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Impact of nuclear fuel cycle centers on shipping special nuclear materials and wastes

Description: The impact of integrated nuclear fuel cycle facilities on the transportation sector appears from this admittedly rather narrow study to be of only marginal significance. However, there are other factors which must be taken into account such as nuclear safeguards, economics, and radiological, ecological, institutional, and sociological impacts. Unless more clear-cut advantages can be shown by on-going studies for some of these other considerations, the regimentation and control of industry that would result from the imposition of the integrated fuel cycle facility concept probably could not be justified. (auth)
Date: January 1, 1975
Creator: Blomeke, J.O.
Partner: UNT Libraries Government Documents Department

Radioactive Waste Treatment and Disposal--a Bibliography of Unclassified Literature, Part 2

Description: This bibliography of the unclassified literature on radioactive waste treatment and disposal has been compiled from references published in the Nuclear Science Abstracts Volume 11, No. 11 (June 15, 1957), through Volume 12, No. 12 (June 30, 1958). The bibliography brings up to date a previous bibliography published as CF 57-8-138. (auth)
Date: July 22, 1958
Creator: Blomeke, J. O.
Partner: UNT Libraries Government Documents Department


Description: A consideration of the characteristics of fused salt-- ThF/sub 4/ solutions suitable for use in homogeneous reactors is presented, together with a brief survey of the literature pertaining to such solutions and a summary of the experimental work accomplished. (auth)
Date: July 19, 1951
Creator: Blomeke, J.O.
Partner: UNT Libraries Government Documents Department

Disposal of radioactive wastes

Description: Radioactive waste management and disposal requirements options available are discussed. The possibility of beneficial utilization of radioactive wastes is covered. Methods of interim storage of transuranium wastes are listed. Methods of shipment of low-level and high-level radioactive wastes are presented. Various methods of radioactive waste disposal are discussed. (DC)
Date: January 1, 1979
Creator: Blomeke, J.O.
Partner: UNT Libraries Government Documents Department

Origin, magnitude, and treatment of radioactive wastes

Description: Radioactive wastes in solid, liquid, and gaseous forms are generated wherever radioactive materials are handled. Unlike conventional industrial wastes, most of them are not amenable (within presently available technology) to treatment that can destroy their toxicity. They become innocuous only through natural decay, and many of the isotopes of concern decay so slowly that they must be isolated from the environment for hundreds of thousands of years. The primary objective of nuclear waste management is to protect man and his environment from these materials by providing containment by means that are available within present or near-future technology. The origin and nature of the wastes are reviewed, and the methods of treatment that are in use or that soon can be made available are described.
Date: January 1, 1979
Creator: Blomeke, J.O.
Partner: UNT Libraries Government Documents Department

Management of radioactive wastes

Description: This paper reviews the nature and characteristics of the radwastes, status of the work that has been done, options that are available in waste management, and plans and expectations for the near future. (DLC)
Date: November 1, 1979
Creator: Blomeke, J.O.
Partner: UNT Libraries Government Documents Department


Description: Disposal of wastes from the processing of solid fuel elements and from solid blanket elements is discussed. The subjects considered include extraction of uranium by several methods, the removal of element jackets, the treatment of uraxium -zirconium fuel elements, disposal into deep wells, the hydraulics of wells, thermal considerations of disposal aquifers regional hydrology, potential deep-well disposal areas in the U. S., aud the cost of disposal. (J.R.D.)
Date: June 13, 1957
Creator: de Laguna, W. & Blomeke, J.O.
Partner: UNT Libraries Government Documents Department


Description: Safe and economic methods of handling radioactive materials off-site are required for the successful operation of nuclear chemical plants. These occasions arise in the shipment of spent fuel, radioactive, isotopes, and liquid wastes. An unsolved problem exists in the development of techniques and sites for the final disposal of waste products. (auth)
Date: March 14, 1962
Creator: Blomeke, J.O. & Shappert, L.B.
Partner: UNT Libraries Government Documents Department

Actinide partitioning and transmutation program. Progress report, July 1--September 30, 1977

Description: In Purex process modifications, two cold runs with mixer-settlers were made on the extraction and stripping of ruthenium and zirconium without the presence of uranium. Efforts in actinide recovery from solids were directed toward the determination of dissolution parameters in various reagents for /sup 241/Am and /sup 239/Pu oxide mixtures, /sup 233/U oxide, /sup 237/Np oxide, /sup 244/Cm oxide, /sup 232/Th oxide, and PuO/sub 2/. Studies in americium-curium recovery with OPIX (oxalate precipitation and ion exchange), Talspeak, and cation exchange chromatography focused on the feasibility of forming oxalate precipitates in continuous systems, the effects of zirconium on Talspeak, and methods for removing solvent degradation products of the Talspeak system. In studies of americium-curium recovery using bidentate extractants, additional distribution coefficients for actinides and other key elements between reduced synthetic LWR waste solution and 30 percent dihexyl-N, N-diethyl-carbamylmethylene phosphonate in diisopropylbenzene were measured. Studies in the americium-curium recovery using inorganic ion exchange media to determine the pH dependence of lanthanide ion affinity for niobate, titanate, and zirconate ion exchange materials have been completed. A modified flowsheet for the extraction of uranium, neptunium, plutonium, americium, and curium from high-level liquid waste is presented. Evaluation of methods for measuring actinides from incinerator ash is continuing. A preliminary evaluation of methods for treatment of salt waste and waste waters was completed. In thermal reactor transmutation studies, waste actinides from an LWR lattice containing mixed uranium-plutonium assemblies were recycled in separate target assemblies. (LK)
Date: February 1, 1978
Creator: Tedder, D.W. & Blomeke, J.O. (comps.)
Partner: UNT Libraries Government Documents Department


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

Actinide partitioning-transmutation program final report. I. Overall assessment

Description: This report is concerned with an overall assessment of the feasibility of and incentives for partitioning (recovering) long-lived nuclides from fuel reprocessing and fuel refabrication plant radioactive wastes and transmuting them to shorter-lived or stable nuclides by neutron irradiation. The principal class of nuclides considered is the actinides, although a brief analysis is given of the partitioning and transmutation (P-T) of /sup 99/Tc and /sup 129/I. The results obtained in this program permit us to make a comparison of the impacts of waste management with and without actinide recovery and transmutation. Three major conclusions concerning technical feasibility can be drawn from the assessment: (1) actinide P-T is feasible, subject to the acceptability of fuels containing recycle actinides; (2) technetium P-T is feasible if satisfactory partitioning processes can be developed and satisfactory fuels identified (no studies have been made in this area); and (3) iodine P-T is marginally feasible at best because of the low transmutation rates, the high volatility, and the corrosiveness of iodine and iodine compounds. It was concluded on the basis of a very conservative repository risk analysis that there are no safety or cost incentives for actinide P-T. In fact, if nonradiological risks are included, the short-term risks of P-T exceed the long-term benefits integrated over a period of 1 million years. Incentives for technetium and iodine P-T exist only if extremely conservative long-term risk analyses are used. Further RD and D in support of P-T is not warranted.
Date: June 1, 1980
Creator: Croff, A.G.; Blomeke, J.O. & Finney, B.C.
Partner: UNT Libraries Government Documents Department


Description: As a means of reducing the quantity of radioactivity released to the environment by radioactive liquid waste discharges at ORNL, it is proposed that two 50,000-gal stainless steel storage tanks and a 600 gph stainless steel, submerged-coil evaporator be installed. The tanks, approximately 10 ft in diameter by 85 ft long, will be equipped with cooling coils attached to their outside surfaces for removal of a maximum of 300,000 Btu/hr of decay heat, and will be supported inside a concrete vauit for containment. The evaporator and a feed tank will be installed inside a cell shielded with 5 ft of concrete and will process mainly intermediate-level waste from the concrete tank farm, but will be able to evaporate high-level waste as well, if required. Two additional cells for the condenser and other off-gas equipment will be housed with the evaporator cell in a building with an operating area and sampling gallery. The capital cost of this installation is estimated to be 226,000. (auth)
Date: April 27, 1962
Creator: Weeren, H.O.; Blomeke, J.O. & Stockdale, W.G.
Partner: UNT Libraries Government Documents Department

Disposal of spent fuel

Description: Based on preliminary analyses, spent fuel assemblies are an acceptable form for waste disposal. The following studies appear necessary to bring our knowledge of spent fuel as a final disposal form to a level comparable with that of the solidified wastes from reprocessing: 1. A complete systems analysis is needed of spent fuel disposition from reactor discharge to final isolation in a repository. 2. Since it appears desirable to encase the spent fuel assembly in a metal canister, candidate materials for this container need to be studied. 3. It is highly likely that some ''filler'' material will be needed between the fuel elements and the can. 4. Leachability, stability, and waste-rock interaction studies should be carried out on the fuels. The major disadvantages of spent fuel as a disposal form are the lower maximum heat loading, 60 kW/acre versus 150 kW/acre for high-level waste from a reprocessing plant; the greater long-term potential hazard due to the larger quantities of plutonium and uranium introduced into a repository; and the possibility of criticality in case the repository is breached. The major advantages are the lower cost and increased near-term safety resulting from eliminating reprocessing and the treatment and handling of the wastes therefrom.
Date: January 1, 1978
Creator: Blomeke, J.O.; Ferguson, D.E. & Croff, A.G.
Partner: UNT Libraries Government Documents Department

Actinide Partitioning and Transmutation Program. Progress report, April 1--June 30, 1977

Description: Experimental work on the 16 tasks comprising the Actinide Partitioning and Transmutation Program was continued. Summaries of work are given on Purex Process modifications, actinide recovery, Am-Cm recovery, radiation effects on ion exchangers, LMFBR transmutation studies, thermal reactor transmutation studies, fuel cycle studies, and partitioning-transmutation evaluation. (JRD)
Date: October 1, 1977
Creator: Tedder, D. W. & Blomeke, J. O.
Partner: UNT Libraries Government Documents Department

Methods for separating actinides from reprocessing and refabrication plant wastes

Description: Chemical processing flowsheets have been developed to partition actinides from all actinide-bearing LWR fuel reprocessing and refabrication plant wastes. These wastes include high-activity-level liquids, scrap recovery liquors, HEPA filters and incinerator ashes, and chemical salt wastes such as sodium carbonate scrub solutions, detergent cleanup streams, and alkaline off-gas scrubber liquors. The separations processes that were adopted for this study are based on solvent extraction, cation exchange chromatography, and leaching with Ce/sup 4 +/-HNO/sub 3/ solution.
Date: January 1, 1979
Creator: Tedder, D.W.; Finney, B.C. & Blomeke, J.O.
Partner: UNT Libraries Government Documents Department


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


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


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