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Gas-phase adsorbents for trapping radioactive iodine and iodine compounds

Description: This standard covers granular adsorbents (activated carbon, mineral base, polymer, etc.) for use in air and gas treatment systems. The absorbents are used in both thin-bed absorber cells and deep-bed systems. The standard includes document list, technical requirements, quality assurance requirements, and preparation for delivery. Test and sampling procedures are given. (DLC)
Date: October 1, 1973
Partner: UNT Libraries Government Documents Department

Endothermic process: application to immobilization of Hanford in-tank solidified waste

Description: Conversion of high-level solid radioactive waste to a nonleachable silicate glass is accomplished in the Endothermic process by simple melting of a mixture of the waste with crushed basalt. Dense (2.5 g cm/sup -3/) green-black glasses are obtained by melting mixtures containing 30 to 40 wt% Hanford In-Tank Solidified (ITS) waste, 50 to 70 wt% basalt, and 0 to 10 wt% B/sub 2/O/sub 3/. Addition of B/sub 2/O/sub 3/ to the process charge is desirable to lower its melting range from about 1100--1150 deg C to 1000--1050 deg C. Leach rates of these glasses (calculated from the sum of the concentrations of Fe, Na, Ca, Si, Mg, Al, Sr, and Cs in the leach liquor) in water at 25 deg C range from 10/sup -7/ to 10/sup -5/ g/cm/sup 2/ day. The lea ch rate, based on /sup 137/Cs, of a typical Endothermic process glass made from actual ITS waste, is 3.0 x 10/sup -8/ g/cm/sup 2/ day. This leach rate corresponds to removal of 2.1 x 10/sup -14/ g / sup 137/Cs per day from a square centimeter of glass containing 57 mu Ci /sup 137/Cs per gram. Judging from initial tests, the Endothermic process is a very promising scheme to increase immobilization of the ITS waste. (auth)
Date: July 1, 1973
Creator: Kupfer, M.J. & Schulz, W.W.
Partner: UNT Libraries Government Documents Department

Post treatment of high-level nuclear fuel wastes

Description: The glass-ceramic product prepared from fluidized-bed calcined synthetic commercial wastes, based on data obtained to date, has many of the properties desired for long-term storage. Although more characterization is necessitated, the product's high-calcine content will decrease the number of storage canisters required and use a minimum of product-forming additives, resulting in significant process cost savings. The product remains in a solid, nonflowing form at temperatures close to the preparation temperature and yet is prepared at relatively low temperatures. The product has void spaces to accommodate radiolytic gas formation, but is hard and dense and has very low leach rates. Process features, such as no direct product contact with furnace or storage canisters, will minimize corrosion of both process equipment and storage canisters. (auth)
Date: January 1, 1975
Creator: Berreth, J.R.; Cole, H.S.; Hoskins, A.P.; Lewis, L.C. & Samsel, E.G.
Partner: UNT Libraries Government Documents Department

Experience with waste vitrification systems at Battelle-Northwest

Description: Three types of melters; in-can, continuous metallic, and joule-heated ceramic are being developed on an engineering scale for conversion of simulated high-level radioactive waste to a glass form. Work with each of the three melters has progressed for over a year, and ton quantities of glass have been produced. The operation and performance of these systems are described. (auth)
Date: January 1, 1975
Creator: Chapman, C.C.; Blair, H.T. & Bonner, W.F.
Partner: UNT Libraries Government Documents Department

Concentration of aqueous radioactive waste with wiped-film evaporators

Description: Tests at the Savannah River Laboratory with two small wiped-film evaporators show that synthetic alkaline (Purex) waste can be converted to a free- flowing slurry that solidifies on cooling to ambient temperature. The desired concentration can be obtained in one pass rather than the several passes required with the bent-tube evaporators presently used at the Savannah River Plant. (auth)
Date: January 1, 1975
Creator: Goodlett, C.B.
Partner: UNT Libraries Government Documents Department

Use of ion exchange for the treatment of liquids in nuclear power plants

Description: The current and future use of ion exchange (demineralization) as a method for treating liquid radioactive streams at nuclear power plants was investigated. Pertinent data were obtained by contacting utility companies, nuclear-steam-supply system vendors, selected AEC-operated facilities, as well as ion exchange resin and equipment manufacturers. Principal emphasis was on obtaining data concerning the decontamination of aqueous solutions characterized by levels of radioactivity that range from 10/sup -7/ to 1 mu Ci/ml. Ion exchange media commonly used in nuclear power plants are synthetic organic resins of polystyrene matrix. They are utilized primarily in the mixed-bed (deep-bed) ion exchange system. Powdered resin (mixed) systems (so-called filter- demineralizer'') are also used in several recent boiling-water-reactor plants. The term decontamination factor (DF), the ratio of the feed to effluent concentration, is widely used and is assumed by designers and operators of the plants to express the ion exchange system performance. In some cases, such DF values may not represent the true system performance. To achieve a desired DF, the feed and effiuent must be sampled for the nuclides of interest and the processing discontinued when the desired effluent concentration is exceeded. Average DF values that can be obtained for various ion-exchange systems and various groups of radionuclides if good engineering practice is used in the design and operation of these systems are listed. These values are based on ion- exchange fundamentals, literature data, laboratory experiments, and plant operating experience. They represent time-average values expected under normal operating conditions rather than maximum values attainable under optimum conditions. (auth)
Date: December 1, 1973
Creator: Lin, K.H.
Partner: UNT Libraries Government Documents Department

Performance of a wiped film evaporator with simulated high level waste slurries

Description: The horizontal, reverse taper, wiped film evaporator that was evaluated demonstrated a number of positive characteristics with respect to its applicability in the solidification of nuclear fuel recovery process wastes. Foremost among these is its ability to remove the bulk (80 to 90 percent) of the liquid associated with any of the purex-type high level, intermediate level, or mixed waste slurries. The major disadvantage of the evaporator is its current inability to discharge a product that is low enough in liquid content to avoid sticking to the evaporator discharge nozzle. Also, while the indirect indications of the torque required to turn the rotor and the power drawn by the drive motor are indicative of the liquid content of the discharged product, no reliable correlation has been found to cover all of the possible flow rates and feed stock compositions that the evaporator may be required to handle. In addition, no reliable means has been found to indicate the presence or absence of product flow through the discharge nozzle. The lack of a positive means of moving the product concentrate out of the evaporator and into a high temperature receiver is an undesirable feature of the evaporator. Pulverized glass former, or frit, was added to the evaporator feedstock in a ratio of frit to metal oxides of 2 to 1, and the resulting mixture successfully evaporated to a concentrate containing about 50 percent solids. In general, the performance of the wiped film evaporator evaluated was favorable for its use in a nuclear waste fixation process, however further development of the rotor design, power input, and operating techniques will be required to produce a free flowing solid product. (auth)
Date: January 1, 1975
Creator: Dierks, R.D. & Bonner, W.F.
Partner: UNT Libraries Government Documents Department