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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

Processing of radioactive waste solutions in a vacuum evaporator- crystallizer

Description: Results of the first 18 months' operation of Hanford's vacuum evaporator- crystallizer are reported. This process reduces the volume of radioactive waste solutions and simultaneously converts the waste to a less mobile salt cake. The evaporator-crystallizer is operating at better than design production rates and has reduced the volume of radioactive wastes by more than 15 million gallons. A process description, plant performance data, mechanical difficulties, and future operating plans are discussed. Also discussed is a computer model of the evaporator-crystallizer process. (auth)
Date: September 26, 1975
Creator: Petrie, J.C.; Donovan, R.I.; Van der Cook, R.E. & Christensen, W.R.
Partner: UNT Libraries Government Documents Department

Research and development activities waste fixation program. Quarterly progress report, October--December 1973

Description: Two 40-hr nonradioactive spray solidification runs were completed in which both PU-4b and PW-6 reference waste compositions were successfully converted into borosilicate glass. The throughput capacity of the 13-in. diameter spray calciner while processing PW-4b waste is over 38 1/hr with a 700 deg C calciner wall and 24-30 1/hr with a 550 deg C wall. Mechanical agitation increases the capacity of the inconel meiter to over 4.5 l of melt per hour and greatly improves product homogeneity. Both PW-4b and PW-6 reference waste compositions were successfully concentrated in the nonradioactive, wiped film evaporator (WFE) facility. Either waste can be concentrated to approximately 60 wt% total solids in the WFE. For the PW-6 waste, this represents removal of about 80% of the initial liquid present in the waste. Two ceramic melters were built and operated, each using molybdenum electrodes for heating. While the performance of these first melters indicated a number of design and construction problems which need to be resolved, a ceramic-electrode heated melter appears to be a viable alternative to metallic melters. The PW-4b melt composition 72-68 contains finely divided crystallites when large quantities are melted at 1150 deg C. The crystallites were identified as CeO/sub 2/, a spinel and zircon, with CeO/ sub 2/ the predominant species. Several options for eliminating the crystallites include increasing the melt temperature, agitating the melt to prevent crystal settling, modification of the composition to promote CeO/sub 2/ solubility, and reducing the concentration of CeO/sub 2/ in the glass. Helium gas formed by alpha particle emission from actinide nuclides in high-level waste glasses can produce large internal stresses in the glass. A porosity of 0.2% in typical waste glasses, from uranium oxide fuel, may be sufficient to reduce the internal stresses to inconsequential levels by providing sites for gas accumulation. However, for ...
Date: January 1, 1974
Creator: Platt, A. M.
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

Salt waste volume reduction by sodium removal

Description: A literature searcha nd preliminary experiments were carried out to determine the feasibility of reducing salt waste volumes by the removal of sodium and purifying the sodium as metal for reuse or less restricted storage for use in the long-term storage of Hanford's radioactive salt waste. Included in the experimental part of the study were oxalate precipitation of sodium, preparation of chloride feed for ecletrolysis, denitration experiments, carbon reduction of sodium-cesium compounds, and distillation of sodium metal for decontamination. Eaeh of these steps was found tn be feasible, but mans problems exist. The most favorable process probable includes: evaporation and denitratipn (possibly preceded be a scavenging precipitatioal, reduction with carbon and distillation or sodium metal from the reaction mixture at low pressures, and purification of sodium metal be filtration and distillaton. At alternative to the first step might be oxalate precipitaton of the sodium. Considerable development work remains before at integrated process could be demonstrated. (auth)
Date: September 15, 1973
Creator: Burger, L.L.; Ryan, J.L.; Swanson, J.L. & Bray, L.A.
Partner: UNT Libraries Government Documents Department