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Strontium Program: Quarterly Summary Report, January 1, 1960

Description: From Abstract: "This report is one of the a sequence of quarterly reports, each designed to up-date its predecessor beginning with HASL-42, "Environmental Contamination from Weapon Tests." Herein are presented data accrued since HASL-69. Plutonium-239 and cesium-137 levels in human tissue and other biological materials are also presented."
Date: January 1, 1960
Creator: Hardy, Edward P., Jr. & Klein, Stanley
Partner: UNT Libraries Government Documents Department

Analysis of water from K west basin canisters (second campaign)

Description: Gas and liquid samples have been obtained from a selection of the approximately 3,820 spent fuel storage canisters in the K West Basin. The samples were taken to characterize the contents of the gas and water in the canisters. The data will provide source term information for two subprojects of the Spent Nuclear Fuel Project (SNFP) (Fulton 1994): the K Basins Integrated Water Treatment System subproject (Ball 1996) and the K Basins Fuel Retrieval System subproject (Waymire 1996). The barrels of ten canisters were sampled in 1995, and 50 canisters were sampled in a second campaign in 1996. The analysis results for the gas and liquid samples of the first campaign have been reported (Trimble 1995a; Trimble 1995b; Trimble 1996a; Trimble 1996b). An analysis of cesium-137 (137CS ) data from the second campaign samples was reported (Trimble and Welsh 1997), and the gas sample results are documented in Trimble 1997. This report documents the results of all analytes of liquid samples from the second campaign.
Date: March 6, 1997
Creator: Trimble, D.J., Fluor Daniel Hanford
Partner: UNT Libraries Government Documents Department

Barrier Attenuation of Air-Scattered Gamma Radiation

Description: Report of a study that "was conducted to determine the attenuation provided by vertical and horizontal barriers exposed only to skyshine radiation from cobalt-60 and cesium-137 sources. Materials of steel, aluminum, concrete, and wood were used as barriers" (p. 5).
Date: December 1964
Creator: Burson, Z. G. & Summers, R. L.
Partner: UNT Libraries Government Documents Department

Design of a Carousel Process for Removing Cesium from SRS Waste Using Crystalline Silicotitanate Ion Exchanger

Description: Designs of a three-column carousel process based on crystalline silicotitanate (CST) ion exchanger have been developed for removing radioactive 137Cs+ from Savannah River Site's (SRS) nuclear wastes. A multicomponent ion exchange equilibrium model (Zheng et al., 1997) from Texas A&M University, which is based on batch data obtained from CST powder, is used to generate cesium loading data at different cesium concentrations for various types of SRS wastes. These loading data are fit to the Langmuir equation to obtain effective single-component cesium isotherm parameters. The predictions are in reasonable agreement with batch test data obtained from CST powder, an early CST pellet batch (38B), and a later batch (IE911) using two SRS waste simulants. The ratios between experimental cesium distribution coefficients and predicted values are between 0.56 and 1.0. The variation appears to be due to inadequate equilibration time in some of the batches. Mass transfer parameters are estimated by analyzing column data of a simulated SRS waste and Melton Valley Storage Tank W29 (MVST-W29) waste. The intraparticle diffusivity estimated for the two wastes can be well correlated by means of the Stokes-Einstein equation.Simulations are performed to determine the length of the mass transfer zone for given feed compositions, Cs+ concentrations, and linear velocities. In order to ensure high column utilization during both the transient and cyclic steady state periods, the length of a single segment in the carousel process is chosen to be the mass transfer zone length after the concentration wave achieves a constant pattern. Analysis of the dimensionless groups in the differential mass balance equations reveals that the normalized mass transfer zone length is linearly proportional to the particle Peclet number. The proportionality constant is a function of the waste composition and the Cs+ concentration in the waste. The higher the effective Cs+ capacity and the higher ...
Date: January 15, 1999
Creator: Walker, D.D.
Partner: UNT Libraries Government Documents Department

Predicting Sediment and Cesium-137 Transported to Offsite During Extreme Floods

Description: This paper presents the methods and results of a research project for predicting contaminated sediment transport from Oak Ridge Reservation to offside under potential extreme flood conditions. A computer model, Hydrologic Simulation Program--FORTRANE (HSPF), was calibrated and validated for White Oak Creek watershed using a five-year data. The model was then used to quantify the effects of a potential 100-year flood event in terms of the sediment transport and {sup 137}Cs movement. Results from computer simulation showed that during a 100-year flood event the watershed and channel bed became the major sources of the {sup 137}Cs. A 100-year flood event may result in 3.2 Ci of the total annual release of {sup 137}Cs which is six times of the averaged annual release observed during a five-year time period.
Date: April 12, 1999
Creator: Bao, Y.
Partner: UNT Libraries Government Documents Department

Radionuclide contaminated soil: Laboratory study and economic analysis of soil washing. Final report

Description: The objective of the work discussed in this report is to determine if soil washing is a feasible method to remediate contaminated soils from the Hazardous Waste Management Facility (HWMF) at Brookhaven National Laboratory (BNL). The contaminants are predominantly Cs-137 and Sr-90. The authors have assumed that the target activity for Cs-137 is 50 pCi/g and that remediation is required for soils having greater activities. Cs-137 is the limiting contaminant because it is present in much greater quantities than Sr-90. This work was done in three parts, in which they: estimated the volume of contaminated soil as a function of Cs-137 content, determined if simple removal of the fine grained fraction of the soil (the material that is less than 0.063 mm) would effectively reduce the activity of the remaining soil to levels below the 50 pCi/g target, assessed the effectiveness of chemical and mechanical (as well as combinations of the two) methods of soil decontamination. From this analysis the authors were then able to develop a cost estimate for soil washing and for a baseline against which soil washing was compared.
Date: May 20, 1996
Creator: Fuhrmann, M.; Zhou, H.; Patel, B.; Bowerman, B. & Brower, J.
Partner: UNT Libraries Government Documents Department

Direct Grout Stabilization of High Cesium Salt Waste: Salt Alternative Phase III Feasibility Study

Description: The direct grout alternative is a viable option for treatment/stabilization and disposal of salt waste containing Cs-137 concentrations of 1-3 Ci/gal. The composition of the direct grout salt solution is higher in sodium salts and contains up to a few hundred ppm Cs-137 more than the current reference salt solution. However it is still similar to the composition of the current reference salt solution. Consequently, the processing, setting, and leaching properties (including TCLP for Cr and Hg) of the direct grout and current saltstone waste forms are very similar. The significant difference between these waste solutions is that the high cesium salt solution will contain between 1 and 3 Curies of Cs-137 per gallon compared to a negligible amount in the current salt solution. This difference will require special engineering and shielding for a direct grout processing facility and disposal units to achieve acceptable radiation exposure conditions. The Cs-137 concentration in the direct grout salt solution will also affect the long-term curing temperature of the waste form since 4.84 Watts of energy are generated per 1000 Ci of Cs-137. The temperature rise of the direct grout during long-term curing has been calculated by A. Shaddy, SRTC.1 The effect of curing temperature on the strength, leaching and physical durability of the direct grout saltstone is described in this report. At the present time, long term curing at 90 degrees C appears to be unacceptable because of cracking which will affect the structural integrity as evaluated in the immersion test. (The experiments conducted in this feasibility study do not address the effect of cracking on leaching of contaminants other than Cr, Hg, and Cs.) No cracking of the direct grout or reference saltstone waste forms was observed for samples cured at 70 degrees C. At the present time the implications of waste ...
Date: December 7, 1998
Creator: Langton, C.A.
Partner: UNT Libraries Government Documents Department

Cesium-137 in K west basin canister water

Description: Liquid and gas samples were taken from 50 K West Basin fuel storage canisters in 1996. The cesium-137 data from the liquid samples and an analysis of the data are presented. The analysis indicated that the cesium-137 data follow a lognormal distribution. Assuming that the total distribution of the K West canister water was predicted, the total K West Basin canister water was estimated to contain about 8,150 curies. The mean canister contains about 2.14 curies with as many as 5% or 190 of the canisters exceeding 19 curies. Opening ten canisters per shift could include a hot canister (cesium-137 > 25 curies) in one out of eight shifts.
Date: January 24, 1997
Creator: Trimble, D.J.
Partner: UNT Libraries Government Documents Department

Characterization of Hanford K basin fuel and sludge: A second look

Description: Characterization of N Reactor metal spent fuel and associated sludge stored in the two Hanford K Basins has entered a more mature stage. Previous campaigns had consisted of top-view visual surveys of-open fuel canisters, limited collection of gas and liquid from sealed canisters, detailed examinations of only a few-fuel elements and collection of sludge from the floor of only one basin. More recent work has included lifting fuel elements from both Basins to ascertain bottom end and circumferential cracks. Sludge collection has now been performed for material residing inside of spent fuel canisters in both Basins. Finally the number of gas and liquid samples from sealed canisters has been greatly expanded leading to a maximum observed cesium-137 content ten times higher than previous reports. Characterization has been a challenge because of the age of the fuel materials, the water environment, and the radiation field.
Date: January 31, 1997
Creator: Makenas, B.J.
Partner: UNT Libraries Government Documents Department

137Cs(90Sr) and Pu isotopes in the Pacific Ocean sources & trends

Description: The main source of artificial radioactivity in the world`s oceans can be attributed to worldwide fallout from atmospheric nuclear weapons testing. Measurements of selected artificial radionuclides in the Pacific Ocean were first conducted in the 1960`s where it was observed that fallout radioactivity had penetrated the deep ocean. Extensive studies carried out during the 1973-74 GEOSECS provided the first comprehensive data on the lateral and vertical distributions of {sup 9O}Sr, {sup 137}Cs and Pu isotopes in the Pacific on a basin wide scale. Estimates of radionuclide inventories in excess of amounts predicted to be delivered by global fallout alone were attributed to close-in fallout and tropospheric inputs from early U.S. tests conducted on Bikini and Enewetak Atolls in the Equatorial Pacific. In general, levels of fallout radionuclides (including {sup 9O}Sr, {sup 137}Cs and Pu isotopes) in the surface waters of the Pacific Ocean have decreased considerably over the past 4 decades and are now much more homogeneously distributed. Resuspension and the subsequent deposition of fallout radionuclides from previously deposited debris on land has become an important source term for the surface ocean. This can be clearly seen in measurements of fallout radionuclides in mineral aerosols over the Korean Peninsula (Yellow dust events). Radionuclides may also be transported from land to sea in river runoff-these transport mechanisms are more important in the Pacific Ocean where large quantities of river water and suspended sands/fluvial sediments reach the coastal zone. Another unique source of artificial radionuclides in the Pacific Ocean is derived from the slow resolubilization and transport of radionuclides deposited in contaminated lagoon and slope sediments near U.S. and French test sites. Although there is a small but significant flux of artificial radionuclides depositing on the sea floor, > 80% of the total 239, {sup 240}Pu inventory and > 95% of the ...
Date: November 1, 1996
Creator: Hamilton, T.F., Millies-Lacrox, J.C. & Hong, G.H.
Partner: UNT Libraries Government Documents Department

Crystalline silicotitanate gate review analysis

Description: Crystalline silicotitanate (CST) is an ion-exchange method for removing radioactive cesium from tank waste to allow the separation of the waste into high- and low-level fractions. The CST, originally developed Sandia National Laboratories personnel in association with Union Oil Products Corporation, has both a high affinity and selectivity for sorbing cesium-137 from highly alkaline or acidic solutions. For several years now, the U.S. Department of Energy has funded work to investigate applying CST to large-scale removal of cesium-137 from radioactive tank wastes. In January 1997, an expert panel sponsored by the Tanks Focus Area met to review the current state of the technology and to determine whether it was ready for routine use. The review also sought to identify any technical issues that must be resolved or additional CST development that must occur before full implementation by end-users. The CST Gate Review Group concluded that sufficient work has been done to close developmental work on CST and turn the remaining site-specific tasks over to the users. This report documents the review group`s findings, issues, concerns, and recommendations as well as responses from the Tanks Focus Area expert staff to specific pretreatment and immobilization issues.
Date: November 1, 1997
Creator: Schlahta, S.N.; Carreon, R. & Gentilucci, J.A.
Partner: UNT Libraries Government Documents Department

Modeling atmospheric deposition using a stochastic transport model

Description: An advanced stochastic transport model has been modified to include the removal mechanisms of dry and wet deposition. Time-dependent wind and turbulence fields are generated with a prognostic mesoscale numerical model and are used to advect and disperse individually released particles that are each assigned a mass. These particles are subjected to mass reduction in two ways depending on their physical location. Particles near the surface experience a decrease in mass using the concept of a dry deposition velocity, while the mass of particles located within areas of precipitation are depleted using a scavenging coefficient. Two levels of complexity are incorporated into the particle model. The simple case assumes constant values of dry deposition velocity and scavenging coefficient, while the more complex case varies the values according to meteorology, surface conditions, release material, and precipitation intensity. Instantaneous and cumulative dry and wet deposition are determined from the mass loss due to these physical mechanisms. A useful means of validating the model results is with data available from a recent accidental release of Cesium-137 from a steel-processing furnace in Algeciras, Spain in May, 1998. This paper describes the deposition modeling technique, as well as a comparison of simulated concentration and deposition with measurements taken for the Algeciras release.
Date: December 17, 1999
Creator: Buckley, R. L.
Partner: UNT Libraries Government Documents Department

Radioisotope inventory of T101AZ thermocouple tree from riser 13D

Description: The radionuclide inventory for the thermocouple tree removed from tank T101-AZ riser 13D was estimated using measured {sup 137}Cs activity. This activity was measured by detectors as the tree was removed from the tank. Other radionuclide activities were estimated using the results of tank samples.
Date: June 12, 1996
Creator: Kessler, S.F., Westinghouse Hanford
Partner: UNT Libraries Government Documents Department

Electrically switched cesium ion exchange. FY 1997 annual report

Description: This paper describes the Electrically Switched Ion Exchange (ESIX) separation technology being developed as an alternative to ion exchange for removing radionuclides from high-level waste. Progress in FY 1997 for specific applications of ESIX is also outlined. The ESIX technology, which combines ion exchange and electrochemistry, is geared toward producing electroactive films that are highly selective, regenerable, and long lasting. During the process, ion uptake and elution can be controlled directly by modulating the potential of an ion exchange film that has been electrochemically deposited onto a high surface area electrode. This method adds little sodium to the waste stream and minimizes the secondary wastes associated with traditional ion exchange techniques. Development of the ESIX process is well underway for cesium removal using ferrocyanides as the electroactive films. Films having selectivity for perrhenate (a pertechnetate surrogate) over nitrate also have been deposited and tested. Based on the ferrocyanide film capacity, stability, rate of uptake, and selectivity shown during performance testing, it appears possible to retain a consistent rate of removal and elute cesium into the same elution solution over several load/unload cycles. In batch experiments, metal hexacyanoferrate films showed high selectivities for cesium in concentrated sodium solutions. Cesium uptake was unaffected by Na/Cs molar ratios of up to 2 x 10{sup 4} , and reached equilibrium within 18 hours. During engineering design tests using 60 pores per inch, high surface area nickel electrodes, nickel ferrocyanide films displayed continued durability. losing less than 20% of their capacity after 1500 load/unload cycles. Bench-scale flow system studies showed no change in capacity or performance of the ESIX films at a flow rate up to 13 BV/h, the maximum flow rate tested, and breakthrough curves further supported once-through waste processing. 9 refs., 24 figs.
Date: September 1, 1997
Creator: Lilga, M.A.; Orth, R.J. & Sukamto, J.P.H.
Partner: UNT Libraries Government Documents Department

Characterization report for Building 301 Hot Cell Facility

Description: During the period from October, 1997, through March, 1998, ANL-E Health Physics conducted a pre-D and D characterization of Building 301, referred to as the Hot Cell Facility. While primary emphasis was placed on radiological evaluation, the presence of non-nuclear hazardous and toxic material was also included in the scope of the characterization. This is one of the early buildings on the ANL-E site, and was heavily used in the 1950`s and 1960`s for various nuclear reaction and reactor design studies. Some degree of cleanup and contamination fixation was done in the 1970`s, so that the building could be used with a minimum of risk of personnel contamination. Work records are largely nonexistent for the early history of the building, so that any assumptions about extent and type of contamination had to be kept very open in the survey planning process. The primary contaminant was found to be painted-over Cs-137 embedded in the concrete floors, although a variety of other nuclides consistent with the work said to have been performed were found in smaller quantities. Due to leaks and drips through the floor, a relatively modest amount of soil contamination was found in the service trench under the building, not penetrating deeply. Two contaminated, disconnected drain lines leaving the building could not be traced by site records, and remain a problem for remediation. The D and D Characterization Plan was fulfilled.
Date: July 1, 1998
Partner: UNT Libraries Government Documents Department

Hanford Isotope Project strategic business analysis Cesium-137 (Cs-137)

Description: The purpose of this business analysis is to address the beneficial reuse of Cesium 137 (Cs-137) in order to utilize a valuable national asset and possibly save millions of tax dollars. Food irradiation is the front runner application along with other uses. This business analysis supports the objectives of the Department of Energy National Isotope Strategy distributed in August 1994 which describes the DOE plans for the production and distribution of isotope products and services. As part of the Department`s mission as stated in that document. ``The Department of Energy will also continue to produce and distribute other radioisotopes and enriched stable isotopes for medical diagnostics and therapeutics, industrial, agricultural, and other useful applications on a businesslike basis. This is consistent with the goals and objectives of the National Performance Review. The Department will endeavor to look at opportunities for private sector to co-fund or invest in new ventures. Also, the Department will seek to divest from ventures that can more profitably or reliably be operated by the private sector.``
Date: October 1, 1995
Partner: UNT Libraries Government Documents Department

Simulation of contaminated sediment transport in White Oak Creek basin

Description: This paper presents a systematic approach to management of the contaminated sediments in the White Oak Creek watershed at Oak Ridge National Laboratory near Oak Ridge, Tennessee. The primary contaminant of concern is radioactive cesium-137 ({sup 137}Cs), which binds to soil and sediment particles. The key components in the approach include an intensive sampling and monitoring system for flood events; modeling of hydrological processes, sediment transport, and contaminant flux movement; and a decision framework with a detailed human health risk analysis. Emphasis is placed on modeling of watershed rainfall-runoff and contaminated sediment transport during flooding periods using the Hydrologic Simulation Program- Fortran (HSPF) model. Because a large number of parameters are required in HSPF modeling, the major effort in the modeling process is the calibration of model parameters to make simulation results and measured values agree as closely as possible. An optimization model incorporating the concepts of an expert system was developed to improve calibration results and efficiency. Over a five-year simulation period, the simulated flows match the observed values well. Simulated total amount of sediment loads at various locations during storms match with the observed values within a factor of 1.5. Simulated annual releases of {sup 137}Cs off-site locations match the data within a factor of 2 for the five-year period. The comprehensive modeling approach can provide a valuable tool for decision makers to quantitatively analyze sediment erosion, deposition, and transport; exposure risk related to radionuclides in contaminated sediment; and various management strategies.
Date: December 31, 1995
Creator: Bao, Y.; Clapp, R.B.; Brenkert, A.L.; Moore, T.D. & Fontaine, T.A.
Partner: UNT Libraries Government Documents Department

Demonstration recommendations for accelerated testing of concrete decontamination methods

Description: A large number of aging US Department of Energy (DOE) surplus facilities located throughout the US require deactivation, decontamination, and decommissioning. Although several technologies are available commercially for concrete decontamination, emerging technologies with potential to reduce secondary waste and minimize the impact and risk to workers and the environment are needed. In response to these needs, the Accelerated Testing of Concrete Decontamination Methods project team described the nature and extent of contaminated concrete within the DOE complex and identified applicable emerging technologies. Existing information used to describe the nature and extent of contaminated concrete indicates that the most frequently occurring radiological contaminants are {sup 137}Cs, {sup 238}U (and its daughters), {sup 60}Co, {sup 90}Sr, and tritium. The total area of radionuclide-contaminated concrete within the DOE complex is estimated to be in the range of 7.9 {times} 10{sup 8} ft{sup 2}or approximately 18,000 acres. Concrete decontamination problems were matched with emerging technologies to recommend demonstrations considered to provide the most benefit to decontamination of concrete within the DOE complex. Emerging technologies with the most potential benefit were biological decontamination, electro-hydraulic scabbling, electrokinetics, and microwave scabbling.
Date: December 1, 1995
Creator: Dickerson, K.S.; Ally, M.R.; Brown, C.H.; Morris, M.I. & Wilson-Nichols, M.J.
Partner: UNT Libraries Government Documents Department

Migration of Sr-20, Cs-137, and Pu-239/240 in Canyon below Los Alamos outfall

Description: Technical Area-21 (TA-21) of Los Alamos National Laboratory (LANL) is on a mesa bordered by two canyons DP Canyon and Los Alamos (LA) Canyon. DP Canyon is a small semiarid watershed with a well defined channel system where the stream flow is ephemeral. TA-21 has had a complex history of waste disposal as research to determine the chemical and metallurgical properties of nuclear materials occurred here from 1945-1978. Due to these operations, the TA-21 mesa top and bordering canyons have been monitored and characterized by the LANL Environmental Restoration Program. Results identify radionuclide values at outfall. 21-011 (k) which exceed Screening Action Levels, and points along DP Canyon which exceed regional background levels. The radiocontaminants considered in this study are strontium-90, cesium-137, and plutonium-239. This research examines sediment transport and speciation of radionuclide contaminant migration from a source term named SWMU 21-011 (k) down DP Canyon. Three dimensional surface plots of data from 1977-1994 are used to portray the transport and redistribution of radioactive contaminants in an alluvial stream channel. An overall decrease in contamination concentration since 1983 has been observed which could be due to more stringent laboratory controls and also to the removal of main plutonium processing laboratories to another site.
Date: April 1, 1996
Creator: Murphy, J.M.; Mason, C.F.V.; Boak, J.M. & Longmire, P.A.
Partner: UNT Libraries Government Documents Department