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HEAT--A ONE-DIMENSIONAL HEAT TRANSFER EQUATION CODE FOR THE IBM-704

Description: dimensional solution to the general heat transfer equation is presented. Specifically written for application in reactor fuel red design, the code requires cylindrical geometry conditions and input parameters of surface temperature and power density. The maximum number of points for which temperature values may be obtained is 251, and the approximate running time for a typical problem varies from 1.0 to 2.0 minutes. (auth)
Date: January 1, 1959
Creator: King, C.M. & Boyle, R.F.
Partner: UNT Libraries Government Documents Department

Dostoman Code: A Compartmental Pathways Computer Model of Contaminant Transport

Description: Assessing the impact of radioactive and hazardous chemical waste disposal on man is an important problem in environmental science and engineering. This report illustrates the mathematical evolution of the compartmental model from small to large systems and provides examples of the use of the compartmental approach in analysis of transport of radionuclide and chemical contaminants.
Date: August 19, 2002
Creator: King, C.M.
Partner: UNT Libraries Government Documents Department

Radionuclide-migration model for buried waste at the Savannah River Plant

Description: Solid waste has been buried at the Savannah River Plant burial ground since 1953. The solid waste is contaminated with alpha-emitting transuranium (TRU) nuclides, with beta-gamma-emitting activation and fission products, and with tritium. To provide guidance for the current use and eventual permanent retirement of the burial site from active service, a radionuclide environmental transport model has been used to project the potential influence on man if the burial site were occupied after decommissioning. The model used to simulate nuclide migration includes the various hydrological, animal, vegetative, atmospheric, and terrestrial pathways in estimating dose to man as a function of time. Specific scenarios include a four-person home farm on the 195-acre burial ground. Key input to the model includes site-specific nuclide migration rates through soil, nuclide distribution coefficients, and site topography. Coupled with literature data on plant and animal concentration factors, transfer coefficients reflecting migration routes are input to a set of linear differential equations for subsequent matrix solution. Output from the model is the nuclide-specific decayed curie intake by man. To discern principal migration routes, model-compartment inventories with time can also be displayed. Dose projections subsequently account for organ concentrations in man for the nuclide of interest. Radionuclide migration has been examined in depth with the dose-to-man model. Movement by vegetative pathways is the primary route for potential dose to man for short-lived isotopes. Hydrological routes provide a secondary scheme for long-lived nuclides. Details of model methodology are reviewed.
Date: January 1, 1982
Creator: King, C M & Root, Jr, R W
Partner: UNT Libraries Government Documents Department

Thermal and Radiolytic Gas Generation Tests on Material from Tanks 241-U-103, 241-AW-101, 241-S-106, and 241-S-102: Status Report

Description: This report summarizes progress in evaluating thermal and radiolytic flammable gas generation in actual Hanford single-shell tank wastes. The work described was conducted at Pacific Northwest National Laboratory (PNNL) for the Flammable Gas Safety Project, whose purpose is to develop information to support DE&S Hanford (DESH) and Project Management Hanford Contract (PHMC) subcontractors in their efforts to ensure the safe interim storage of wastes at the Hanford Site. This work is related to gas generation studies performed by Numatec Hanford Corporation (formerly Westinghouse Hanford Company). This report describes the results of laboratory tests of gas generation from actual convective layer wastes from Tank 241-U-103 under thermal and radiolytic conditions. Accurate measurements of gas generation rates from highly radioactive tank wastes are needed to assess the potential for producing and storing flammable gases within the tanks. The gas generation capacity of the waste in Tank 241-U-103 is a high priority for the Flammable Gas Safety Program due to its potential for accumulating gases above the flammability limit (Johnson et al, 1997). The objective of this work was to establish the composition of gaseous degradation products formed in actual tank wastes by thermal and radiolytic processes as a function of temperature. The gas generation tests on Tank 241-U-103 samples focused first on the effect of temperature on the composition and rate of gas generation Generation rates of nitrogen, nitrous oxide, methane, and hydrogen increased with temperature, and the composition of the product gas mixture varied with temperature.
Date: June 17, 1999
Creator: King, C.M. & Bryan, S.A.
Partner: UNT Libraries Government Documents Department

Isolating /sup 241/Am from waste solutions containing Al, Ca, Fe, and Cr

Description: About 2.4 kg of /sup 241/Am contaminated with calcium and aluminum had been recovered from low-activity waste during recycle of 11% /sup 240/Pu. A process was developed and demonstrated to purify the americium before shipment as /sup 241/AmO/sub 2/. The americium and some of the calcium were batch extracted into 50% TBP-n-paraffin from 2.2M Al(NO/sub 3/)/sub 3/ - 0.3M HNO/sub 3/ solution in a canyon tank. Pregnant solvent was scrubbed first with 2.1M Al/sup 3 +/-0.3M Li/sup +/-6.7M NO/sub 3/- and then with 7M LiNO/sub 3/ to reduce the calcium content and to displace the aluminum. Americium was then stripped from the solvent with water and concentrated by evaporation. Before precipitating the americium with oxalic acid, the nitric acid was adjusted with NH/sub 4/OH to yield a 1M NH/sub 4/NO/sub 3/ solution. Recovery across the batch extraction step was 97.8%, while 93% of the calcium and >99% of the aluminum was rejected. Recovery across precipitation averaged >96% while producing a product which was >99.3% pure /sup 241/AmO/sub 2/. The major impurities were water, carbon, calcium, iron, and zinc.
Date: January 1, 1982
Creator: Gray, L.W.; Burney, G.A. & King, C.M.
Partner: UNT Libraries Government Documents Department

DRACO--A THREE-DIMENSIONAL FEW-GROUP DEPLETION CODE FOR THE IBM-704

Description: A three-dimensional few-group depletion code prograrnmed for the IBM-704 is presented, The code, called DRACO, is used in studying the neutron flux, the power level, and the related buildup and depletion of materials at different stages in a reactor lifetime. The three sections of the code are described, and the background documents are referenced. (auth)
Date: December 1, 1958
Creator: McCarty, D.S.; King, C.M.; Mandel, J.T. & Henderson, H.P.
Partner: UNT Libraries Government Documents Department

Risk assessment of mixed waste sites

Description: As part of its ongoing efforts to ensure environmental regulation compliance at DOE facilities, DOE published on April 26, 1985, a notice of intent to write an Environmental Impact Statement on Waste Management Activities for Groundwater Protection (Groundwater EIS) at the Savannah River Plant (SRP). To perform a human health risk assessment of each waste site for each closure action considered, DuPont organized a project team led by personnel from the Savannah River Laboratory (SRL) and supported by outside contractors specializing in risk assessment work. As part of that team, JBF Associates, Inc. (JBFA) performed an atmospheric containment transport analysis and human health risk assessment of nonradioactive contaminants from SRP waste sites. For each waste site, three closure actions were examined: (1) excavate the site, backfill it, and cap it followed by regular groundwater monitoring (Option 1); (2) backfill and cap the site followed by regular groundwater monitoring (Option 2); and (3) no remedial action, regular groundwater monitoring, and some site maintenance work (Option 3). The human health risk assessment performed by JBFA estimated the public and worker risks from contaminants released to the atmosphere from each waste site for each closure option. This paper first presents the methodology JBFA used to estimate the public and worker risks attributable to the inhalation and ingestion of airborne, nonradioactive contaminants. Following the description of the analysis methodology, the authors present the risk results for the waste sites that were due to atmospherically released nonradioactive contaminants. Both worker risks and public risks are presented. Finally, the authors present the results and conclusions derived from their analysis of the risk from airborne, nonradioactive contaminants.
Date: December 31, 1987
Creator: Montague, D.F.; Holton, G.A. & King, C.M.
Partner: UNT Libraries Government Documents Department

Factors for assessment of human health risk associated with remedial action at hazardous waste sites

Description: A risk assessment strategy that is cost effective and minimized human health risks was developed for closure of hazardous waste sites at the Savannah River Plant. The strategy consists of (1) site characterization, (2) contaminant transport modeling, and (3) determination of relative merits of alternative remedial actions according to the degree of health protection they provide.
Date: January 1, 1985
Creator: Stephenson, D E; King, C M; Looney, B B; Holmes, W G & Gordon, D E
Partner: UNT Libraries Government Documents Department

Waste migration studies at the Savannah River Plant burial ground

Description: The low-level radioactive waste burial ground at the Savannah River Plant is a typical shallow-land-burial disposal site in a humid region. Studies of waste migration at this site provide generic data for designing other disposal facilities. A program of field, laboratory, and modeling studies for the SRP burial ground has been conducted for several years. Recent results of lysimeter tests, soil-water chemistry studies, and transport modeling are reported. The lysimeter experiments include ongoing tests with 40 lysimeters containing a variety of defense wastes, and recently concluded lysimeter tests with tritium and plutonium waste forms. The tritium lysimeter operated 12 years. In chemistry studies, measurements of soil-water distribution coefficients (K/sub d/) were concluded. Current emphasis is on identification of trace organic compounds in groundwater from the burial site. Development of the dose-to-man model was completed, and the computer code is available for routine use. 16 refs., 2 figs., 2 tabs.
Date: January 1, 1985
Creator: Stone, J A; Oblath, S B; Hawkins, R H; Grant, M W; Hoeffner, S L & King, C M
Partner: UNT Libraries Government Documents Department

Concepts for detritiation of waste liquids

Description: Tritium is formed in thermal nuclear reactors both by neutron activation of elements such as deuterium and lithium and by ternary fission in the fuel. It is a weak beta-emitter with a short half-life, 12.3 years, and its radiological significance in reactor discharges is very low. In heavy-water-cooled and -moderated reactors, such as the SRS reactors, the tritium concentration in the moderator is sufficiently high to cause a potential hazard to operators, so research and development programs have been carried out on processes to remove the tritium. Detritiation of light water has also been the subject of major R&D efforts world-wide, because reprocessing operations can generate significant quantities of tritium in liquid waste, and high concentrations of tritium may arise in some aqueous streams in future fusion reactors. This paper presents a review of some of the methods that have been proposed, studied, and developed for removal of tritium from light and heavy water, along with some new concepts for aqueous detritiation directly from liquid oxide (HTO) bearing feed streams.
Date: December 31, 1991
Creator: King, C. M.; Van Brunt, V.; Garber, A. R. & King, R. B.
Partner: UNT Libraries Government Documents Department

Spectroscopic probes of the structure of hydrous uranium oxide precursors to UO{sub 2} ceramic fuel

Description: Fourier Transform infrared spectroscopy, x-ray powder diffraction and thermal analysis show that one example of ``ammonium diuranate`` observed as an intermediate in the U(VI) sol-gel process is a layered hydrous uranium oxide with a proposed structural formula of (NH){sub 4}{sub 2}[(UO{sub 2}){sub 8}O{sub 4}(OH){sub 10}]{center_dot}8H{sub 2}O, an ammonium ion intercalate. Examples of polyamine intercalation compounds hydrous uranium oxide are also given.
Date: December 31, 1989
Creator: Thompson, M. C.; King, C. M. & King, R. B.
Partner: UNT Libraries Government Documents Department

Thermal and radiolytic gas generation from Tank 241-S-102 waste

Description: This report summarizes progress in evaluating thermal and radiolytic rate parameters for flammable gas generation in Hanford single-shell tank wastes based on the results of laboratory tests using actual waste from Tank 241-S-102 (S-102). Work described in this report was conducted at Pacific Northwest National Laboratory (PNNL) for the Flammable Gas Safety Project, whose purpose is to develop information to support Fluor Daniel Hanford (FDH) and its Project Management Hanford Contract (PHMC) subcontractors in their efforts to ensure the safe interim storage of wastes at the Hanford Site. This work is related to gas generation studies being performed at Georgia Institute of Technology (GIT) under subcontract to PNNL, using simulated wastes, and to studies being performed at Numatec Hanford Corporation (formerly Westinghouse Hanford Company) using actual wastes. The results of gas generation from Tank S-102 waste under thermal and radiolytic conditions are described in this report. The accurate measurement of gas generation rates in actual waste from highly radioactive waste tanks is needed to assess the potential for producing and storing flammable gases within the waste tanks. This report addresses the gas generation capacity of the waste from Tank S-102, a waste tank listed as high priority by the Flammable Gas Safety Program due to its potential for flammable gas accumulation above the flammability limit.
Date: July 1, 1997
Creator: King, C.M.; Pederson, L.R. & Bryan, S.A.
Partner: UNT Libraries Government Documents Department

Gas generation from Tank 241-SY-103 waste

Description: This report summarizes progress made in evaluating mechanisms by which flammable gases are generated in Hanford double-shell tank wastes, based on the results of laboratory tests using actual waste from Tank 241-SY-103. The objective of this work is to establish the identity and stoichiometry of degradation products formed in actual tank wastes by thermal and radiolytic processes as a function of temperature. The focus of the gas generation tests on Tank 241-SY-103 samples is first the effect of temperature on gas generation (volume and composition). Secondly, gas generation from irradiation of Tank 241-SY-103 samples at the corresponding temperatures as the thermal-only treatments will be measured in the presence of an external radiation source (using a {sup 137}Cs capsule). The organic content will be measured on a representative sample prior to gas generation experiments and again at the termination of heating and irradiation. The gas generation will be related to the extent of organic species consumption during heating. Described in this report are experimental methods used for producing and measuring gases generated at various temperatures from highly radioactive actual tank waste, and results of gas generation from Tank 241-SY-103 waste taken from its convective layer. The accurate measurement of gas generation rates from actual waste from highly radioactive waste tanks is needed to assess the potential for producing and storing flammable gases within the waste tanks. This report addresses the gas generation capacity of the waste from the convective layer of Tank 241-SY-103, a waste tank listed on the Flammable Gas Watch List due to its potential for flammable gas accumulation above the flammability limit.
Date: April 1, 1996
Creator: Bryan, S.A.; King, C.M.; Pederson, L.R.; Forbes, S.V. & Sell, R.L.
Partner: UNT Libraries Government Documents Department

Evaulation of B{sub 4}C as an ablator material for NIF capsules. Revision 1

Description: Boron carbide (B{sub 4}C) is examined as a potential fuel container and ablator for implosion capsules on the National Ignition Facility (NIF). A capsule of pure B{sub 4}C encasing a layer of solid DT implodes stably and ignites with anticipated NIF x-ray drives, producing 18 MJ of energy. Thin films of B{sub 4}C were found to be resistant to oxidation and modestly transmitting in the infrared (IR), possibly enabling IR fuel characterization and enhancement for thin permeation barriers but not for full-thickness capsules. Polystyrene mandrels 0.5 mm in diameter were successfully coated with 0.15-2.0 micrometers of B{sub 4}C. Thickness estimated from optical density agreed well with those measured by scanning electron microscopy (SEM). The B{sub 4}C microstructure was columnar but finer than for Be made at the same conditions. B{sub 4}C is a very strong material, with a fiber tensile strength capable of holding NIF fill pressures at room temperature, but it is also very brittle, and microscopic flaws or grain structure may limit the noncryogenic fill pressure. Argon (Ar) permeation rates were measured for a few capsules that had been further coated with 5 micrometers of plasma polymer. The B{sub 4}C coatings tended to crack under tensile load. Some shells filled more slowly than they leaked, suggesting that the cracks open and close under opposite pressure loading. As observed earlier for Ti coatings, 0.15-micrometer layers of B{sub 4}C had better gas retention properties than 2-micrometer layers, possibly because of fewer cracks. Permeation and fill strength issues for capsules with a full ablator thickness of B{sub 4}C are unresolved. 21 refs., 6 figs.
Date: March 26, 1997
Creator: Burnham, A.K.; Alford, C.S.; Makowiecki, D.M.; Dittrich, T.R.; Wallace, R.J.; Honea, E.C. et al.
Partner: UNT Libraries Government Documents Department

Oxygen-17 NMR studies on uranium (VI) hydrolysis and gelation

Description: Hydrolysis and gelation processes in uranyl solutions are observed using the strong sharp uranyl oxygen-17 resonance. The ability to follow the hydrolysis of uranyl salts by observation of the sharp uranyl oxygen-17 resonance provides a clear indication of the dependence of uranyl hydrolysis on the counteranion (nitrate versus chloride) but not on the means of introducing hydroxide into the solution (Me{sub 4}NOH versus R{sub 3}N extraction). In addition, two different pathways for gelation are suggested. In the first pathway the uranyl hydrolysis is conducted with a base (HMTA in these studies) which preferentially forms trimeric (UO{sub 2}){sub 3} ({mu}{sub 3}-O) units which can then condense into the polymeric UO{sub 2}O{sub 6/3} layers of a gel based on the hexagonal structure of {proportional_to}UO{sub 2}(OH){sub 2}. In the second gelation pathway a uranyl derivative is treated with excess hydroxide in the absence of a metal or hydrogen-bonding ammonium cations which form insoluble solids uranates. Consensation of the resulting solution of soluble UO{sub 2}(OH)n{sup 2-n} anions can then lead to a similar polymer UO{sub 2}O{sub 4/2} or UO{sub 2}O{sub 6/3} structure of a gel. 9 refs., 2 figs.
Date: December 31, 1989
Creator: King, R. B.; King, C. M. & Garber, A. R.
Partner: UNT Libraries Government Documents Department

Savannah River Laboratory DOSTOMAN code: a compartmental pathways computer model of contaminant transport

Description: The Savannah River Laboratory DOSTOMAN code has been used since 1978 for environmental pathway analysis of potential migration of radionuclides and hazardous chemicals. The DOSTOMAN work is reviewed including a summary of historical use of compartmental models, the mathematical basis for the DOSTOMAN code, examples of exact analytical solutions for simple matrices, methods for numerical solution of complex matrices, and mathematical validation/calibration of the SRL code. The review includes the methodology for application to nuclear and hazardous chemical waste disposal, examples of use of the model in contaminant transport and pathway analysis, a user's guide for computer implementation, peer review of the code, and use of DOSTOMAN at other Department of Energy sites. 22 refs., 3 figs.
Date: January 1, 1985
Creator: King, C M; Wilhite, E L; Root, Jr, R W; Fauth, D J; Routt, K R; Emslie, R H et al.
Partner: UNT Libraries Government Documents Department

Concepts for detritiation of waste liquids

Description: Tritium is formed in thermal nuclear reactors both by neutron activation of elements such as deuterium and lithium and by ternary fission in the fuel. It is a weak beta-emitter with a short half-life, 12.3 years, and its radiological significance in reactor discharges is very low. In heavy-water-cooled and -moderated reactors, such as the SRS reactors, the tritium concentration in the moderator is sufficiently high to cause a potential hazard to operators, so research and development programs have been carried out on processes to remove the tritium. Detritiation of light water has also been the subject of major R D efforts world-wide, because reprocessing operations can generate significant quantities of tritium in liquid waste, and high concentrations of tritium may arise in some aqueous streams in future fusion reactors. This paper presents a review of some of the methods that have been proposed, studied, and developed for removal of tritium from light and heavy water, along with some new concepts for aqueous detritiation directly from liquid oxide (HTO) bearing feed streams.
Date: January 1, 1991
Creator: King, C.M. (Westinghouse Savannah River Co., Aiken, SC (United States)); Van Brunt, V.; Garber, A.R. (South Carolina Univ., Columbia, SC (United States)) & King, R.B. (Georgia Univ., Athens, GA (United States). Dept. of Chemistry)
Partner: UNT Libraries Government Documents Department

Spectroscopic probes of the structure of hydrous uranium oxide precursors to UO sub 2 ceramic fuel

Description: Fourier Transform infrared spectroscopy, x-ray powder diffraction and thermal analysis show that one example of ammonium diuranate'' observed as an intermediate in the U(VI) sol-gel process is a layered hydrous uranium oxide with a proposed structural formula of (NH){sub 4}{sub 2}((UO{sub 2}){sub 8}O{sub 4}(OH){sub 10}){center dot}8H{sub 2}O, an ammonium ion intercalate. Examples of polyamine intercalation compounds hydrous uranium oxide are also given.
Date: January 1, 1989
Creator: Thompson, M.C.; King, C.M. (Westinghouse Savannah River Co., Aiken, SC (United States)) & King, R.B. (Georgia Univ., Athens, GA (United States). Dept. of Chemistry)
Partner: UNT Libraries Government Documents Department

New insights into uranium (VI) sol-gel processing

Description: Nuclear Magnetic Resonance (NMR) investigations on the Oak Ridge National Laboratory process for sol-gel synthesis of microspherical nuclear fuel (UO{sub 2}), has been extremely useful in sorting out the chemical mechanism in the sol-gel steps. {sup 13}C, {sup 15}N, and {sup 1}H NMR studies on the HMTA gelation agent (Hexamethylene tetramine, C{sub 6}H{sub 12}N{sub 4}) has revealed near quantitative stability of this adamantane-like compound in the sol-gel process, contrary to its historical role as an ammonia source for gelation from the worldwide technical literature. {sup 17}O NMR of uranyl (UO{sub 2}{sup ++}) hydrolysis fragments produced in colloidal sols has revealed the selective formation of a uranyl trimer, ((UO{sub 2}){sub 3}({mu}{sub 3}-O)({mu}{sub 2}-OH){sub 3}){sup +}, induced by basic hydrolysis with the HMTA gelation agent. Spectroscopic results will be presented to illustrate that trimer condensation occurs during sol-gel processing leading to layered polyanionic hydrous uranium oxides in which HMTAH{sup +} is occluded as an intercalation'' cation. Subsequent sol-gel processing of microspheres by ammonia washing results in in-situ exchange and formation of a layered hydrous ammonium uranate with a proposed structural formula of (NH{sub 4}){sub 2} ((UO{sub 2}){sub 8} O{sub 4} (OH){sub 10}) {center dot} 8H{sub 2}O. This compound is the precursor to sintered UO{sub 2} ceramic fuel. 23 refs., 10 figs.
Date: January 1, 1990
Creator: King, C.M.; Thompson, M.C.; Buchanan, B.R. (Westinghouse Savannah River Co., Aiken, SC (USA)); King, R.B. (Georgia Univ., Athens, GA (USA). Dept. of Chemistry) & Garber, A.R. (South Carolina Univ., Columbia, SC (USA). Dept. of Chemistry)
Partner: UNT Libraries Government Documents Department

Oxygen-17 NMR studies on uranium (VI) hydrolysis and gelation

Description: Hydrolysis and gelation processes in uranyl solutions are observed using the strong sharp uranyl oxygen-17 resonance. The ability to follow the hydrolysis of uranyl salts by observation of the sharp uranyl oxygen-17 resonance provides a clear indication of the dependence of uranyl hydrolysis on the counteranion (nitrate versus chloride) but not on the means of introducing hydroxide into the solution (Me{sub 4}NOH versus R{sub 3}N extraction). In addition, two different pathways for gelation are suggested. In the first pathway the uranyl hydrolysis is conducted with a base (HMTA in these studies) which preferentially forms trimeric (UO{sub 2}){sub 3} ({mu}{sub 3}-O) units which can then condense into the polymeric UO{sub 2}O{sub 6/3} layers of a gel based on the hexagonal structure of {proportional to}UO{sub 2}(OH){sub 2}. In the second gelation pathway a uranyl derivative is treated with excess hydroxide in the absence of a metal or hydrogen-bonding ammonium cations which form insoluble solids uranates. Consensation of the resulting solution of soluble UO{sub 2}(OH)n{sup 2-n} anions can then lead to a similar polymer UO{sub 2}O{sub 4/2} or UO{sub 2}O{sub 6/3} structure of a gel. 9 refs., 2 figs.
Date: January 1, 1989
Creator: King, R.B. (Georgia Univ., Athens, GA (United States). Dept. of Chemistry); King, C.M. (Westinghouse Savannah River Co., Aiken, SC (United States)) & Garber, A.R. (South Carolina Univ., Columbia, SC (United States). Dept. of Chemistry)
Partner: UNT Libraries Government Documents Department