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History of prototype high level waste Canister SS-9 while in air and water storage

Description: Canister SS-9 was filled with high-level phosphate ceramic waste material in March 1969. Following 1.2 years water storage at 50/sup 0/C, 3.5 years hot air storage at 400 to 500/sup 0/C, and 10 months water storage at 50/sup 0/C, the canister failed. The canister has three visible cracks, one of which is 13 in. long. It was concluded from metallography that failure was due to stress-assisted intergranular attack enhanced by metal sensitization during the hot air storage period, and a high chloride ion concentration in the canister storage water. Cores were taken from Canister SS-9 and the leach rate of the material in deionized water was determined to be 5.1 x 10/sup -4/ g/cm/sup 2/-day for the first day. Averaged over 90 days, the material leach rate was 3.1 x 10/sup -5/ g/cm/sup 2/-day. Since it was known that the failure occurred sometime between quarterly canister storage water sampling periods, these leach rates were used to calculate an effective waste surface area presented by the canister cracks. Thus, the leach rates for the first day and the average for 90 days represent the extremes that could have occurred. The effective waste surface area contacted via the canister cracks was calculated to be between 1 and 20 m/sup 2/. Based on the calculated effective surface area and the above leach rates of the phosphate ceramic material, 6.1 g of waste were dissolved per day. This is related to the activity of the canister as follows: Curies of /sup 137/Cs in Canister SS-9 (corrected to May 1977) = 1.78 x 10/sup 4/; Curies of /sup 137/Cs leached in one day, based on above leach rates = 5.5 x 10/sup -1/; Percent of total cesium in canister = 3.1 x 10/sup -3/; Curies of /sup 137/Cs leached in one day based on averaged ...
Date: November 1, 1977
Creator: Bradley, D.J.
Partner: UNT Libraries Government Documents Department

Theromdynamics of carbon in nickel-based multicomponent solid solutions

Description: The activity coefficient of carbon in nickel, nickel-titanium, nickel-titanium-chromium, nickel-titanium-molybdenum and nickel-titanium-molybdenum-chromium alloys has been measured at 900, 1100 and 1215/sup 0/C. The results indicate that carbon obeys Henry's Law over the range studied (0 to 2 at. percent). The literature for the nickel-carbon and iron-carbon systems are reviewed and corrected. For the activity of carbon in iron as a function of composition, a new relationship based on re-evaluation of the thermodynamics of the CO/CO/sub 2/ equilibrium is proposed. Calculations using this relationship reproduce the data to within 2.5 percent, but the accuracy of the calibrating standards used by many investigators to analyze for carbon is at best 5 percent. This explains the lack of agreement between the many precise sets of data. The values of the activity coefficient of carbon in the various solid solutions are used to calculate a set of parameters for the Kohler-Kaufman equation. The calculations indicate that binary interaction energies are not sufficient to describe the thermodynamics of carbon in some of the nickel-based solid solutions. The results of previous workers for carbon in nickel-iron alloys are completely described by inclusion of ternary terms in the Kohler-Kaufman equation. Most of the carbon solid solution at high temperatures in nickel and nickel-titantium alloys precipitates from solution on quenching in water. The precipitate is composed of very small particles (greater than 2.5 nm) of elemental carbon. The results of some preliminary thermomigration experiments are discussed and recommendations for further work are presented.
Date: April 1, 1978
Creator: Bradley, D. J.
Partner: UNT Libraries Government Documents Department

Proceedings of the Task 2 workshop Waste Isolation Safety Assessment Program

Description: The reports from the workshop on waste form release rate analysis are presented. The workshop started with overview presentations on the Office of Nuclear Waste Isolation (ONWI), the Waste Isolation Safety Assessment Program (WISAP), WISAP Task 2 (Waste Form Release Rate Analysis), and WISAP Task 4 (Sorption/Desorption Analysis). Technical presentations followed in these areas: leaching studies on spent fuels, leaching studies on high-level waste glass, waste form surface science experiments, radiation effects, and leach modeling. Separate abstracts were prepared for each.
Date: January 1, 1979
Creator: Bradley, D.J.
Partner: UNT Libraries Government Documents Department

Leaching of fully radioactive high-level waste glass

Description: As part of continuing Department of Energy (DOE)-sponsored studies in waste management, the Pacific Northwest Laboratory (PNL) has been conducting the High-Level Waste Immobilization Program. The purpose of this program is to develop and demonstrate technology for incorporating nuclear wastes into final waste forms. The preparation and leach testing of fully radioactive, zinc borosilicate glass, which was prepared from power reactor waste, are described. Leach testing using the International Atomic Energy Association (IAEA) procedure was performed in deionized water for a period of 1.75 years. Leach rates were determined for activation products, fission products, and actinides. These rates ranged from 4 x 10/sup -5/ g of glass/cm/sup 2/-day, based on cesium, to 4 x 10/sup -9/ g of glass/cm/sup 2/-day, based on cerium. Following is the ranking of the release rates of the elements, from highest to lowest: Cs > Sr > Co > Sb > Mn > Pu > Eu > Rh > Cm > Ce. A similar leach test, using the same glass composition but with nonradioactive elements, has recently been completed. The leach rates of Cs and Sr for the nonradioactive glass were found to be in close agreement with those in this study. Slopes calculated from curves of cumulative fractions leached show that radioisotope release begins with a diffusion-type mechanism and changes gradually to a silicate lattice alteration mechanism. Changes in sampling frequency altered the apparent release mechanism when leachant changes were longer than one month. The leach rates were quite constant for samples taken from the top to the bottom of the glass melt, indicating a homogeneous product. Safety assessment studies and modeling programs use leach rates to predict the amount of radioactive material released should the waste be contacted by aqueous solutions. Further tests, focusing on geologic storage conditions and using fully radioactive wastes, ...
Date: September 1, 1978
Creator: Bradley, D.J.
Partner: UNT Libraries Government Documents Department

Overview of contamination from US and Russian nuclear complexes

Description: This paper briefly compares the United States and Russian weapons complexes and provides a perspective on the releases of radioactivity to the environment in both countries. Fortunately, the technologies, data, models, and scientific experience that have been gained over the last 50 years are being shared between the US Department of Energy (DOE) and Ministry of Atomic Energy of the Russian Federation (MINATOM) which constitutes a new environmental partnership between the two countries.
Date: June 1, 1995
Creator: Bradley, D.J.
Partner: UNT Libraries Government Documents Department

Radioactive waste management in the former USSR

Description: Radioactive waste materials--and the methods being used to treat, process, store, transport, and dispose of them--have come under increased scrutiny over last decade, both nationally and internationally. Nuclear waste practices in the former Soviet Union, arguably the world's largest nuclear waste management system, are of obvious interest and may affect practices in other countries. In addition, poor waste management practices are causing increasing technical, political, and economic problems for the Soviet Union, and this will undoubtedly influence future strategies. this report was prepared as part of a continuing effort to gain a better understanding of the radioactive waste management program in the former Soviet Union. the scope of this study covers all publicly known radioactive waste management activities in the former Soviet Union as of April 1992, and is based on a review of a wide variety of literature sources, including documents, meeting presentations, and data base searches of worldwide press releases. The study focuses primarily on nuclear waste management activities in the former Soviet Union, but relevant background information on nuclear reactors is also provided in appendixes.
Date: June 1, 1992
Creator: Bradley, D.J.
Partner: UNT Libraries Government Documents Department

Long-term leaching of irradiated spent fuel

Description: Spent Light Water Reactor (LWR) fuel with burnups of 9, 28 and 54 MWd/kg U were leach tested at 25/sup 0/C in deionized water in a Paige apparatus. No discernible differences in leach rates were observed due to burnup. Additionally, the 28 MWd/kg U fuel was IAEA leach tested in five different leachants using the IAEA method. Deionized water gave the highest leach rates and a calcium chloride solution gave the lowest leach rates. An accelerated leaching period was observed during the Paige leach test of the 54 MWd/kg U spent fuel. Comparison between spent fuel and borosilicate waste glass leach rates was made. In sodium bicarbonate solution the leach rates are near equal and the glass becomes increasingly more durable with CaCl/sub 2/ solution, followed by sodium chloride solution, WIPP B brine and deionized water where the glass is two to three orders of magnitude more leach resistant than spent fuel. 16 figures.
Date: January 1, 1979
Creator: Katayama, Y. B. & Bradley, D. J.
Partner: UNT Libraries Government Documents Department

Definition of the waste package environment for a repository located in salt

Description: The expected environmental conditions for emplaced waste packages in a salt repository are simulated in the materials testing program to evaluate performance. Synthetic brines, based on the analyses of actual brines (both intrusion and inclusion), are used for corrosion and leach testing. Elevated temperatures (to 150/sup 0/C) and radiation fields of up to 10/sup 3/ rad/h are employed as conservative conditions to bracket expected performance and provide data for worst case scenarios. Obtaining a precise definition of the waste package environment in a salt repository and its change with time is closely tied to detailed site characterization of the candidate salt repository horizon. It is expected that field testing can augment some of the materials testing currently under way and can provide increased confidence in the predicted site-specific near-field conditions. 17 references, 5 figures, 1 table.
Date: January 1, 1983
Creator: Clark, D. E. & Bradley, D. J.
Partner: UNT Libraries Government Documents Department

Initial waste package interaction tests: status report

Description: This report describes the results of some initial investigations of the effects of rock media on the release of simulated fission products from a sngle waste form, PNL reference glass 76-68. All tests assemblies contained a minicanister prepared by pouring molten, U-doped 76-68 glass into a 2-cm-dia stanless steel tube closed at one end. The tubes were cut to 2.5 to 7.5 cm in length to expose a flat glass surface rimmed by the canister wall. A cylindrical, whole rock pellet, cut from one of the rock materials used, was placed on the glass surface then both the canister and rock pellet were packed in the same type of rock media ground to about 75 ..mu..m to complete the package. Rock materials used were a quartz monzonite basalt and bedded salt. These packages were run from 4 to 6 weeks in either 125 ml digestion bombs or 850 ml autoclaves capable of direct solution sampling, at either 250 or 150/sup 0/C. Digestion bomb pressures were the vapor pressure of water, 600 psig at 250/sup 0/C, and the autoclaves were pressurized at 2000 psig with an argon overpressure. In general, the solution chemistry of these initial package tests suggests that the rock media is the dominant controlling factor and that rock-water interaction may be similar to that observed in some geothermal areas. In no case was uranium observed in solution above 15 ppB. The observed leach rates of U glass not in contact with potential sinks (rock surfaces and alteration products) have been observed to be considerably higher. Thus the use of leach rates and U concentrations observed from binary leach experiments (waste-form water only) to ascertain long-term environmental consequences appear to be quite conservative compared to actual U release in the waste package experiments. Further evaluation, however, of fission product ...
Date: December 1, 1980
Creator: Shade, J.W. & Bradley, D.J.
Partner: UNT Libraries Government Documents Department

Radioactive contamination of the Arctic Region, Baltic Sea, and the Sea of Japan from activities in the former Soviet Union

Description: Contamination of the Arctic regions of northern Europe and Russia, as well as the Sea of Japan, may become a potential major hazard to the ecosystem of these large areas. Widespread poor radioactive waste management practices from nuclear fuel cycle activities in the former Soviet Union have resulted in direct discharges to this area as well as multiple sources that may continue to release additional radioactivity. Information on the discharges of radioactive materials has become more commonplace in the last year, and a clearer picture is emerging of the scale of the contamination. Radioactivity in the Arctic oceans is now reported to be four times higher than would be derived from fallout from weapons tests. Although the characteristics and extent of the contamination are not well known, it has been stated that the contamination in the Arctic may range from 1 to 3.5 billion curies. As yet, no scientific sampling or measurement program has occurred that can verify the amount or extent of the contamination, or its potential impact on the ecosystem.
Date: September 1, 1992
Creator: Bradley, D.J.
Partner: UNT Libraries Government Documents Department

Radioactive contamination of the Arctic Region, Baltic Sea, and the Sea of Japan from activities in the former Soviet Union

Description: Contamination of the Arctic regions of northern Europe and Russia, as well as the Sea of Japan, may become a potential major hazard to the ecosystem of these large areas. Widespread poor radioactive waste management practices from nuclear fuel cycle activities in the former Soviet Union have resulted in direct discharges to this area as well as multiple sources that may continue to release additional radioactivity. Information on the discharges of radioactive materials has become more commonplace in the last year, and a clearer picture is emerging of the scale of the contamination. Radioactivity in the Arctic oceans is now reported to be four times higher than would be derived from fallout from weapons tests. Although the characteristics and extent of the contamination are not well known, it has been stated that the contamination in the Arctic may range from 1 to 3.5 billion curies. As yet, no scientific sampling or measurement program has occurred that can verify the amount or extent of the contamination, or its potential impact on the ecosystem.
Date: September 1, 1992
Creator: Bradley, D. J.
Partner: UNT Libraries Government Documents Department

RADIOACTIVE WASTE MANAGEMENT IN THE USSR: A REVIEW OF UNCLASSIFIED SOURCES, 1963-1990

Description: The Soviet Union operates a vast and growing radioactive waste management system. Detailed information on this system is rare and a general overall picture only emerges after a review of a great deal of literature. Poor waste management practices and slow implementation of environmental restoration activities have caused a great deal of national concern. The release of information on the cause and extent of an accident involving high-level waste at the Kyshtym production reactor site in 1957, as well as other contamination at the site, serve to highlight past Soviet waste management practices. As a result, the area of waste management is now receiving greater emphasis, and more public disclosures. Little is known about Soviet waste management practices related to uranium mining, conversion, and fuel fabrication processes. However, releases of radioactive material to the environment from uranium mining and milling operations, such as from mill tailings piles, are causing public concern. Official Soviet policy calls for a closed fuel cycle, with reprocessing of power reactor fuel that has been cooled for five years. For power reactors, only VVER-440 reactor fuel has been reprocessed in any significant amount, and a decision on the disposition of RBMK reactor fuel has been postponed indefinitely. Soviet reprocessing efforts are falling behind schedule; thus longer storage times for spent fuel will be required, primarily at multiple reactor stations. Information on reprocessing in the Soviet Union has been severely limited until 1989, when two reprocessing sites were acknowledged by the Soviets. A 400-metric ton (MT) per year reprocessing facility, located at Kyshtym, has been operational since 1949 for reprocessing production reactor fuel. This facility is reported to have been reprocessing VVER-440 and naval reactor fuel since 1978, with about 2000 MT of VVER-440 fuel being reprocessed by July 1989. A second facility, located near Krasnoyarsk and ...
Date: March 1, 1990
Creator: Bradley, D. J. & Schneider, K. J.
Partner: UNT Libraries Government Documents Department

Leaching of actinides and technetium from simulated high-level waste glass

Description: Leach tests were conducted using a modified version of the IAEA procedure to study the behavior of glass waste-solution interactions. Release rates were determined for Tc, U, Np, Pu, Am, Cm, and Si in the following solutions: WIPP B salt brine, NaCl (287 g/l), NaCl (1.76 g/1), CaCl/sub 2/ (1.66 g/l), NaHCO/sub 3/ (2.52 g/l), and deionized water. The leach rates for all elements decreased an order of magnitude from their initial values during the first 20 to 30 days leaching time. The sodium bicarbonate solution produced the highest elemental release rates, while the saturated salt brine and deionized water in general gave the lowest release. Technetium has the highest initial release of all elements studied. The technetium release rates, however, decreased by over four orders of magnitude in 150 days of leaching time. In the prepared glass, technetium was phase separated, concentrating on internal pore surfaces. Neptunium, in all cases except CaCl/sub 2/ solution, shows the highest actinide release rate. In general, curium and uranium have the lowest release rates. The range of actinide release rates is from 10/sup -5/ to 10/sup -8/ g/cm/sup 2//day. 25 figures, 7 tables.
Date: August 1, 1979
Creator: Bradley, D.J.; Harvey, C.O. & Turcotte, R.P.
Partner: UNT Libraries Government Documents Department

Leach test methodology for the Waste/Rock Interactions Technology Program

Description: Experimental leach studies in the WRIT Program have two primary functions. The first is to determine radionuclide release from waste forms in laboratory environments which attempt to simulate repository conditions. The second is to elucidate leach mechanisms which can ultimately be incorporated into nearfield transport models. The tests have been utilized to generate rates of removal of elements from various waste forms and to provide specimens for surface analysis. Correlation between constituents released to the solution and corresponding solid state profiles is invaluable in the development of a leach mechanism. Several tests methods are employed in our studies which simulate various proposed leach incident scenarios. Static tests include low temperature (below 100/sup 0/C) and high temperature (above 100/sup 0/C) hydrothermal tests. These tests reproduce nonflow or low-flow repository conditions and can be used to compare materials and leach solution effects. The dynamic tests include single-pass, continuous-flow(SPCF) and solution-change (IAA)-type tests in which the leach solutions are changed at specific time intervals. These tests simulate repository conditions of higher flow rates and can also be used to compare materials and leach solution effects under dynamic conditions. The modified IAEA test is somewhat simpler to use than the one-pass flow and gives adequate results for comparative purposes. The static leach test models the condition of near-zero flow in a repository and provides information on element readsorption and solubility limits. The SPCF test is used to study the effects of flowing solutions at velocities that may be anticipated for geologic groundwaters within breached repositories. These two testing methods, coupled with the use of autoclaves, constitute the current thrust of WRIT leach testing.
Date: May 1, 1980
Creator: Bradley, D.J.; McVay, G.L. & Coles, D.G.
Partner: UNT Libraries Government Documents Department

Status report on LWR spent fuel IAEA leach tests

Description: Spent light-water-reactor (LWR) fuel with an average burnup of 28,000 MWd/MTU was leach-tested at 25/sup 0/C using a modified version of the International Atomic Energy Agency procedure. Leach rates were determined from tests conducted in five different solutions: deionized water, sodium chloride (NaCl), sodium bicarbonate (NaHCO/sub 3/), calcium chloride (CaCl/sub 2/) and Waste Isolation Pilot Plant B brine solutions. Elemental leach rates are reported based on the release of /sup 90/Sr + /sup 90/Y, /sup 106/Ru, /sup 137/Cs, /sup 144/Ce, /sup 154/Eu, /sup 239 + 240/Pu, /sup 244/Cm and total uranium. After 467 days of cumulative leaching, the elemental leach rates are highest in deionized water. The elemental leach rates uin the different solutions generally decreased from deionized water to the 0.03M NaCl solution to the WIPP B brine solution to the 0.03M NaHCO/sub 3/ solution and was a factor of 20 lower in 0.015M CaCl/sub 2/ solution than in deionized water. The leach rates of spent fuel and borosilicate waste-glass were also compared. In sodium bicarbonate solution, the leach rates of the two waste forms were nearly equal, but the glass was increasingly more resistant than spent fuel in calcium chloride solution, followed by sodium chloride solution, WIPP B brine solution and deionized water. In deionized water the glass, based on the elemental release of plutonium and curium, was 50 to 400 times more leach resistant than spent fuel.
Date: March 1, 1980
Creator: Katayama, Y.B.; Bradley, D.J. & Harvey, C.O.
Partner: UNT Libraries Government Documents Department

Annual report on the Characterization on the high-level waste glasses.

Description: The waste compo itions PW-7c and PW-9 were defined and glass development was completed. Major variations in major oxide concentration would not grossly affect the leach rates of the glass. Impact and strength tests on nonradioactive glass showed that the waste glasses produced slightly less fine particulate than commercial glass. Waste glass had 60% of the strength of the soda-lime glass. A water-quench reduced thermal conductivity about 20%, and a 24-h hold at devitrification temperatures did not produce a significant change. Densities of waste glass at process temperature were 6.6 to 9.3% lower than at room temperature. The effects of glass composition on volatility were measured. Leach tests of highly devitrified samples of 72-68 have shown that leach rates of Cs, Sr and U are increased up to 10X and that Zn leach rates are reduced by nearly 200X. In glass 76-68, where devitrification is much slower, elemental differencesbetween as-formed and thermally-treated samples have not been significant. Average Cs leach rates from the 76-68 glass in an IAEA type long-term test have decreased to 3.3 x 10/sup -8/ g/cm/sup 2//day. High temperature (250 and 350/sup 0/C) leach tests showed that glass is comparable to other ceramic materials. In salt brine the glass is rapidly depleted of Cs, Rb and Mo; in water the glass structure is rapidly rearranged to a crystalline structure, and Cs and Rb tend to remain bound in the solid. 76-68 glass (low ZnO) has slow devitrification kinetics compared to 72-68 glass (high ZnO). After equivalent radiation exposures of 300,000 years, the glass buttons still retain their original physical appearance. Stored energy is not a problem for HLW glasses. Density changes are small and do not affect the integrity of the samples. (DLC)
Date: June 1, 1978
Creator: Ross, W.A.; Bradley, D.J. & Bunnell, L.R.
Partner: UNT Libraries Government Documents Department

Waste/Rock Interactions Technology Program. Status report on LWR spent-fuel leach tests

Description: Spent fuels with burnups of 9000, 28,000 and 54,500 MWd/MTU have been leach tested at 25/sup 0/C. Three leach-test procedures (Paige, IAEA and static) were used. IAEA and static tests were conducted in five different solutions: deionized water, sodium bicarbonate, sodium chloride, calcium chloride and Waste Isolation Pilot Plant B brine solutions. Elemental leach data are reported based on the release of /sup 90/Sr/sup +90/Y, /sup 106/Ru, /sup 137/Cs, /sup 144/Ce, /sup 154/Eu, /sup 239 +240/Pu, /sup 125/Sb, /sup 244/Cm, /sup 129/I, /sup 99/Tc, and total uranium. This is the first report on /sup 129/I and /sup 99/Tc from spent fuel. Termination of the Paige test showed that the plateout (radionuclide adsorption) on the test apparatus had negligible effect on the leach rate of cesium and plutonium, but a major (up to a factor of 50 times) effect on the curium leach rate. Three-hundred additional days of leach testing by the IAEA procedure, from 467 to 769 d, showed a continuation of the leaching trends observed during the first 467 d. Results from the first two static leach test series, 2 and 8 d, gave the /sup 129/I and /sup 99/Tc release numbers.
Date: November 1980
Creator: Katayama, Y. B.; Bradley, D. J. & Harvey, C. O.
Partner: UNT Libraries Government Documents Department

Composition of eta carbide in Hastelloy N after aging 10,000 hr at 815/sup 0/C

Description: The composition of the eta carbide in Hastelloy N containing 0.7 wt percent Si in the alloy approaches M/sub 12/C, rather than M/sub 6/C as indicated in the alloy literature. The silicon content of the eta phase in this case was about 25 at. percent, much higher than has been observed in less highly alloyed material. The data do not permit a definition of the limiting compositions of the phases.
Date: November 1, 1977
Creator: Leitnaker, J.M.; Potter, G.A.; Bradley, D.J.; Franklin, J.C. & Laing, W.R.
Partner: UNT Libraries Government Documents Department

Nuclear waste package materials testing report: basaltic and tuffaceous environments

Description: The disposal of high-level nuclear wastes in underground repositories in the continental United States requires the development of a waste package that will contain radionuclides for a time period commensurate with performance criteria, which may be up to 1000 years. This report addresses materials testing in support of a waste package for a basalt (Hanford, Washington) or a tuff (Nevada Test Site) repository. The materials investigated in this testing effort were: sodium and calcium bentonites and mixtures with sand or basalt as a backfill; iron and titanium-based alloys as structural barriers; and borosilicate waste glass PNL 76-68 as a waste form. The testing also incorporated site-specific rock media and ground waters: Reference Umtanum Entablature-1 basalt and reference basalt ground water, Bullfrog tuff and NTS J-13 well water. The results of the testing are discussed in four major categories: Backfill Materials: emphasizing water migration, radionuclide migration, physical property and long-term stability studies. Structural Barriers: emphasizing uniform corrosion, irradiation-corrosion, and environmental-mechanical testing. Waste Form Release Characteristics: emphasizing ground water, sample surface area/solution volume ratio, and gamma radiolysis effects. Component Compatibility: emphasizing solution/rock, glass/rock, glass/structural barrier, and glass/backfill interaction tests. This area also includes sensitivity testing to determine primary parameters to be studied, and the results of systems tests where more than two waste package components were combined during a single test.
Date: March 1, 1983
Creator: Bradley, D.J.; Coles, D.G.; Hodges, F.N.; McVay, G.L. & Westerman, R.E.
Partner: UNT Libraries Government Documents Department

Thermodynamics of high temperature brines

Description: Osmotic and activity coefficient data and enthalpy and heat capacity data for NaCl solutions at saturation pressure of water from 0 to 300{sup 0}C and to saturation composition have been simultaneously fit to a 30 parameter equation. The data are reproduced by the equation, in most cases, to within experimental error. Calculated values of the osmotic coefficient, the activity of water, the activity of NaCl, and the heat capacity, enthalpy and entropy of the solution are given in Tables in 25{sup 0}C intervals from 0 to 300{sup 0}C and concentrations from 0.25 to 25 wt% NaCl.
Date: April 1, 1979
Creator: Pitzer, K.S.; Bradley, D.J.; Rogers, P.S.Z. & Peiper, J.C.
Partner: UNT Libraries Government Documents Department

Leaching characteristics of actinides from simulated reactor waste glass

Description: Two methods for measuring the leach rates of simulated high level waste glass are compared. One is a modification of the standard IAEA method and the other is a one-pass method in which fresh leachant solution is pumped over the sample at a controlled flow rate and temperature. For times up to 3 days, there is close agreement between results from the two methods at 25.0/sup 0/C. Leach rates from the one-pass method show a correlation with flow rate only on day 1 at 25.0/sup 0/C, whereas they show a correlation with flow rate for all three days at 75.0/sup 0/C. /sup 237/Np rates at 75.0/sup 0/C are greater than those at 25.0/sup 0/C, but /sup 239/Pu rates at 75.0/sup 0/C are less than or equal to those at 25.0/sup 0/C.
Date: January 18, 1979
Creator: Weed, H.C.; Coles, D.G.; Bradley, D.J.; Mensing, R.W. & Schweiger, J.S.
Partner: UNT Libraries Government Documents Department

Leaching characteristics of actinides from simulated reactor waste glass

Description: Two methods for measuring the leach rates of simulated high level waste glass are compared. One is a modification of the standard IAEA method and the other is a one-pass method in which fresh leachant solution is pumped over the sample at a controlled flow rate and temperature. For times up to 3 days, there is close agreement between results from the two methods at 25.0C. Leach rates from the one-pass method show no correlation with flow rate at 25.0C, but at 75.0C leach rate increases with flow rate. Np-237 rates at 75.0C are greater than those at 25.0C, but /sup 239/Pu rates at 75.0C are less than or equal to those at 25.0C. Resorption of /sup 239/Pu associated with (SiO/sub 2/)/sub x/ polymers at high temperature is suggested as a possible cause.
Date: October 24, 1978
Creator: Weed, H.C.; Coles, D.G.; Bradley, D.J.; Mensing, R.W. & Schweiger, J.S.
Partner: UNT Libraries Government Documents Department

Comparison of selected foreign plans and practices for spent fuel and high-level waste management

Description: This report describes the major parameters for management of spent nuclear fuel and high-level radioactive wastes in selected foreign countries as of December 1989 and compares them with those in the United States. The foreign countries included in this study are Belgium, Canada, France, the Federal Republic of Germany, Japan, Sweden, Switzerland, and the United Kingdom. All the countries are planning for disposal of spent fuel and/or high-level wastes in deep geologic repositories. Most countries (except Canada and Sweden) plan to reprocess their spent fuel and vitrify the resultant high-level liquid wastes; in comparison, the US plans direct disposal of spent fuel. The US is planning to use a container for spent fuel as the primary engineered barrier. The US has the most developed repository concept and has one of the earliest scheduled repository startup dates. The repository environment presently being considered in the US is unique, being located in tuff above the water table. The US also has the most prescriptive regulations and performance requirements for the repository system and its components. 135 refs., 8 tabs.
Date: April 1, 1990
Creator: Schneider, K.J.; Mitchell, S.J.; Lakey, L.T.; Johnson, A.B. Jr.; Hazelton, R.F. & Bradley, D.J.
Partner: UNT Libraries Government Documents Department

Leaching characteristics of actinides from simulated reactor waste

Description: Leach rates for /sup 237/Np and /sup 239/Pu are investigated with a single-pass leaching system. The factorial experimental design uses several combinations of solution composition and flow rate; and two temperatures, 25/sup 0/ and 75/sup 0/C. The 25/sup 0/C results are compared with those form a modified IAEA procedure. At 25/sup 0/C, leach rates decrease with time. Agreement between results from the single-pass and modified IAEA methods is fair with WIPP brine leachant, good with NaHCO/sub 3/, and good with distilled H/sub 2/O. Leach rates are approximately independent of flow rates at room temperature, but increase with flow rates at high temperature. Rates for /sup 237/Np increase with temperature, but those fro /sup 239/Pu either decrease or do not change with temperature.
Date: January 1, 1979
Creator: Weed, H.C.; Coles, D.G.; Bradley, D.J.; Mensing, R.W.; Schweiger, J.S. & Rego, J.H.
Partner: UNT Libraries Government Documents Department