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Computer simulation of noncondensible gas behavior in geothermal power plants utilizing direct contact heat exchange. Report of work, February 1, 1980-February 28, 1981

Description: A computer model was developed to simulate the behavior of carbon dioxide and hydrogen sulfide in a geothermal power plant using direct contact heat exchange with isobutane as a working fluid. This computer program was modified to simulate the particular equipment characteristics of the 500 kW direct contact pilot plant at East Mesa. Vapor and liquid compositions and temperatures can be calculated throughout the heat exchangers in the pilot plant. The program is now available for analysis of the pilot plant operation and for design of similar plants.
Date: January 1, 1981
Creator: Perona, J.J.
Partner: UNT Libraries Government Documents Department

OPTIMUM FILL VOLUMES IN POT CALCINATION OF RADIOACTIVE WASTES

Description: The 15,000 MW nuclear economy assumed for the long range study of pot calcination costs reported earlier was used as a basis for calculating optimum fill volumes. An algebraic expression was developed for cost as a functmon of the normalized radius of the central void space in a partially filled vessel. Minima of this expression were found for acmdmc and neutralized wastes in 6, 12, and 24in.-diameter vessels. Optimum fill volumes decreased as vessel diameter increased, varying for acidic wastes from 99.8% for 6-in.-diameter vessels to 92.5% for 24-in.diameter vessels. Decreases in costs by using optimum fill volumes instead of the 90% fill volume assumed for all cases in the long range study were small, the largest being an 8% decrease for neutralized wastes in 6- in.-diameter vessels. (auth)
Date: November 17, 1961
Creator: Perona, J.J.
Partner: UNT Libraries Government Documents Department

THE EFFECTS OF INTERNAL HEAT GENERATION ON POT CALCINATION RATES FOR RADIOACTIVE WASTES

Description: Methods by which the radial deposition mechanism was determined in experiments with simulated waste solutions are reviewed. Based on this mechanism, an expression for the rate of solid deposition with internal heat generation was developed by a combined heat and material balance. A sample calculation for Purex waste showed that a moderate heat generation rate of 5000 Btu/hr/ft/sup 3/ would decrease the time to fill a 12-in.-dia calcination vessel from 78 to 55 hr. For the calcination stage of the process in which the deposited solids are heated in the absence of a liquid phase, a solution was developed for the equation of heat transfer with the temperature profile from the solid deposition stage as an initial condition. For the example Purex waste with a heat generation rate of 5000 Btu/hr ft/sup 3/, less than 15 min would be required for calcination, compared to about 8 hr in experiments with simulated wastes. (auth)
Date: October 23, 1961
Creator: Perona, J.J.
Partner: UNT Libraries Government Documents Department

CALCULATION OF TEMPERATURE RISE IN DEEPLY BURIED RADIOACTIVE CYLINDERS

Description: Temperatures were calculated relative to the storage of radioactive solid waste as a function of time and radial distance for radioactive solid cylinders in infinite solid media of "average soil," "average rock," and salt. A resistance at the cylinder--infinite medium boundary was included in the form of an air space. For the range of parameters used and withia the practical limits of accuracy, the maximum temperature rise increased linearly with the heat generation rate. The fission product spectrum was not significant in the determination of the maximum temperature rise. Under the pessimistic storage conditions assumed, the storage of cylinders of a practical size appears feasible without excessive temperature rise. A maximum temperature rise of 1000 deg F would be produced with an initial heat generation rate of 1300 to 1600 Btu/hr-ft/ sup 3/ for cylinders with a 5-in. radius, with 350 to 450 Btu/hr-ft/sup 3/ for a 10-in. radius, and with 175 to 210 Btu/hr-ft/sup 3/ for a 15-in. radius, assuming a thermal conductivity of the radioactive cylinder of 0. 1 Btu/hr-ft- deg F. (auth)
Date: February 25, 1960
Creator: Perona, J. J. & Whatley, M. E.
Partner: UNT Libraries Government Documents Department

A PRELIMINARY STUDY OF PRE-SOLVENT EXTRACTION TREATMENT OF STAINLESS STEEL- URANIUM FUELS WITH DILUTE AQUA REGIA

Description: The continuous dissolution of 304 stainless steel and stainless steel - UO/sub 2/ alloy in dilute aqua regia was studied with subsequent stripping of the dissolver product to remove chloride ion. The process has the advantage of producing, by means of a simple head end treatment, a solvent extract feed in a conventional nitric acid medium so that existing solvent extraction processes, materials of construction and waste disposal methods can be used. The purposes of this study were to investigate the the variables affecting the dissolution process and to obtain dissolver scale-up data, and to investigate the removal of chloride from the dissolver product and the variables affecting the stripping operation. A continuous flooded pot dissolver was used. It has the advantages of stability of operation and ease of control in comparison with column dissolvers and requires a minimum of mechanical processing prior to dissolution. Stripping of the dissolver prcduct to remove chloride ion was studied in a 4-in. diameter Pyrek bubblecap column containing 12 single babble cap plates. Continuous dissolution rates and dissolver product stainiess steel loadings were correlatsd with aqua regla feed composition, acid feed rats and surface area exposed to reaction. Profiles of chloride concentration down the stripping column were obtained for various vapor to liquid mole ratios and for several nitric acid stripping vapor concentrations. Noncondensable off-gas compositions and rates were also measured. (auth)
Date: October 11, 1957
Creator: Kitts, F.G. & Perona, J.J.
Partner: UNT Libraries Government Documents Department

CALCULATIONAL MODELS OF POT CALCINATION

Description: A simplified model for solids deposition in the pot calcination of waste was analyzed, and numerical calculations were made. In long calcination pots of 10 to 12 in. diameter, calcination times should not exceed 24 hours and might be as low as three hours if the pot is kept full. If the pots are fed at a constant rate, the cake might form with a steady state V'' when viewed in vertical section which would progress from bottom to top. Cake deposition rates appear to be independent of pot radius. Several advantages to using larger diameter pots are discussed. (auth)
Date: March 23, 1959
Creator: Whatley, M.E. & Perona, J.J.
Partner: UNT Libraries Government Documents Department

Sulfur hexafluoride purification from mixtures with air: a process feasibility study

Description: Studies were made of the purification of SF/sub 6/ vapor contaminated with air for application at the Holifield Heavy-Ion Research Facility. Liquefaction appears to be a good method for recovering about 90% of the SF/sub 6/ if it is badly contaminated (15% air), and an even greater fraction can be recovered from mixtures containing less air. In cases where liquefaction is insufficient by itself, adsorption of SF/sub 6/ on activated carbon at -50/sup 0/F is promising. Two carbon beds, each containing about 500 lb of carbon, should be sufficient. The refrigeration system for liquefaction and adsorption would have a capacity of about 2 tons. As an alternative, the use of molecular sieves to trap out the air was investigated, but such a bed would require at least 15,000 lb of molecular sieves and very long cycle times. A large-scale desublimer was also investigated and appears workable, but it would require some development effort before the design could proceed with confidence.
Date: October 1, 1979
Creator: Perona, J.J. & Watson, J.S.
Partner: UNT Libraries Government Documents Department

Analysis of mass transfer processes in geothermal power cycles utilizing direct contact heat exchange. Report of work, September 21, 1978 to September 30, 1979

Description: A computer program was developed which calculates the isobutane content of the spent brine and the liquid-vapor distribution of carbon dioxide and hydrogen sulfide throughout the components of a geothermal power plant using direct contact heat exchange. The program model assumes separate boiler and preheater vessels, with the preheater being a spray tower. The condenser model is a horizontal tube surface condenser with condensation on the outside. The program was written in Fortran language. The Fortran source deck consists of 976 cards. The program utilizes 320K for compilation and 72K for execution on an IBM 370/3031. Sample cases were run which illustrate the effects of salt concentration in the brine and isobutane-to-brine ratio on isobutane and noncondensible gas content of the spent brine.
Date: January 1, 1979
Creator: Knight, J.J. & Perona, J.J.
Partner: UNT Libraries Government Documents Department

DEMONSTRATION DISPOSAL OF HIGH-LEVEL RADIOACTIVE SOLIDS IN LYONS, KANSAS, SALT MINE: BACKGROUND AND PRELIMINARY DESIGN OF EXPERIMENTAL ASPECTS

Description: A demonstration of the disposal of high-level radioactive waste solids to be carried out in a salt mine at Lyons, Kansas, will have as its objectives: (1) the demonstration of required waste-handling equipment and techniques, (2) the determination of the stability of salt under the influence of heat and radiation, and (3) the collection of information on creep and plastic flow of salt which is needed for the design of an actual disposal facility. As presently conceived, 14 irradiated fuel assemblies from the Engineering Test Reactor will serve as a source of radiation in lieu of actual solidified wastes. The assemblies will be placed in a circular array of holes in the floor with one can in the center and other six cans located peripherally, spaced 5 ft on centers. During the course of the 2-year test, four sets of assemblies will be used to achieve a peak dose to the salt of about 8 x 10/sup 8/ rad and the temperature of the adjacent salt will be maintained at 200 deg C with electrical heaters. A second array, consisting only of heaters, will be operated as a control to determine those effects due solely to heat. In addition to the radioactive and control arrays, a ribpillar located between the two arrays will be heated electrically around its base to produce significant information on salt flow characteristics at elevated temperatures. (auth)
Date: January 10, 1964
Creator: Bradshaw, R.L.; Perona, J.J. & Blomeke, J.O.
Partner: UNT Libraries Government Documents Department

CALCULATED TRANSIENT PRESSURES DUE TO IMPULSE AND RAMP PERTURBATIONS TO VENTILATING SYSTEMS IN BUILDINGS 3019, 3026, 3508, AND 4507

Description: As part of a general hazard review survey conducted by the Chemical Technology Division of its facilities, transient pressures due to impulse and ramp perturbations to the cell ventilating systems of buildings 3019, 3026, and 4607 and the closed glove box system of 3508 were calculated. From the portions of the pressure curves above atmospheric pressure, volumes of gas outleakage were estimated; thus the amount of activity released can be calculated if an estimate of the activity concentration is available. The volumes of outleakage for all four ventilating systems were small for reasonable sizes of perturbations. For an impulse perturbation causing an instantaneous rise of +8.0 in- H/sub 2/0, the length of time above atmospheric pressure and estimated outleakages for PRFP cells in 3019 are 1.5 sec and 3.1 ft/sup 3/, respectively; for volatility cell 1 in 3019, 0.33 sec and 0.45 ft3; for cell A in 3026, 2.1 sec and 3.0 ft/sup 3/; for a glove box in 3508, 0.066 sec and 0.04 ft/sup 3/; and a cell in 4507, 0.26 sec and 0.03 ft/sup 3/. (auth)
Date: August 15, 1961
Creator: Perona, J.J.; Dunn, W.E. & Johnson, H.F.
Partner: UNT Libraries Government Documents Department

Vacuum sorption pumping at cryogenic temperatures of argon and oxygen on molecular sieves

Description: Cryosorption pumping is a method of excavating enclosed volumes by adsorbing gas on a deep bed of solid sorbent (typically a zeolite) at cryogenic temperatures. Modeling the dynamic behavior of these systems for air pumping requires information on two major constituents of air, oxygen and argon, which had not been previously studied, as well as data on a nonadsorbing specie, helium. Deep beds of Davison 4A molecular sieves were subjected to a metered flow of pure gas and the pressure history of the experiment was monitored, using computer data acquisition techniques. Particle size variations is the major variable in determining the mechanism of the process. The data acquired in the current study compare favorably with previous experiments. Previously developed models for the dynamic sorption behavior of deep beds under vacuum for two extreme conditions, micropore and micropore control were tested in this study. The sorption behavior of argon clearly fit into the category of macropore controlled sorption, indicating that these species are adsorbed primarily on the surface of the zeolite crystals, much like the theoretical and experimental results for N/sub 2/ cryosorption on the same sieves of Crabb. On the other hand oxygen sorption is most likely micropore controlled, and may be molded by the method of Praznick. 11 refs., 7 figs., 1 tab.
Date: January 1, 1988
Creator: Perona, J.J.; Gibson, M.R. & Byers, C.H.
Partner: UNT Libraries Government Documents Department

Low-pressure transfer operations

Description: Low-pressure transfer operations have been used or proposed which utilize sorption or condensation processes to induce flow. The applications include vacuum pumping, as well as transport of gases between vessels. The unsteady-state operation and the coupling of sorption/condensation with the flow rate produce interestingly complex behavior.
Date: January 1, 1982
Creator: Watson, J.S.; Fisher, P.W. & Perona, J.J.
Partner: UNT Libraries Government Documents Department

EVALUATION OF ULTIMATE DISPOSAL METHODS FOR LIQUID AND SOLID RADIOACTIVE WASTES. IV. SHIPMENT OF CALCINED SOLIDS

Description: The costs of shipping caleined Purex and Thorex wastes were calculated assuming the wastes were produced by a plant processing 1500 metric tons/year of U converter fuel at a burnup of 10,000 Mwd ton, 270 metric tons/year of Th converter fuel at 20,000 Mwd/ton. Calculations were made for Purex waste calcined in acidic and reacidified (after alkaline storage) forms and for Thorex waste calcined in acidic and reacidified forms and with constituents added for producing an acidic Thorex glass. Shipping casks of Fe, Pb, and U were considered at 25, 0.75, and 00/ lb. Casks were cylindrical in shape and up to 60 in. ID, which is large enough to contain four 24-in.-dia., nine 12in.- dia., or thirty six 6-in.-dia. cylinders of calcined waste. Cask weights ranged up to 100 tons. The cask design did not include liquid coolants or mechanical cooling equipment, and couriers were assumed not required. Minimum waste age prior to shipping because of temperature limitations ranged up to 11 years for acidic Purex with four 24-in.-dia. cylinders/cask. Rail freight rates of , , and /ton were assumed for distances of turn of the empty casks. Total costs were lowest in all cases for lead casks, and for 1000 mile round-trip shipments ranged from 5.5 x 10/sup -4/ mil1/kwh/sub e/ for acidic Purex waste at 30 years of age in casks containing four 24-in.-dia. cylinders to 1.6 x 10/sup -2/ mill/kwh/sub e/ for acidic Thorex at 0.33 years in casks containing four 6-in.- dia. or one 12-in.-dia. cylinders. Costs for 3000 mile roundtrip shipments were higher by factors of 2.0 to 2.4. (auth)
Date: October 18, 1962
Creator: Perona, J.J.; Bradshaw, R.L.; Blomeke, J.O. & Roberts, J.T.
Partner: UNT Libraries Government Documents Department

EVALUATION OF ULTIMATE DISPOSAL METHODS FOR LIQUID AND SOLID RADIOACTIVE WASTES. PART III. INTERIM STORAGE OF SOLIDIFIED WASTES

Description: The costs of interim storage of solidified Purex and Thorex wastes in water-filled canals were estimated as the third part of a study to evaluate, from the standpoint of econoNonemics and hazards, the various steps leading to and including the permanent disposal of highly radioactive liquid and solid wastes. The wastes were assumed to have been solidified following their production in a plant proccessed 1500 metric tons per year of uranium converter fuel it a burnup of 10,000 Mwd/ton and 270 tons/yr of thorium converter fuel at 20,000 Mwd/ton. Separate facilities where designed for the storage of the calcined wastes in the acid and reacified forms, and for the Thorex waste made into a glass. Consideration was given also to storage (in the same facilities) of the combinations acid Purex-acid Thorex and reacified Purex-reacidified Thorex wastes. Costs for interim storage times from 1 to 30 yr were computed for wastes decayed 120 days and 1, 3, and 10 yr at time of initial storage. Costs ranged from 1.5 x 10/sup -3/ mill/kwh, for 1-yr storage of calcined 10-yr-old acid Purex waste to 18 x 10/sup -3/ mill/kwh/sub e/ for 30-yr storage of calcined, reacified 120-day- old Thorex wastes. Costs of storage of the Purex and Thorex wastes together in the same facility ranged from 1.5 x 10/sup -3/ mill/kwh/sub e/ for 1-yr storage to 4.8 x 10/sup -3/ mill/kwh/sub e/ for 10-yr storage for the calcined acid wastes and from 1.8 x 10/sup -3/ to 6.3 x 10/sup -3/ wastes at time of storage was not a very significant factor, the costs for storage of 10-yr-decayed wastes being only 10 to 15% less than those for storage of the same wastes aged 120 days. Storage of acid wastes as solids was cheaper by factors of 2 to 2.7 than storage ...
Date: October 21, 1963
Creator: Blomeke, J O; Perona, J J; Weeren, H O & Bradshaw, T L
Partner: UNT Libraries Government Documents Department

EVALUATION OF ULTIMATE DISPOSAL METHODS FOR LIQUID AND SOLID RADIOACTIVE WASTES. V. EFFECTS OF FISSION PRODUCT REMOVAL ON COSTS OF WASTE MANAGEMENT

Description: In a study based on optimistic expectations of waste composition from future fission product separations processes, estimated costs for management of wastes from which 90 and 99% of all fission products were removed were from 70 to 80% of those for management of waste from which no fission products were removed. This cost difference is not believed to be sufficient to pay for the separation and final disposal of the fission products, which was not included in the waste management costs; hence, separation does not represent an economic route for waste management unless a substantial market for the fission products exists to pay most of the costs. As a basis for this study, it was assumed that after fission product removal the waste was identical to neutralized Purex waste in volume and composition of major ingredients. The sequential steps in the management of waste from processing 1500 metric tons per year of uranium converter fuel irradiated to 10,000 Mwd/ton were: interim storage of liquid waste, conversion to solids by pot calcination, interim storage of calcined solid waste, shipment of 1000 miles, and final disposal in a salt mine. Minimum-cost schemes were worked out involving optimum choices of interim liquid and solid storage times, diameter of the waste-calcination cylinder, and age at time of burial in the sa1t. Costs for wastes from which fission products were removed were 1east for calcination in 24-in.-dia. vessels, were not strongly affected by age, and fell in the range of 0.017 to 0.019 mill/kwh(e). The lowest cost for acid Purex waste without fission product removal was about 0.024 mill/kwh(e), obtained by using either 12- or 24-in.-dia. calcination vessels and buried in salt after allowing 30 years for decay of the fission products in the calcined wastes. These costs are equivalent to about 00 per ton ...
Date: June 26, 1963
Creator: Perona, J.J.; Blomeke, J.O.; Bradshaw, R.L. & Roberts, J.T.
Partner: UNT Libraries Government Documents Department

Sludge mobilization with submerged nozzles in horizontal cylindrical tanks

Description: The Melton Valley Storage Tanks (MVSTs) and the evaporator service tanks at the Oak Ridge National Laboratory (ORNL) are used for the collection and storage of liquid low-level waste (LLLW). Wastes collected in these tanks are typically acidic when generated and are neutralized with sodium hydroxide to protect the tanks from corrosion; however, the high pH of the solution causes the formation of insoluble compounds that precipitate. These precipitates formed a sludge layer approximately 0.6 to 1.2 m (2 to 4 ft) deep in the bottom of the tanks. The sludge in the MVSTs and the evaporator service tanks will eventually need to be removed from the tanks and treated for final disposal or transferred to another storage facility. The primary options for removing the sludge include single-point sluicing, use of a floating pump, robotic sluicing, and submerged-nozzle sluicing. The objectives of this study were to (1) evaluate the feasibility of submerged-nozzle sluicing in horizontal cylindrical tanks and (2) obtain experimental data to validate the TEMPEST (time-dependent, energy, momentun, pressure, equation solution in three dimensions) computer code.
Date: October 1995
Creator: Hylton, T. D.; Cummins, R. L.; Youngblood, E. L. & Perona, J. J.
Partner: UNT Libraries Government Documents Department

EVALUATION OF ULTIMATE DISPOSAL METHOD FOR LIQUID AND SOLID RADIOACTIVE WASTES. PART I. INTERIM LIQUID STORAGE

Description: As the first part of a study to evaluate the economics of the various steps leading to and including the permanent disposal of high-activity liquid and solid radioactive waste, costs of interim liquid storage of acid and alkaline Purex and Thorex wastes were estimated for storage times of 0.5 to 30 years. A 6- ton/day plant was assumed, processing 1500 tons/year of uranium converter fuel at a burnup of 10,000 Mwd/ton and 270 tons/year of thorium converter fuel at a burnup of 20,000 Mwd/ton. Tanks of Savannah River design were assumed, with stainless steel construction for acid wastes and mild steel construction for neutralized wastes. The operating cycle of each tank was assumed to consist of equal filling and emptying periods plus a full (or dead) period. With interim storage time defined as filling time plus full time, tank costs were minimum when full time was 40 to 70% of the interim storage time, using present worth considerations. For waste storage times of 0.5 to 30 years, costs ranged from 2.2 x 10/sup -3/ to 9.5 x 10/sup -3/ mill/kwh/sub e/ for acid wastes and from 1.7 x 10/sup -3/ to 5.1 x 10/sup -3/ mill/kwh/sub e/ for neutralized wastes. (auth)
Date: August 22, 1961
Creator: Bradshaw, R.L.; Perona, J.J.; Roberts, J.T. & Blomeke, J.O.
Partner: UNT Libraries Government Documents Department

EVALUATION OF ULTIMATE DISPOSAL METHODS FOR LIQUID AND SOLID RADIOACTIVE WASTES. PART II. CONVERSION TO SOLID BY POT CALCINATION

Description: The costs of pot calcination of Purex and Thorex wastes were calculated. The wastes were assumed produced by a plant processing 1500 ton/year of U converter fuel at a burnup of 10,000 Mwd/ton and 270 ton/year of Th converter fuel at 20,000 Mwd/ton. Costs were calculated for processing Purex waste in acidic and reacidified forms and for processing Thorex wastes in acidic and reacidified forms and with constituents added for producing an acidic Thorex glass. Calcination vessel designs were right circular cylinders similar to those used in engineering development studies. Costs were calculated for processing in 6-, 12-, and 24-in.-dia vessels with a fixed length of 10 ft. Vessel costs used, based on estimates from private industry, were calculated for wastes decayed 120 days and 1, 3, 10, and 30 years after reactor discharge prior to calcination. Aging had negligible effect on costs, except as it permitted larger diameter vessels to be used, because vessel and operating costs were much larger than capital costs in all cases. The lowest cost was 0.87 x 10/sup -2/ mill/kwh/sub e/ for processing acidic Purex and Thorex wastes in 24-in.-dia vessels, and the highest was 5.0 x 10/sup -2/ mill/kwh/sub e/ for processing reacidified Purex and Thorex wastes in 6-in.-dia vessels. About 7 years of interim liquid storage would be required before acidic Purex wastes could be processed in 24-in.-dia vessels. (auth)
Date: October 16, 1961
Creator: Perona, J.J.; Bradshaw, R.L.; Roberts, J.T. & Blomeke, J.O.
Partner: UNT Libraries Government Documents Department

Nitrogen oxide absorption into water and dilute nitric acid in an engineering-scale sieve-plate column with plates designed for high gas-liquid interfacial area

Description: The absorption of gaseous NO/sub x/ compounds into water and dilute HNO/sub 3/ was studied in a three-stage sieve-plate column with plates designed for high gas-liquid interfacial area. The performance of the column was measured while several operating parameters were varied. The results of the study indicate the importance of three mechanisms in the absorption of gaseous NO/sub x/ (NO/sub 2/ + 2N/sub 2/O4 + NO) compounds: (1) the absorption of NO/sub 2//sup */ (NO/sub 2/ + 2N/sub 2/O4) which results in production of liquid HNO/sub 3/ and HNO/sub 2/; (2) the dissociation of the liquid HNO/sub 2/ into HNO/sub 3/ and gaseous NO; and (3) the gas-phase oxidation of NO to NO/sub 2/. A mathematical model based on these mechanisms was developed and is presented to explain the observed phenomena.
Date: January 1, 1978
Creator: Counce, R.M.; Groenier, W.S.; Klein, J.A. & Perona, J.J.
Partner: UNT Libraries Government Documents Department

Treatability studies for decontamination of Melton Valley Storage Tank supernate

Description: Liquid low-level waste, primarily nitric acid contaminated with radionuclides and minor concentrations of organics and heavy metals, is neutralized with sodium hydroxide, concentrated by evaporation, and stored for processing and disposal. The evaporator concentrate separates into sludge and supernate phases upon cooling. The supernate is 4 to 5 mol/L sodium nitrate contaminated with soluble radionuclides, principally {sup 137}Cs, {sup 90}Sr, and {sup 14}C, while the sludge consists of precipitated carbonates and hydroxides of metals and transuranic elements. Methods for treatment and disposal of this waste are being developed. In studies to determine the feasibility of removing {sup 137}Cs from the supernates before solidification campaigns, batch sorption measurements were made from four simulated supernate solutions with four different samples of potassium hexacyanocobalt ferrate (KCCF). Cesium decontamination factors of 1 to 8 were obtained with different KCCF batches from a highly-salted supernate at pH 13. Decontamination factors as high as 50 were measured from supernates with lower salt content and pH, in fact, the pH had a greater effect than the solution composition on the decontamination factors. The decontamination factors were highest after 1 to 2 d of mixing and decreased with longer mixing times due to decomposition of the KCCF in the alkaline solution. The decontamination factors decreased with settling time and were lower for the same total contact time (mixing + settling) for the longer mixing times, indicating more rapid KCCF decomposition during mixing than during settling. There was no stratification of cesium in the tubes as the KCCF decomposed.
Date: August 1, 1992
Creator: Arnold, W.D.; Fowler, V.L.; Perona, J.J. & McTaggart, D.R.
Partner: UNT Libraries Government Documents Department

Evaporation studies on Oak Ridge National Laboratory liquid low-level waste

Description: Evaporation studies were performed with Melton Valley storage tank liquid low-level radioactive waste concentrate and with surrogates (nonradioactive) to determine the feasibility of a proposed out-of-tank-evaporation project. Bench-scale tests indicated that volume reductions ranging from 30 to 55% could be attained. Vendor-site tests were conducted (with surrogate waste forms) using a bench-scale single-stage, low-pressure (subatmospheric), low-temperature (120 to 173{degree}F) evaporator similar to units in operation at several nuclear facilities. Vendor tests were successful; a 30% volume reduction was attained with no crystallization of solids and no foaming, as would be expected from a high pH solution. No fouling of the heat exchanger surfaces occurred during these tests. It is projected that 52,000 to 120,000 gal of water could be evaporated from the supernate stored in the Melton and Bethel Valley liquid low-level radioactive waste (LLLW) storage tanks with this type of evaporator.
Date: March 1, 1993
Creator: Fowler, V. L. & Perona, J. J.
Partner: UNT Libraries Government Documents Department