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Technology of the light water reactor fuel cycle

Description: This essay presents elements of the processes used in the fuel cycle steps and gives an indication of the types of equipment used. The amounts of radioactivity released in normal operation of the processes are indicated and related to radiation doses. Types and costs of equipment or processes required to lower these radioactivity releases are in some cases suggested. Mining and milling, conversion of uranium concentrate to UF/sub 6/, uranium isotope separation, LWR fuel fabrication, fuel reprocessing, transportation, and waste management are covered in this essay. 40 figures, 34 tables. (DLC)
Date: January 1, 1979
Creator: Wymer, R.G.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department

DEVELOPMENT OF A METHOD FOR PREDICTING CONTINUOUS TRICKLE-TYPE COLUMN DISSOLVER BEHAVIOR FROM BATCH DATA

Description: A method for determining dissolver solution concentration as a function of length of column traversed for trickletype continuous dissolvers has been developed. An illustrative example is worked out for the mercury-catalyzed nitric acid dissolution of U-Al cast alloy slugs. The method employs a graphical integration of simple functions of dissolution rates and acid concentrations. These rates and concentrations are easily obtained from batch experiments made over the important ranges of concentration. (auth)
Date: May 14, 1954
Creator: Wymer, R.G.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department

CERENKOV RADIATION INTENSITY CALCULATIONS FOR Sr$sup 90$ AND Co$sup 60$ IN WATER

Description: A method for calculating Cherenkov radiation intensity from an initial electron energy distribution is presented. The Cherenkov radiation intensity from 1 curie of Sr/sup 90/ in secular equilibrium with Y/sup 90/ in water was calculated from the beta energy spectrum to illustrate the use of the method for a pure beta emitter. The Cherenkov radiation intensity from 1 curie of Co/sup 60/ in water was calculated from the Compton electron energy spectrum to illustrate the use of the method for a gamma emitter. The steps necessary to obtain the Compton electron energy spectrum from a gamma emitter are indicated. (auth)
Date: September 19, 1961
Creator: Wymer, R.G. & Biggers, R.E.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department

Thorium fuel cycle: a nuclear strategy and fuel recycle technology

Description: Use of thorium fuel cycles in thermal reactors appears to permit a moderate rate of introduction of fast breeder reactors into the nuclear economy and helps maintain a relatively low ratio of FBRs to thermal reactors in the future. To implement the benefits of thorium fuel cycles, however, will require fuel recycle research and development. Fuel recycle technology developed for uranium and plutonium cycles will be beneficial to thorium fuel cycle development; however, significant additional R and D is required to implement either the HEUTH or the DUTH cycles. The metal-clad reactors in general have relatively common generic technology development requirements, although there are significant differences between fast and thermal reactor fuel recycle needs. The thorium fuel recycle R and D requirements of HTGRs are more reactor-specific than of the other reactor types; however, some specific efforts will be required for all the different reactor types.
Date: January 1, 1978
Creator: Kasten, P.R.; Dahlberg, R.C. & Wymer, R.G.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department

Workshop on instrumentation and analyses for a nuclear fuel reprocessing hot pilot plant

Description: In order to assist in the study of instrumentation and analytical needs for reprocessing plants, a workshop addressing these needs was held at Oak Ridge National Laboratory from May 5 to 7, 1980. The purpose of the workshop was to incorporate the knowledge of chemistry and of advanced measurement techniques held by the nuclear and radiochemical community into ideas for improved and new plant designs for both process control and inventory and safeguards measurements. The workshop was athended by experts in nuclear and radiochemistry, in fuel recycle plant design, and in instrumentation and analysis. ORNL was a particularly appropriate place to hold the workshop since the Consolidated Fuel Reprocessing Program (CFRP) is centered there. Requirements for safeguarding the special nuclear materials involved in reprocessing, and for their timely measurement within the process, within the reprocessing facility, and at the facility boundaries are being studied. Because these requirements are becoming more numerous and stringent, attention is also being paid to the analytical requirements for these special nuclear materials and to methods for measuring the physical parameters of the systems containing them. In order to provide a focus for the consideration of the workshop participants, the Hot Experimental Facility (HEF) being designed conceptually by the CFRP was used as a basis for consideration and discussions.
Date: May 1, 1980
Creator: Babcock, S.M.; Feldman, M.J.; Wymer, R.G. & Hoffman, D.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department

Options for treating high-temperature gas-cooled reactor fuel for repository disposal

Description: This report describes the options that can reasonably be considered for disposal of high-temperature gas-cooled reactor (HTGR) fuel in a repository. The options include whole-block disposal, disposal with removal of graphite (either mechanically or by burning), and reprocessing of spent fuel to separate the fuel and fission products. The report summarizes what is known about the options without extensively projecting or analyzing actual performance of waste forms in a repository. The report also summarizes the processes involved in convert spent HTGR fuel into the various waste forms and projects relative schedules and costs for deployment of the various options. Fort St. Vrain Reactor fuel, which utilizes highly-enriched {sup 235}U (plus thorium) and is contained in a prismatic graphite block geometry, was used as the baseline for evaluation, but the major conclusions would not be significantly different for low- or medium-enriched {sup 235}U (without thorium) or for the German pebble-bed fuel. Future US HTGRs will be based on the Fort St. Vrain (FSV) fuel form. The whole block appears to be a satisfactory waste form for disposal in a repository and may perform better than light-water reactor (LWR) spent fuel. From the standpoint of process cost and schedule (not considering repository cost or value of fuel that might be recycled), the options are ranked as follows in order of increased cost and longer schedule to perform the option: (1) whole block, (2a) physical separation, (2b) chemical separation, and (3) complete chemical processing.
Date: February 1, 1992
Creator: Lotts, A.L.; Bond, W.D.; Forsberg, C.W.; Glass, R.W.; Harrington, F.E.; Micheals, G.E. et al.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department

Options for treating high-temperature gas-cooled reactor fuel for repository disposal

Description: This report describes the options that can reasonably be considered for disposal of high-temperature gas-cooled reactor (HTGR) fuel in a repository. The options include whole-block disposal, disposal with removal of graphite (either mechanically or by burning), and reprocessing of spent fuel to separate the fuel and fission products. The report summarizes what is known about the options without extensively projecting or analyzing actual performance of waste forms in a repository. The report also summarizes the processes involved in convert spent HTGR fuel into the various waste forms and projects relative schedules and costs for deployment of the various options. Fort St. Vrain Reactor fuel, which utilizes highly-enriched {sup 235}U (plus thorium) and is contained in a prismatic graphite block geometry, was used as the baseline for evaluation, but the major conclusions would not be significantly different for low- or medium-enriched {sup 235}U (without thorium) or for the German pebble-bed fuel. Future US HTGRs will be based on the Fort St. Vrain (FSV) fuel form. The whole block appears to be a satisfactory waste form for disposal in a repository and may perform better than light-water reactor (LWR) spent fuel. From the standpoint of process cost and schedule (not considering repository cost or value of fuel that might be recycled), the options are ranked as follows in order of increased cost and longer schedule to perform the option: (1) whole block, (2a) physical separation, (2b) chemical separation, and (3) complete chemical processing.
Date: February 1, 1992
Creator: Lotts, A. L.; Bond, W. D.; Forsberg, C. W.; Glass, R. W.; Harrington, F. E.; Micheals, G. E. et al.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department