increases, more hydrogen will be produced when this sludge is processed. In the SRS Tank Farm it takes approximately two years to prepare a sludge batch for processing in the DWPF. This length of time is necessary to mix the appropriate sludges, blend them to form a sludge batch and then wash it to enable processing in the DWPF. This means that the exact composition of a sludge batch is not known for {approx}two years. During this time, studies with simulated nonradioactive sludges must be performed to determine the desired DWPF processing parameters for the new sludge batch. Consequently, prediction of the noble metal concentrations is desirable to prepare appropriate simulated sludges for studies of the DWPF process for that sludge batch. These studies give a measure of the amount of hydrogen that will be produced when that sludge batch is processed. This report describes in detail the measurement of these noble metal concentrations in sludges and a way to predict their concentrations from an estimate of the lanthanum concentration in the sludge. Results for two sludges are presented in this report. These are Sludge Batch 3 (SB3) currently being processed by the DWPF and a sample of unwashed sludge from Tank 11 that will be part of Sludge Batch 4. The concentrations ...
Date: April 5, 2005
Creator: Bibler, N
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

THE PRODUCT CONSISTENCY TEST HOW AND WHY IT WAS DEVELOPED

Description: The Product Consistency Test (PCT), American Society for Testing Materials (ASTM) Standard C1285, is currently used world wide for testing glass and glass-ceramic waste forms for high level waste (HLW), low level waste (LLW), and hazardous wastes. Development of the PCT was initiated in 1986 because HLW glass waste forms required extensive characterization before actual production began and required continued characterization during production ({ge}25 years). Non-radioactive startup was in 1994 and radioactive startup was in 1996. The PCT underwent extensive development from 1986-1994 and became an ASTM consensus standard in 1994. During the extensive laboratory testing and inter- and intra-laboratory round robins using non-radioactive and radioactive glasses, the PCT was shown to be very reproducible, to yield reliable results rapidly, to distinguish between glasses of different durability and homogeneity, and to easily be performed in shielded cell facilities with radioactive samples. In 1997, the scope was broadened to include hazardous and mixed (radioactive and hazardous) waste glasses. In 2002, the scope was broadened to include glass-ceramic waste forms which are currently being recommended for second generation nuclear wastes yet to be generated in the nuclear renaissance. Since the PCT has proven useful for glass-ceramics with up to 75% ceramic component and has been used to evaluate Pu ceramic waste forms, the use of this test for other ceramic/mineral waste forms such as geopolymers, hydroceramics, and fluidized bed steam reformer mineralized product is under investigation.
Date: December 15, 2008
Creator: Jantzen, C & Ned Bibler, N
Partner: UNT Libraries Government Documents Department

DETERMINATION OF REPORTABLE RADIONUCLIDES FOR DWPF SLUDGE BATCH 5 (MACROBATCH 6)

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 comply with 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 may meet the greater than 0.01% criterion for Curie content. Specification 1.6 of the WAPS, International Atomic Energy Agency (IAEA) Safeguards Reporting for High Level Waste (HLW), requires that the ratio by weights of the following uranium and plutonium isotopes be reported: U-233, U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, and Pu-242. Therefore, the complete set of reportable radionuclides must also include this set of U and Pu isotopes. The DWPF is receiving radioactive sludge slurry from HLW Tank 40. The radioactive sludge slurry in Tank 40 is a blend of the heel from Tank 40 (Sludge Batch 4 (SB4)), Sludge Batch 5 (SB5) that was transferred to Tank 40 from Tank 51, and H-Canyon Np transfers completed after the start of processing. The blend of sludge in Tank 40 is also referred to as Macrobatch 6 (MB6). This report develops ...
Date: February 4, 2010
Creator: Bannochie, C.; Bibler, N. & Diprete, D.
Partner: UNT Libraries Government Documents Department

CHARACTERIZATION OF DWPF MELTER OFF-GAS QUENCHER AND STEAM ATOMIZED SCRUBBER DEPOSIT SAMPLES

Description: This report summarizes the results from the characterization of deposits from the inlets of the primary off-gas Quencher and Steam Atomized Scrubber (SAS) in the Defense Waste Processing Facility (DWPF), as requested by a technical assistance request. DWPF requested elemental analysis and compound identification to help determine the potential causes for the substance formation. This information will be fed into Savannah River National Laboratory modeling programs to determine if there is a way to decrease the formation of the deposits. The general approach to the characterization of these samples included x-ray diffraction (XRD), scanning electron microscopy (SEM), and chemical analysis. The following conclusions are drawn from the analytical results found in this report: (1) The deposits are not high level waste glass from the DWPF melt pool based on comparison of the compositions of deposits to the composition of a sample of glass taken from the pour stream of the melter during processing of Sludge Batch 3. (2) Chemical composition results suggest that the deposits are probably a combination of sludge and frit particles entrained in the off-gas. (3) Gamma emitters, such as Co-60, Cs-137, Eu-154, Am-241, and Am-243 were detected in both the Quencher and SAS samples with Cs-137 having the highest concentration of the gamma emitters. (4) No evidence existed for accumulation of fissile material (U-233, U-235, and Pu-239) relative to Fe in either deposit. (5) XRD results indicated both samples were primarily amorphorous and contained some crystals of the iron oxides, hematite and magnetite (Fe{sub 2}O{sub 3} and Fe(Fe{sub 2}O{sub 4})), along with sodium nitrate (NaNO{sub 3}). The other main crystalline compound in the SAS deposit was mercurous chloride. The main crystalline compound in the Quencher deposit was a uranium oxide compound. These are all sludge components. (6) SEM analysis of the Quencher deposit revealed crystalline uranium ...
Date: June 6, 2007
Creator: Zeigler, K & Ned Bibler, N
Partner: UNT Libraries Government Documents Department

CHARACTERIZATION OF RADIOACTIVE MACROBATCH 4 GLASS BEING PRODUCED BY THE DWPF AT SAVANNAH RIVER SITE

Description: At the Savannah River Site (SRS) the Defense Waste Processing Facility (DWPF) has been immobilizing SRS's radioactive high-level waste (HLW) sludge into a borosilicate glass for approximately nine years. Currently the DWPF is immobilizing HLW sludge in Macrobatch 4 (MB4). Each macrobatch is nominally five hundred thousand gallons of HLW and produces nominally five hundred stainless steel canisters two feet in diameter ten feet tall filled with the borosilicate glass. This paper presents results of the characterization of a sample of MB4 glass taken directly from the pour stream of the DWPF melter during the filling of the canister S02312. This canister was the 275th canister filled during immobilizing MB4. The purpose of the sample was to confirm that the leachability of the glass in a standard ASTM test was less than the leachability criterion set forth in the Waste Acceptance Product Specifications (WAPS) for vitrified waste forms for permanent disposal in a Federal geologic repository. The sample was sent to the Savannah River National Laboratory (SRNL) for characterization.
Date: January 3, 2006
Creator: BIBLER, N. & BANNOCHIE, C.
Partner: UNT Libraries Government Documents Department

RADIOLYTIC HYDROGEN GENERATION INSAVANNAH RIVER SITE (SRS) HIGH LEVEL WASTETANKS COMPARISON OF SRS AND HANFORDMODELING PREDICTIONS

Description: In the high level waste tanks at the Savannah River Site (SRS), hydrogen is produced continuously by interaction of the radiation in the tank with water in the waste. Consequently, the vapor spaces of the tanks are purged to prevent the accumulation of H{sub 2} and possible formation of a flammable mixture in a tank. Personnel at SRS have developed an empirical model to predict the rate of H{sub 2} formation in a tank. The basis of this model is the prediction of the G value for H{sub 2} production. This G value is the number of H{sub 2} molecules produced per 100 eV of radiolytic energy absorbed by the waste. Based on experimental studies it was found that the G value for H{sub 2} production from beta radiation and from gamma radiation were essentially equal. The G value for H{sub 2} production from alpha radiation was somewhat higher. Thus, the model has two equations, one for beta/gamma radiation and one for alpha radiation. Experimental studies have also indicated that both G values are decreased by the presence of nitrate and nitrite ions in the waste. These are the main scavengers for the precursors of H{sub 2} in the waste; thus the equations that were developed predict G values for hydrogen production as a function of the concentrations of these two ions in waste. Knowing the beta/gamma and alpha heat loads in the waste allows one to predict the total generation rate for hydrogen in a tank. With this prediction a ventilation rate can be established for each tank to ensure that a flammable mixture is not formed in the vapor space in a tank. Recently personnel at Hanford have developed a slightly different model for predicting hydrogen G values. Their model includes the same precursor for H{sub 2} as ...
Date: April 15, 2009
Creator: Crawford, C & Ned Bibler, N
Partner: UNT Libraries Government Documents Department

RESULTS FOR THE FIRST QUARTER 2009 TANK 50 WAC SLURRY SAMPLE: CHEMICAL AND RADIONUCLIDE CONTAMINANT RESULTS

Description: This report details the chemical and radionuclide contaminant results for the characterization of the 2009 First Quarter sampling of Tank 50 for the Saltstone Waste Acceptance Criteria (WAC). Information from this characterization will be used by Liquid Waste Operations (LWO) to support the transfer of low-level aqueous waste from Tank 50 to the Salt Feed Tank in the Saltstone Facility in Z-Area, where the waste will be immobilized. This information is also used to update the Tank 50 Waste Characterization System. The following conclusions are drawn from the analytical results provided in this report: (1) The concentrations of the chemical and radioactive contaminants were all less than their respective WAC Targets or Limits except for Am-242m. (2) The radionuclide Am-242m was not detected; however, its detection limit is above the WAC Target given in Attachment 8.4. The higher detection limit was expected based on current analytical capabilities as stated in the Task Technical and Quality Assurance Plan (TTQAP). (3) The reported detection limit of isopropanol was lower than its WAC Limit for accident analysis but higher than its WAC concentration given in Table 4 for vault flammability. The higher detection limit was expected based on current analytical capabilities and is documented in the Task Technical and Quality Assurance Plan (TTQAP). (4) The reported detection limit for Isopar L is lower than its WAC limit for accident analysis in Appendix 8.1 but higher than its WAC concentration given in Table 3 in reference to vault flammability. The higher detection limit was expected based on current analytical capabilities as stated in the Task Technical and Quality Assurance Plan (TTQAP). (5) Isopar L and Norpar 13 have limited solubility in aqueous solutions making it difficult to obtain consistent and reliable sub-samples. The values reported in this memo are the concentrations in the sub-sample ...
Date: October 6, 2009
Creator: Reigel, M.; Diprete, C. & Bibler, N.
Partner: UNT Libraries Government Documents Department