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Sample records for cimarron uranium fuel plant

  1. Decontamination and decommissioning of the Kerr-McGee Cimarron Plutonium Fuel Plant

    Energy Technology Data Exchange (ETDEWEB)

    1994-05-01

    This final report is a summary of the events that completes the decontamination and decommissioning of the Cimarron Corporation`s Mixed Oxides Fuel Plant (formally Sequoyah Fuels Corporation and formerly Kerr-McGee Nuclear Corporation - all three wholly owned subsidiaries of the Kerr-McGee Corporation). Included are details dealing with tooling and procedures for performing the unique tasks of disassembly decontamination and/or disposal. That material which could not be economically decontaminated was volume reduced by disassembly and/or compacted for disposal. The contaminated waste cleaning solutions were processed through filtration and ion exchange for release or solidified with cement for L.S.A. waste disposal. The L.S.A. waste was compacted, and stabilized as required in drums for burial in an approved burial facility. T.R.U. waste packaging and shipping was completed by the end of July 1987. This material was shipped to the Hanford, Washington site for disposal. The personnel protection and monitoring measures and procedures are discussed along with the results of exposure data of operating personnel. The shipping containers for both T.R.U. and L.S.A. waste are described. The results of the decommissioning operations are reported in six reports. The personnel protection and monitoring measures and procedures are contained and discussed along with the results of exposure data of operating personnel in this final report.

  2. Uranium Fuel Plant. Applicants environmental report

    International Nuclear Information System (INIS)

    The Uranium Fuel Plant, located at the Cimarron Facility, was constructed in 1964 with operations commencing in 1965 in accordance with License No. SNM-928, Docket No. 70-925. The plant has been in continuous operation since the issuance of the initial license and currently possesses contracts extending through 1978, for the production of nuclear fuels. The Uranium Plant is operated in conjunction with the Plutonium Facility, each sharing common utilities and sanitary wastes disposal systems. The operation has had little or no detrimental ecological impact on the area. For the operation of the Uranium Fuel Fabrication Plant, initial equipment provided for the production of UO2, UF4, uranium metal and recovery of scrap materials. In 1968, the plant was expanded by increasing the UO2 and pellet facilities by the installation of another complete production line for the production of fuel pellets. In 1969, fabrication facilities were added for the production of fuel elements. Equipment initially installed for the recovery of fully enriched scrap has not been used since the last work was done in 1970. Economically, the plant has benefited the Logan County area, with approximately 104 new jobs with an annual payroll of approximately $1.3 million. In addition, $142,000 is annually paid in taxes to state, local and federal governments, and local purchases amount to approximately $1.3 million. This was all in land that was previously used for pasture land, with a maximum value of approximately 37,000 dollars. Environmental effects of plant operation have been minimal. A monitoring and measurement program is maintained in order to ensure that the ecology of the immediate area is not affected by plant operations

  3. Radiation protection training at uranium hexafluoride and fuel fabrication plants

    International Nuclear Information System (INIS)

    This report provides general information and references useful for establishing or operating radiation safety training programs in plants that manufacture nuclear fuels, or process uranium compounds that are used in the manufacture of nuclear fuels. In addition to a brief summary of the principles of effective management of radiation safety training, the report also contains an appendix that provides a comprehensive checklist of scientific, safety, and management topics, from which appropriate topics may be selected in preparing training outlines for various job categories or tasks pertaining to the uranium nuclear fuels industry. The report is designed for use by radiation safety training professionals who have the experience to utilize the report to not only select the appropriate topics, but also to tailor the specific details and depth of coverage of each training session to match both employee and management needs of a particular industrial operation. 26 refs., 3 tabs

  4. Evaluation of bioassay program at uranium fuel fabrication plants

    International Nuclear Information System (INIS)

    Results of a comprehensive study of urinalysis, lung burden and personal air sample measurements for workers at a uranium fuel fabrication plant are presented. Correlations between measurements were found and regression models used to explain the relationship between lung burden, daily intakes and urinary excretions of uranium. Assuming the ICRP lung model, the lung burden histories of ten workers were used to estimate the amounts in each of the long-term compartments of the lung. Estimates of the half lives of each compartment and of the maximum relative contributions to the urine from each compartment are given. These values were then used to predict urinary excretions from the long-term compartments for workers at another fuel fabrication plant. The standard error of estimate compared well with the daily variation in urinary excretion. (author)

  5. Reprocessed uranium influence on clearance application in uranium fuel fabrication plant

    International Nuclear Information System (INIS)

    Clearance levels for uranium isotopes have been recently authorized in Japan. The measurement of those elements can be disregarded when the nominal of the element (D/C), expressed as (D/C)*, is less than 10-3, where D is the specific radioactivity concentration of nuclides, C is the clearance level of nuclides, and (D/C)* is defined as (D/C) divided by the highest value of (D/C)'s in the element constitutions of uranium waste. In this study, the concentration of nuclides in reprocessed uranium was evaluated using ORIGEN2 burnup analyses and the most appropriate decontamination factors for determining the (D/C)* values and their influence on clearance application in the uranium fuel fabrication plant. It was concluded that nuclides other than five isotopes, 232U, 234U, 235U, 236U, and 238U, can be disregarded with regards to clearance application in the uranium fabrication plant, regardless of operation conditions, whether the fuel is fabricated by receiving reprocessed uranium or not. (author)

  6. Spectrophotometric determination of uranium in liquid waste generated in Fuel Fabrication Plant

    International Nuclear Information System (INIS)

    During fabrication of uranium bearing nuclear fuels, liquid waste is being generated. The liquid waste contains impurities such as Ca, Na, Fe, Ni, Cr etc. The total dissolved solids (TDS) are high, upto 400 gram per litre (gpl). Study has been carried out for spectrophotometric determination of uranium in solution employing Arsenazo-III as metal indicator. The absorbance was measured at 655 nm. For U: Ca ratio 1:10 no interference was observed. For U:Ca ratio of 1:125, uranium concentration was reduced by ∼5%. The method can be applied for determination of uranium in liquid waste generated in fuel fabrication plant. (author)

  7. Uranium recovery from waste of the nuclear fuel cycle plants at IPEN-CNEN/SP, Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Freitas, Antonio A.; Ferreira, Joao C.; Zini, Josiane; Scapin, Marcos A.; Carvalho, Fatima Maria Sequeira de, E-mail: afreitas@ipen.b, E-mail: jcferrei@ipen.b, E-mail: jzini@ipen.b, E-mail: mascapin@ipen.b, E-mail: fatimamc@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    Sodium diuranate (DUS) is a uranium concentrate produced in monazite industry with 80% typical average grade of U{sup 3}O{sup 8}, containing sodium, silicon, phosphorus, thorium and rare earths as main impurities. Purification of such concentrate was achieved at the nuclear fuel cycle pilot plants of uranium at IPEN by nitric dissolution and uranium extraction into an organic phase using TBP/Varsol, while the aqueous phase retains impurities and a small quantity of non extracted uranium; both can be recovered later by precipitation with sodium hydroxide. Then the residual sodium diuranate goes to a long term storage at a safeguards deposit currently reaching 20 tonnes. This work shows how uranium separation and purification from such bulk waste can be achieved by ion exchange chromatography, aiming at decreased volume and cost of storage, minimization of environmental impacts and reduction of occupational doses. Additionally, the resulting purified uranium can be reused in nuclear fuel cycle.(author)

  8. Uranium recovery from waste of the nuclear fuel cycle plants at IPEN-CNEN/SP, Brazil

    International Nuclear Information System (INIS)

    Sodium diuranate (DUS) is a uranium concentrate produced in monazite industry with 80% typical average grade of U3O8, containing sodium, silicon, phosphorus, thorium and rare earths as main impurities. Purification of such concentrate was achieved at the nuclear fuel cycle pilot plants of uranium at IPEN by nitric dissolution and uranium extraction into an organic phase using TBP/Varsol, while the aqueous phase retains impurities and a small quantity of non extracted uranium; both can be recovered later by precipitation with sodium hydroxide. Then the residual sodium diuranate goes to a long term storage at a safeguards deposit currently reaching 20 tonnes. This work shows how uranium separation and purification from such bulk waste can be achieved by ion exchange chromatography, aiming at decreased volume and cost of storage, minimization of environmental impacts and reduction of occupational doses. Additionally, the resulting purified uranium can be reused in nuclear fuel cycle.(author)

  9. Automation of remote handling in uranium and mixed oxide fuel element fabrication plants

    International Nuclear Information System (INIS)

    The subject of the analyses are plants for the fabrication or uranium oxide and uranium-plutonium mixed oxide fuel elements. The reference basis of the paper is an overview of the state-of-the-art of manufacturing technologies with regard to automation and remote handling during fuel element fabrication in national and foreign plants, and in comparabel sectors of conventional technologies. Proceeding from ambient dose rates, residence times, and technical conditions or individual doses at typical work-places during fuel element fabrication, work processes are pointed out which, taking into account technical possibilities, should be given priority when automating, and technical solutions for it are sought. Advantages and disadvantages of such measures are outlined, and reduction of radiation exposure is shown (example: mixed oxide fuel fabrication plant at Hanau). (orig./HP)

  10. Uranium plutonium oxide fuels

    International Nuclear Information System (INIS)

    Uranium plutonium oxide is the principal fuel material for liquid metal fast breeder reactors (LMFBR's) throughout the world. Development of this material has been a reasonably straightforward evolution from the UO2 used routinely in the light water reactor (LWR's); but, because of the lower neutron capture cross sections and much lower coolant pressures in the sodium cooled LMFBR's, the fuel is operated to much higher discharge exposures than that of a LWR. A typical LMFBR fuel assembly is shown. Depending on the required power output and the configuration of the reactor, some 70 to 400 such fuel assemblies are clustered to form the core. There is a wide variation in cross section and length of the assemblies where the increasing size reflects a chronological increase in plant size and power output as well as considerations of decreasing the net fuel cycle cost. Design and performance characteristics are described

  11. Uranium's transformation from mineral to fuel bundles

    International Nuclear Information System (INIS)

    Uranium undergoes chemical transformation phases before it can be used in the nuclear power plant. In first phase uranium is transformed from mineral to yellow cake, in which uranium is in the form of U3O8. After that comes conversion (U3O8-UF6) and enrichment (0.7%-3% U-235). Finally, uranium is converted in fuel fabrication to uranium dioxide, UO2, and fuel pellets are made

  12. Environmental consequences of uranium atmospheric releases from fuel cycle facility: II. The atmospheric deposition of uranium and thorium on plants

    International Nuclear Information System (INIS)

    Uranium and thorium isotopes were measured in cypress leaves, wheat grains and lettuce taken in the surroundings of the uranium conversion facility of Malvési (South of France). The comparison of activity levels and activity ratios (namely 238U/232Th and 230Th/232Th) in plants with those in aerosols taken at this site and plants taken far from it shows that aerosols emitted by the nuclear site (uranium releases in the atmosphere by stacks and 230Th-rich particles emitted from artificial ponds collecting radioactive waste mud) accounts for the high activities recorded in the plant samples close to the site. The atmospheric deposition process onto the plants appears to be the dominant process in plant contamination. Dry deposition velocities of airborne uranium and thorium were measured as 4.6 × 10−3 and 5.0 × 10−3 m s−1, respectively. - Highlights: • Uranium and thorium were measured in plants near the uranium conversion facility. • Activity ratios show that emissions account for the high activities recorded in the plants. • The atmospheric deposition process appears to dominate in plant contamination. • Dry deposition velocities of airborne uranium and thorium were determined

  13. Environmental consequences of uranium atmospheric releases from fuel cycle facility: II. The atmospheric deposition of uranium and thorium on plants.

    Science.gov (United States)

    Pourcelot, L; Masson, O; Renaud, P; Cagnat, X; Boulet, B; Cariou, N; De Vismes-Ott, A

    2015-03-01

    Uranium and thorium isotopes were measured in cypress leaves, wheat grains and lettuce taken in the surroundings of the uranium conversion facility of Malvési (South of France). The comparison of activity levels and activity ratios (namely (238)U/(232)Th and (230)Th/(232)Th) in plants with those in aerosols taken at this site and plants taken far from it shows that aerosols emitted by the nuclear site (uranium releases in the atmosphere by stacks and (230)Th-rich particles emitted from artificial ponds collecting radioactive waste mud) accounts for the high activities recorded in the plant samples close to the site. The atmospheric deposition process onto the plants appears to be the dominant process in plant contamination. Dry deposition velocities of airborne uranium and thorium were measured as 4.6 × 10(-3) and 5.0 × 10(-3) m s(-1), respectively. PMID:25500060

  14. Criticality accident in uranium fuel processing plant. Questionnaires from Research Committee of Nuclear Safety

    International Nuclear Information System (INIS)

    The Research Committee of Nuclear Safety carried out a research on criticality accident at the JCO plant according to statement of president of the Japan Atomic Energy Society on October 8, 1999, of which results are planned to be summarized by the constitutions shown as follows, for a report on the 'Questionnaires of criticality accident in the Uranium Fuel Processing Plant of the JCO, Inc.': general criticality safety, fuel cycle and the JCO, Inc.; elucidation on progress and fact of accident; cause analysis and problem picking-up; proposals on improvement; and duty of the Society. Among them, on last two items, because of a conclusion to be required for members of the Society at discussions of the Committee, some questionnaires were send to more than 1800 of them on April 5, 2000 with name of chairman of the Committee. As results of the questionnaires contained proposals and opinions on a great numbers of fields, some key-words like words were found on a shape of repeating in most questionnaires. As they were thought to be very important nuclei in these two items, they were further largely classified to use for summarizing proposals and opinions on the questionnaires. This questionnaire had a big characteristic on the duty of the Society in comparison with those in the other organizations. (G.K.)

  15. Uranium speciation in plants

    International Nuclear Information System (INIS)

    Detailed knowledge of the nature of uranium complexes formed after the uptake by plants is an essential prerequisite to describe the migration behavior of uranium in the environment. This study focuses on the determination of uranium speciation after uptake of uranium by lupine plants. For the first time, time-resolved laser-induced fluorescence spectroscopy and X-ray absorption spectroscopy were used to determine the chemical speciation of uranium in plants. Differences were detected between the uranium speciation in the initial solution (hydroponic solution and pore water of soil) and inside the lupine plants. The oxidation state of uranium did not change and remained hexavalent after it was taken up by the lupine plants. The chemical speciation of uranium was identical in the roots, shoot axis, and leaves and was independent of the uranium speciation in the uptake solution. The results indicate that the uranium is predominantly bound as uranyl(VI) phosphate to the phosphoryl groups. Dandelions and lamb's lettuce showed uranium speciation identical to lupine plants. (orig.)

  16. Study of internal exposure to uranium compounds in fuel fabrication plants in Brazil

    International Nuclear Information System (INIS)

    The International Commission on Radiological Protection (ICRP) Publication 66 and Supporting Guidance 3) strongly recommends that specific information on lung retention parameters should be used in preference to default values wherever appropriate, for the derivation of effective doses and for bioassay interpretation of monitoring data. A group of 81 workers exposed to UO2 at the fuel fabrication facility in Brazil was selected to evaluate the committed effective dose. The workers were monitored for determination of uranium content in the urinary and faecal excretion. The contribution of intakes by ingestion and inhalation were assessed on the basis of the ratios of urinary to fecal excretion. For the selected workers it was concluded that inhalation dominated intake. According to ICRP 66, uranium oxide is classified as insoluble Type S compound. The ICRP Supporting Guidance 3 and some recent studies have recommended specific lung retention parameters to UO2. The solubility parameters of the uranium oxide compound handled by the workers at the fuel fabrication facility in Brazil was evaluated on the basis of the ratios of urinary to fecal excretion. Excretion data were corrected for dietary intakes. This paper will discuss the application of lung retention parameters recommended by the ICRP models to these data and also the dependence of the effective committed dose on the lung retention parameters. It will also discuss the problems in the interpretation of monitoring results, when the worker is exposed to several uranium compounds of different solubilities. (author)

  17. Individual monitoring of internal exposure to uranium oxides in two fuel fabrication plants

    International Nuclear Information System (INIS)

    Individual monitoring of personal exposure to inhalation of uranium oxides throughout the manufacture of fuel for pressurized water reactor (PWR) includes lung gamma-spectrometry, fecal analysis and urine analysis. Examination of the results shows the following: internal exposure is the consequence of repeated intake incidents as revealed by early peaks of urinary and particularly fecal elimination; a shift is often observed with the results of aerosol concentration measured through air collectors; the measured variations of uranium lung incorporations are relatively fast (apparent mean period 165 d). Correct evaluation of the effective dose equivalent from inhalation requires further information concerning the aerosol size distribution at work stations, the physico-chemical characteristics of the product leading to an estimate of its actual biological solubility, and the measurement of the fraction of aerosol liable to intake with an individual portable collector

  18. Compucea: A high performance analysis procedure for timely On-site Uranium Accountancy Verification in Leu Fuel Fabrication Plants

    International Nuclear Information System (INIS)

    COMPUCEA (Combined Procedure for Uranium Concentration and Enrichment Assay) is used for on-site analytical measurements in support of joint EURATOM-IAEA inspections during physical inventory verification (PIV) campaigns in European Low-Enriched Uranium (Leu) fuel fabrication plants. The analytical technique involves the accurate determination of the uranium element content by energy-dispersive X-ray absorption edge spectrometry (L-edge densitometry) and of the 235 U enrichment by gamma spectrometry with a LaBr3(Ce) detector. For evaluation of the LaBr3 spectra a modified version of the NaIGEM code is used, which has recently been adapted to handle the presence of reprocessed uranium. This paper describes the technique, setup and calibration procedure of the instrument. Results from PIV campaigns in 2007 and 2008 are presented, which demonstrate the performance of the technique. First results obtained with a sandwich detector configuration for enhanced detection efficiency of the passive gamma spectrometry are discussed.

  19. Uranium resources, production and fuel fabrication

    International Nuclear Information System (INIS)

    Almost all the known disseminated and vein-type uranium deposits in India are located in the Precambrian igneous and metamorphic complexes in the Peninsular Shield; the most significant reserves occur in the Singhbhum Thrust Belt of Bihar. Adequate resources of uranium to meet the country's fuel requirements for the nuclear power programme have been established. The Uranium Corporation of India has been operating commercially an underground uranium mine and a mill at Jaduguda (Bihar) since 1968. The uranium ore body is mined by the cut-and-fill method. The present mine workings, 530 m below ground level, comprise many innovative features, namely, a tower-mounted Koepe winder system, skip-loading with an underground crushing system, concrete headframe, etc. Surveillance, control and monitoring systems, especially required in the mining of low grade uranium ores, have been successfully introduced. The uranium mill adjacent to the mine uses the acid leach and ion-exchange processes of recovery. The effluents are suitably treated in a specially designed tailings pond. Other accessory economic minerals, namely chalcopyrite, molybdenite and magnetite, are profitably recovered as by-products. Fuel fabrication commenced in India with the manufacture of aluminium-clad metallic uranium fuel for the CIR reactor. Power reactor oxide fuel manufacture has been carried out initially at Trombay for the Rajasthan Power Reactor I (RAPP-I). For transferring the technology developed, industrial-scale plants have been set up in the Nuclear Fuel Complex (NFC) at Hyderabad for the manufacture of zirconium-clad natural uranium fuel for PHWRs and low enrichment uranium fuel for the BWR Tarapur Power Station

  20. Transfer of Plutonium-Uranium Extraction Plant and N Reactor irradiated fuel for storage at the 105-KE and 105-KW fuel storage basins, Hanford Site, Richland Washington

    International Nuclear Information System (INIS)

    The U.S. Department of Energy (DOE) needs to remove irradiated fuel from the Plutonium-Uranium Extraction (PUREX) Plant and N Reactor at the Hanford Site, Richland, Washington, to stabilize the facilities in preparation for decontamination and decommissioning (D ampersand D) and to reduce the cost of maintaining the facilities prior to D ampersand D. DOE is proposing to transfer approximately 3.9 metric tons (4.3 short tons) of unprocessed irradiated fuel, by rail, from the PUREX Plant in the 200 East Area and the 105 N Reactor (N Reactor) fuel storage basin in the 100 N Area, to the 105-KE and 105-KW fuel storage basins (K Basins) in the 100 K Area. The fuel would be placed in storage at the K Basins, along with fuel presently stored, and would be dispositioned in the same manner as the other existing irradiated fuel inventory stored in the K Basins. The fuel transfer to the K Basins would consolidate storage of fuels irradiated at N Reactor and the Single Pass Reactors. Approximately 2.9 metric tons (3.2 short tons) of single-pass production reactor, aluminum clad (AC) irradiated fuel in four fuel baskets have been placed into four overpack buckets and stored in the PUREX Plant canyon storage basin to await shipment. In addition, about 0.5 metric tons (0.6 short tons) of zircaloy clad (ZC) and a few AC irradiated fuel elements have been recovered from the PUREX dissolver cell floors, placed in wet fuel canisters, and stored on the canyon deck. A small quantity of ZC fuel, in the form of fuel fragments and chips, is suspected to be in the sludge at the bottom of N Reactor's fuel storage basin. As part of the required stabilization activities at N Reactor, this sludge would be removed from the basin and any identifiable pieces of fuel elements would be recovered, placed in open canisters, and stored in lead lined casks in the storage basin to await shipment. A maximum of 0.5 metric tons (0.6 short tons) of fuel pieces is expected to be recovered

  1. Chemwes Uranium Plant

    International Nuclear Information System (INIS)

    The Chemwes Uranium Plant is located in an area which is underlain to a major extent by pinnacled dolomite. It was decided to adopt a replacement fill for support of light structures in preference to alternatives such as the installation of piles or 'bridging' between pinnacles. The 3 m thick soil 'raft' resulting from the fill replacement technique made it possible to support all but a very small number of foundations upon shallow spread footings or raft slabs. This article describes a replacement fill for support of light structures at the Chemwes Uranium Plant

  2. Decommissioning of uranium conversion plant

    International Nuclear Information System (INIS)

    Since about 20 years have passed after the construction of the uranium conversion plant, most equipments installed have worn out. Liquid wastes stored in lagoons which were generated during the operation of this plant are needed to be treated safely. Therefore, the decommissioning project on the uranium conversion plant was started from 2001. This study is a preliminary step for the decommissioning of the uranium conversion plant. It was reviewed on the plant status overall, especially facility descriptions and operational histories for the installations located inside and outside of the plant and methods of decontamination and of dismantling to the contamination conditions. And some proper options on each main object was proposed

  3. Uranium hexafluoride production plant decommissioning

    International Nuclear Information System (INIS)

    The Institute of Energetic and Nuclear Research - IPEN is a research and development institution, located in a densely populated area, in the city of Sao Paulo. The nuclear fuel cycle was developed from the Yellow Cake to the enrichment and reconversion at IPEN. After this phase, all the technology was transferred to private enterprises and to the Brazilian Navy (CTM/SP). Some plants of the fuel cycle were at semi-industrial level, with a production over 20 kg/h. As a research institute, IPEN accomplished its function of the fuel cycle, developing and transferring technology. With the necessity of space for the implementation of new projects, the uranium hexafluoride (UF6) production plant was chosen, since it had been idle for many years and presented potential leaking risks, which could cause environmental aggression and serious accidents. This plant decommission required accurate planning, as this work had not been carried out in Brazil before, for this type of facility, and there were major risks involving gaseous hydrogen fluoride aqueous solution of hydrofluoric acid (HF) both highly corrosive. Evaluations were performed and special equipment was developed, aiming to prevent leaking and avoid accidents. During the decommissioning work, the CNEN safety standards were obeyed for the whole operation. The environmental impact was calculated, showing to be not relevant.The radiation doses, after the work, were within the limits for the public and the area was released for new projects. (author)

  4. Radiological safety aspects of uranium fuel fabrication facilities at Nuclear Fuel Complex, Hyderabad

    International Nuclear Information System (INIS)

    The Health Physics Division of the Bhabha Atomic Research Centre is operating a Health Physics Unit at Nuclear Fuel Complex, Hyderabad which carried out radiological, industrial hygiene and environmental surveillances. Nuclear Fuel Complex has two batteries of plants - one for natural UO2 fuel bundles for Pressurised Heavy Water Reactors (PHWRs) and the other for enriched UO2 fuel assemblies for Boiling Water Reactor (BWR) in the country. For natural UO2 fuel the Uranium Oxide Plant (UOP) converts magnesium diuranate to UO2 powder. The Ceramic Fuel Fabrication Plant (CFFP) processes the UO2 powder to dense sintered UO2 pellets and further to fuel assemblies for PHWR. The Enriched Uranium Oxide Plant (EUOP) starts with uranium hexa-fluoride and converts to UO2 powder and Enriched Fuel Fabrication Plant (EFFP) processes the UO2 powder to sintered pellets and fuel assemblies for BWR

  5. Vaal Reefs South uranium plant

    International Nuclear Information System (INIS)

    The Vaal Reefs mining complex, part of the Anglo American Corporation, is the largest gold and uranium producing complex in the world, being South Africa's principal producer, accounting for about a quarter of the country's uranium production. Vaal Reefs South uranium plant in the Orkney district was recently officially opened by Dr AJA Roux, the retiring president of the Atomic Energy Board and chairman of the Uranium Enrichment Corporation and will increase the country's uranium production. In the field of technology, and particularly processing technology, South Africa has shown the world unprecedented technology achievement in the field of uranium extraction from low grade ores and the development of the unique uranium enrichment process. New technical innovations that have been incorporated in this new plant are discussed

  6. Rapid determination of trace uranium in liquid wastes from spent nuclear-fuel reprocessing plants. Using on-line solid-phase extraction/electrochemical detection

    International Nuclear Information System (INIS)

    An on-line analysis system using a solid-phase extraction column coupled to electrochemical detection has been developed for the rapid determination of small amounts of uranium in liquid waste samples of spent nuclear-fuel reprocessing plants. A sample solution with a concentration of 3 M HNO3 was loaded onto a column: packed with U/TEVA resin. The interference elements were rinsed by passing 3 M HNO3 through the column. The adsorbed uranium was eluted with 0.1 M HNO3. The eluate was directly introduced into a flow-electrolysis cell. The reduction current of U(VI)→U(V) was monitored and recorded. The uranium concentration was calculated from the relation between the peak current and the concentration of the standard uranium solution. The result of five repeated analyses for a standard solution containing 2.5 μg (0.1 mL at 25 μg mL-1) of uranium was found to be 2.5 ± 0.025 μg (mean ±1σ). The detection limit calculated from 3-times the standard deviation of the background current was 56 ng. The analysis time required for one sample was within 5 min. The recoveries of uranium in actual nuclear waste reprocessing solutions were 92-112%. (author)

  7. Decommissioning of an uranium hexafluoride pilot plant

    International Nuclear Information System (INIS)

    The Institute of Nuclear and Energetic Researches has completed fifty years of operation, belongs to the National Commission for Nuclear Energy, it is situated inside the city of Sao Paulo. The IPEN-CNEN/SP is a Brazilian reference in the nuclear fuel cycle, researches in this field began in 1970, having dominance in the cycle steps from Yellow Cake to Uranium Hexafluoride technology. The plant of Uranium Hexafluoride produced 35 metric tonnes of this gas by year, had been closed in 1992, due to domain and total transference of know-how for industrial scale, demand of new facilities for the improvement of recent researches projects. The Institute initiates decommissioning in 2002. Then, the Uranium Hexafluoride pilot plant, no doubt the most important unit of the fuel cycle installed at IPEN-CNEN/SP, beginning decommissioning and dismantlement (D and D) in 2005. Such D and D strategies, planning, assessment and execution are described, presented and evaluated in this paper. (author)

  8. Simulation study for purification, recovery of plutonium and uranium from plant streams of Fast Reactor Fuel Reprocessing Plant

    International Nuclear Information System (INIS)

    A method for removal of plutonium from the lean organic streams obtained after co-stripping of uranium -plutonium was developed. Plutonium from lean organic phase was stripped using U4+/hydrazine as the stripping agent. The effect of concentrations of stripping agent U4+ and feed Pu concentration in the lean organic phase was studied. Lean organic phases having higher plutonium concentration require three stages of stripping to bring plutonium concentration 4+ stabilized by hydrazine reduces Pu (IV) to Pu (III) thereby stripping plutonium from the organic phase. The non-extractability of Pu (III) by TBP was utilized for development of flow sheet for obtaining a uranium product lean of plutonium for ease of handling. (author)

  9. Field test of short-notice random inspections for inventory-change verification at a low-enriched-uranium fuel-fabrication plant

    International Nuclear Information System (INIS)

    An approach of short-notice random inspections (SNRIs) for inventory-change verification can enhance the effectiveness and efficiency of international safeguards at natural or low-enriched uranium (LEU) fuel fabrication plants. According to this approach, the plant operator declares the contents of nuclear material items before knowing if an inspection will occur to verify them. Additionally, items about which declarations are newly made should remain available for verification for an agreed time. Then a statistical inference can be made from verification results for items verified during SNRIs to the entire populations, i.e. the entire strata, even if inspectors were not present when many items were received or produced. A six-month field test of the feasibility of such SNRIs took place at the Westinghouse Electric Corporation Commercial Nuclear Fuel Division during 1993. Westinghouse personnel made daily declarations about both feed and product items, uranium hexafluoride cylinders and finished fuel assemblies, using a custom-designed computer ''mailbox''. Safeguards inspectors from the IAEA conducted eight SNRIs to verify these declarations. They arrived unannounced at the plant, in most cases immediately after travel from Canada, where the IAEA maintains a regional office. Items from both strata were verified during the SNRIs by meant of nondestructive assay equipment

  10. Uranium dioxide Caramel fuel

    International Nuclear Information System (INIS)

    The work performed in France on Caramel fuels for research reactors reflects the reality of a program based on non proliferation criteria, as they have already appeared several years ago. This work actually includes the following different aspects: identification of the non proliferation criterion defining this action; determination of the economical and technical goals to be reached; realization of research and development studies finalized in a full scale demonstration; transposition to an industrial and commercial level

  11. Demand of natural uranium to satisfy the requirements of nuclear fuel of new nuclear power plants in Mexico

    International Nuclear Information System (INIS)

    Due to the expectation of that in Mexico new plants of nuclear energy could be installed, turns out from the interest to evaluate the uranium requirements to operate those plants and to also evaluate if the existing reserves in the country could be sufficient to satisfy that demand. Three different scenes from nuclear power plant expansion for the country are postulated here that are desirable for the diversification of generation technologies. The first scene considers a growth in the generation by nuclear means of two reactors of type ABWR that could enter operation by years 2015 and 2020, in the second considers the installation of four reactors but as of 2015 and new every 5 years, in the scene of high growth considers the installation of 6 reactors of the same type that in the other scenes, settling one every three years as of 2015. The results indicate that the uranium reserves could be sufficient to only maintain in operation to one of the reactors proposed by the time of their useful life. (Author)

  12. Nuclear fuel control in fuel fabrication plants

    International Nuclear Information System (INIS)

    The basic control problems of measuring uranium and of the environment inside and outside nuclear fuel fabrication plants are reviewed, excluding criticality prevention in case of submergence. The occurrence of loss scraps in fabrication and scrap-recycling, the measuring error, the uranium going cut of the system, the confirmation of the presence of lost uranium and the requirement of the measurement control for safeguard make the measurement control very complicated. The establishment of MBA (material balance area) and ICA (item control area) can make clearer the control of inventories, the control of loss scraps and the control of measuring points. Besides the above basic points, the following points are to be taken into account: 1) the method of confirmation of inventories, 2) the introduction of reliable NDT instruments for the rapid check system for enrichment and amount of uranium, 3) the introduction of real time system, and 4) the clarification of MUF analysis and its application to the reliability check of measurement control system. The environment control includes the controls of the uranium concentration in factory atmosphere, the surface contamination, the space dose rate, the uranium concentration in air and water discharged from factories, and the uranium in liquid wastes. The future problems are the practical restudy of measurement control under NPT, the definite plan of burglary protection and the realization of the disposal of solid wastes. (Iwakiri, K.)

  13. Status of Uranium Conversion Plant Decommissioning

    International Nuclear Information System (INIS)

    KAERI (Korea Atomic Energy Research Institute) constructed a pilot plant for the uranium conversion process for the development of the technologies and the localization of nuclear fuels for HWR (heavy water reactor) in 1982. The final product of the plant was a UO2 powder of ceramic grade for HWR and its capacity was 100 ton-U/year. After that, a part of the AUC (Ammonium Uranyl Carbonate) process was added and the process was improved for automatic operation. 320 tons of UO2 powder was produced and supplied to the fabrication plant at KAERI for the fuel of the Wolsong-1 CANDU (Canadian deuterium uranium) reactor. The conversion plant has building area of 2916 m2 and two main conversion processes. ADU (Ammonium Di-Uranate) and AUC process are installed in the backside and the front side of the building, respectively. Conversion plant has two lagoons, which is to store all wastes generated from the plant operation. Sludge wastes stored 150m3 and 100m3 in Lagoon 1 and 2, respectively. Main compounds of sludge are ammonium nitrate, sodium nitrate, calcium nitrate, and calcium carbonate. In early 1992, it was determined that the plant operation would be stopped due to a much higher production cost than that of the international market. The conversion plant has been shutdown and minimally maintained for the prevention of contamination by deterioration of the equipment and the lagoon

  14. Fuel powder production from ductile uranium alloys

    International Nuclear Information System (INIS)

    Metallic uranium alloys are candidate materials for use as the fuel phase in very-high-density LEU dispersion fuels. These ductile alloys cannot be converted to powder form by the processes routinely used for oxides or intermetallics. Three methods of powder production from uranium alloys have been investigated within the US-RERTR program. These processes are grinding, cryogenic milling, and hydride-dehydride. In addition, a gas atomization process was investigated using gold as a surrogate for uranium. (author)

  15. Manufacture of uranium compounds for research reactors fuel elements. Participation of the UCPP (Uranium compound production plant) in the Egyptian project

    International Nuclear Information System (INIS)

    UCPP is an international qualified supplier of U3O8 with up to 20 % enrichment in U-235. The characteristics of this powder are those specified for fuel plates manufacture for test reactors. This paper describes the works performed in the plant since its beginning, emphasising those undertaken during the last years. The transference of U3O8 manufacturing technology to INVAP SE, the enterprise that installed a plant of similar characteristics in the Arabian Republic of Egypt, is especially described. (author)

  16. Nuclear fuel cycle based on thorium and uranium-233

    International Nuclear Information System (INIS)

    The analysis of activities carried out in this country and abroad on a complex solution of principal problems of nuclear power advance. Demonstration of the potentiality of the above problems solution on the basis of conventional reactor plant development (light water cooled reactors and BN-type fast reactors) within the framework of nuclear fuel cycle using uranium-235, plutonium and uranium-233. 28 refs.; 1 fig.; 8 tabs

  17. On line spectrophotometry with optical fibers. Application to uranium-plutonium separation in a spent fuel reprocessing plant

    International Nuclear Information System (INIS)

    Optimization of mixer-settler operation for uranium-plutonium separation in the Purex process can be obtained by remote spectrophotometry with optical fibers. Data acquisition on uranium VI, uranium IV and plutonium III is examined in function of acidity and nitrate content of the solution. Principles for on line multicomponent monitoring and mathematical modelization of the measurements are described

  18. Uranium uptake by hydroponically cultivated crop plants

    International Nuclear Information System (INIS)

    Hydroponicaly cultivated plants were grown on medium containing uranium. The appropriate concentrations of uranium for the experiments were selected on the basis of a standard ecotoxicity test. The most sensitive plant species was determined to be Lactuca sativa with an EC50 value about 0.1 mM. Cucumis sativa represented the most resistant plant to uranium (EC50 = 0.71 mM). Therefore, we used the uranium in a concentration range from 0.1 to 1 mM. Twenty different plant species were tested in hydroponic solution supplemented by 0.1 mM or 0.5 mM uranium concentration. The uranium accumulation of these plants varied from 0.16 mg/g DW to 0.011 mg/g DW. The highest uranium uptake was determined for Zea mays and the lowest for Arabidopsis thaliana. The amount of accumulated uranium was strongly influenced by uranium concentration in the cultivation medium. Autoradiography showed that uranium is mainly localized in the root system of the plants tested. Additional experiments demonstrated the possibility of influencing the uranium uptake from the cultivation medium by amendments. Tartaric acid was able to increase uranium uptake by Brassica oleracea and Sinapis alba up to 2.8 times or 1.9 times, respectively. Phosphate deficiency increased uranium uptake up to 4.5 times or 3.9 times, respectively, by Brassica oleracea and S. alba. In the case of deficiency of iron or presence of cadmium ions we did not find any increase in uranium accumulation. - Highlights: → The uranium accumulation in twenty different plant species varied from 0.160 to 0.011 mg/g DW. → Uranium is mainly localized in the root system. → Tartaric acid was able to increase uranium uptake by Brassica oleracea and Sinapis alba. → The phosphates deficiency increase the uranium uptake.

  19. Uranium uptake by hydroponically cultivated crop plants

    Energy Technology Data Exchange (ETDEWEB)

    Soudek, Petr; Petrova, Sarka [Laboratory of Plant Biotechnologies, Joint Laboratory of Institute of Experimental Botany AS CR, v.v.i. and Crop Research Institute, v.v.i., Rozvojova 263, 162 05 Prague 6 (Czech Republic); Benesova, Dagmar [Laboratory of Plant Biotechnologies, Joint Laboratory of Institute of Experimental Botany AS CR, v.v.i. and Crop Research Institute, v.v.i., Rozvojova 263, 162 05 Prague 6 (Czech Republic); Faculty of Environment Technology, Institute of Chemical Technology, Technicka 5, 166 28 Prague 6 (Czech Republic); Dvorakova, Marcela [Laboratory of Plant Biotechnologies, Joint Laboratory of Institute of Experimental Botany AS CR, v.v.i. and Crop Research Institute, v.v.i., Rozvojova 263, 162 05 Prague 6 (Czech Republic); Vanek, Tomas, E-mail: vanek@ueb.cas.cz [Laboratory of Plant Biotechnologies, Joint Laboratory of Institute of Experimental Botany AS CR, v.v.i. and Crop Research Institute, v.v.i., Rozvojova 263, 162 05 Prague 6 (Czech Republic)

    2011-06-15

    Hydroponicaly cultivated plants were grown on medium containing uranium. The appropriate concentrations of uranium for the experiments were selected on the basis of a standard ecotoxicity test. The most sensitive plant species was determined to be Lactuca sativa with an EC{sub 50} value about 0.1 mM. Cucumis sativa represented the most resistant plant to uranium (EC{sub 50} = 0.71 mM). Therefore, we used the uranium in a concentration range from 0.1 to 1 mM. Twenty different plant species were tested in hydroponic solution supplemented by 0.1 mM or 0.5 mM uranium concentration. The uranium accumulation of these plants varied from 0.16 mg/g DW to 0.011 mg/g DW. The highest uranium uptake was determined for Zea mays and the lowest for Arabidopsis thaliana. The amount of accumulated uranium was strongly influenced by uranium concentration in the cultivation medium. Autoradiography showed that uranium is mainly localized in the root system of the plants tested. Additional experiments demonstrated the possibility of influencing the uranium uptake from the cultivation medium by amendments. Tartaric acid was able to increase uranium uptake by Brassica oleracea and Sinapis alba up to 2.8 times or 1.9 times, respectively. Phosphate deficiency increased uranium uptake up to 4.5 times or 3.9 times, respectively, by Brassica oleracea and S. alba. In the case of deficiency of iron or presence of cadmium ions we did not find any increase in uranium accumulation. - Highlights: > The uranium accumulation in twenty different plant species varied from 0.160 to 0.011 mg/g DW. > Uranium is mainly localized in the root system. > Tartaric acid was able to increase uranium uptake by Brassica oleracea and Sinapis alba. > The phosphates deficiency increase the uranium uptake.

  20. Innovations over old plant techniques in Jaduguda Uranium Mill expansion

    International Nuclear Information System (INIS)

    India's first Uranium Mines and Mills was commissioned at Jaduguda in 1968. The plant's flowsheet was developed at BARC after extensive tests, for extraction of uranium as yellow cake from the ore. The designed capacity of the process plant was initially 1000 MT/day of ore treatment supplied from nearby mines. Subsequently, due to growing demand of uranium fuel, opening of Bhatin mines and setting up of three plants for recovery of uranium mineral from copper tailings of Hindustan Copper Ltd. was perceived. The capacity of the Jaduguda Plant was increased to 1400 MT/day in 1987 to meet this requirement. A new mine at Narwapahar is under development which will necessitate augmentation of the capacity of the Jaduguda plant by 700 MT/day. Major changes are contemplated in equipment selection for the expansion besides incorporation of a high degree of automation based on microprocessor technology which are discussed in this paper. (author)

  1. Analysis of causes of criticality accidents at nuclear fuel processing facilities in foreign countries. Similarities to the criticality accident at JCO's uranium processing plant

    International Nuclear Information System (INIS)

    On September 30, 1999, a criticality accident occurred at the JCO's uranium processing plant, which resulted in the first nuclear accident involving a fatality, in Japan, and forced the residents in the vicinity of the site to be evacuated and be sheltered indoors. Before the JCO accident, 21 criticality accidents have been reported at nuclear fuel processing facilities in foreign countries. The present paper describes the overall trends observed in the 21 accidents and discusses the sequences and causes of the accidents analyzed in terms of similarities to the JCO accident. Almost all of them occurred with the uranium or plutonium solution and in vessels/tanks with unfavorable geometry. In some cases, the problems similar to those observed in the JCO accident were identified: violations of procedures and/or technical specifications for improving work efficiencies, procedural changes without any application to and permission from the regulatory body, lack of understanding of criticality hazards, and complacency that a criticality accident would not occur. (author)

  2. Analysis of fuel cycles with natural uranium

    International Nuclear Information System (INIS)

    A method was developed and a computer code was written for analysis of fuel cycles and it was applied for heavy water and graphite moderated power reactors. Among a variety of possibilities, three methods which enable best utilization of natural uranium and plutonium production were analyzed. Analysis has shown that reprocessing of irradiated uranium and plutonium utilization in the same or similar type of reactor could increase significantly utilization of natural uranium. Increase of burnup is limited exclusively by costs of reprocessing, plutonium extraction and fabrication of new fuel elements

  3. Fossile fuel and uranium resources

    International Nuclear Information System (INIS)

    The world's resources of coal, lignite, oil, natural gas, shale oil and uranium are reviewed. These quantities depend on the prices which make new resources exploitable. Uranium resources are given exclusively for the USSR, Eastern Europe and China. Their value in terms of energy depends heavily on the reactor type used. All figures given are estimated to be conservative

  4. Research Establishment progress report 1978 - uranium fuel cycle

    International Nuclear Information System (INIS)

    A report of research programs continuing in the following areas is presented: mining and treatment of uranium ores, uranium enrichment, waste treatment, reprocessing and the uranium fuel cycle. Staff responsible for each project are indicated

  5. Nondestructive analysis at B and W's uranium conversion plant

    International Nuclear Information System (INIS)

    Containers and processing lines bearing high and low enriched uranium are routinely analyzed by nondestructive assay. Measurement systems used at Babcock and Wilcox's nuclear fuels plant in Apollo, Pennsylvania include the segmented gamma scanner (SGS) and the stabilized assay meter (SAM-II). These systems have been calibrated and used for a variety of tasks including uranium holdup measurements prior to decommissioning, in situ filter analysis and assay of calcined waste. 2 refs

  6. Uranium and nuclear fuel market - state and purposes

    International Nuclear Information System (INIS)

    The most important factors influenced the world market of natural uranium and uranium enriched for reactor fuels have been presented and discussed. The costs of particular steps of nuclear fuel production have been also shown

  7. Internal dosimetry for uranium fuel manufacture at BNFL

    International Nuclear Information System (INIS)

    At its Springfields Works, near Preston, UK BNFL manufactures uranium fuels and fuel intermediates, in a range of chemical and metallurgical processes. Uranium ore concentrate is converted to uranium metal for the Magnox reactors, uranium hexafluoride (UF6) to uranium dioxide (UO2) for AGR and other oxide reactors, and various intermediate products are produced to meet customer requirements. Thus, uranium compounds with biological retention periods ranging from days (UF6) to years (UO2) are handled on multi-hundred, or thousand, tonne per year scales. Control and minimisation of workforce exposure is exercised primarily by engineered methods (e.g. total enclosures and high integrity plant), backed up by use of respiratory and other protective equipment. A high profile is given to good standards of housekeeping. Assessment of intake is by methods approved by HSE (NII) in the Approved Laboratory Statement on internal dosimetry. The principal method is assessment by use of continuous air sampling combined with occupancy. This is back up by routine personal air sampling (PAS) in selected relevant areas in which ceramic UO2 is handled. Further assurance is provided by programmed PAS in other areas and by systematic, and routine, urinalysis and whole-body monitoring of all relevant members of the workforce. The results of the above are presented in detail. (Author)

  8. Criticality accident in uranium fuel processing plant. Emergency medical care and dose estimation for the severely overexposed patients

    Energy Technology Data Exchange (ETDEWEB)

    Akashi, Makoto; Ishigure, Nobuhito [National Inst. of Radiological Sciences, Chiba (Japan)

    2000-08-01

    A criticality accident occurred in JCO, a plant for nuclear fuel production in 1999 and three workers were exposed to extremely high-level radiation (neutron and {gamma}-ray). This report describes outlines of the clinical courses and the medical cares for the patients of this accident and the emergent medical system for radiation accident in Japan. One (A) of the three workers of JCO had vomiting and diarrhea within several minutes after the accident and another one (B) had also vomiting within one hour after. Based on these evidences, the exposure dose of A and B were estimated to be more than 8 and 4 GyEq, respectively. Generally, acute radiation syndrome (ARS) is assigned into three phases; prodromal phase, critical or manifestation phase and recovery phase or death. In the prodromal phase, anorexia, nausea, vomiting and diarrhea often develop, whereas the second phase is asymptotic. In the third phase, various syndromes including infection, hemorrhage, dehydration shock and neurotic syndromes are apt to occur. It is known that radiation exposure at 1 Gy or more might induce such acute radiation syndromes. Based on the clinical findings of Chernobyl accident, it has been thought that exposure at 0.5 Gy or more causes a lowering of lymphocyte level and a decrease in immunological activities within 48 hours. Lymphocyte count is available as an indicator for the evaluation of exposure dose in early phase, but not in later phase The three workers of JCO underwent chemical analysis of blood components, chromosomal analysis and analysis of blood {sup 24}Na immediately after the arrival at National Institute of Radiological Sciences via National Mito Hospital specified as the third and the second facility for the emergency medical care system in Japan, respectively. (M.N.)

  9. Corrosion Evaluation of RERTR Uranium Molybdenum Fuel

    Energy Technology Data Exchange (ETDEWEB)

    A K Wertsching

    2012-09-01

    As part of the National Nuclear Security Agency (NNSA) mandate to replace the use of highly enriched uranium (HEU) fuel for low enriched uranium (LEU) fuel, research into the development of LEU fuel for research reactors has been active since the late 1970’s. Originally referred to as the Reduced Enrichment for Research and Test Reactor (RERTR) program the new effort named Global Threat Reduction Initiative (GTRI) is nearing the goal of replacing the standard aluminum clad dispersion highly enriched uranium aluminide fuel with a new LEU fuel. The five domestic high performance research reactors undergoing this conversion are High Flux Isotope reactor (HFIR), Advanced Test Reactor (ATR), National Institute of Standards and Technology (NIST) Reactor, Missouri University Research Reactor (MURR) and the Massachusetts Institute of Technology Reactor II (MITR-II). The design of these reactors requires a higher neutron flux than other international research reactors, which to this point has posed unique challenges in the design and development of the new mandated LEU fuel. The new design utilizes a monolithic fuel configuration in order to obtain sufficient 235U within the LEU stoichoimetry to maintain the fission reaction within the domestic test reactors. The change from uranium aluminide dispersion fuel type to uranium molybdenum (UMo) monolithic configuration requires examination of possible corrosion issues associated with the new fuel meat. A focused analysis of the UMo fuel under potential corrosion conditions, within the ATR and under aqueous storage indicates a slow and predictable corrosion rate. Additional corrosion testing is recommended for the highest burn-up fuels to confirm observed corrosion rate trends. This corrosion analysis will focus only on the UMo fuel and will address corrosion of ancillary components such as cladding only in terms of how it affects the fuel. The calculations and corrosion scenarios are weighted with a conservative bias to

  10. Recycled uranium: An advanced fuel for CANDU reactors

    International Nuclear Information System (INIS)

    The use of recycled uranium (RU) fuel offers significant benefits to CANDU reactor operators particularly if used in conjunction with advanced fuel bundle designs that have enhanced performance characteristics. Furthermore, these benefits can be realised using existing fuel production technologies and practices and with almost negligible change to fuel receipt and handling procedures at the reactor. The paper will demonstrate that the supply of RU as a ceramic-grade UO2 powder will increasingly become available as a secure option to virgin natural uranium and slightly enriched uranium(SEU). In the context of RU use in Canadian CANDU reactors, existing national and international transport regulations and arrangements adequately allow all material movements between the reprocessor, RU powder supplier, Canadian CANDU fuel manufacturer and Canadian CANDU reactor operator. Studies have been undertaken of the impact on personnel dose during fuel manufacturing operations from the increased specific activity of the RU compared to natural uranium. These studies have shown that this impact can be readily minimised without significant cost penalty to the acceptable levels recognised in modem standards for fuel manufacturing operations. The successful and extensive use of RU, arising from spent Magnox fuel, in British Energy's Advanced Gas-Cooled reactors is cited as relevant practical commercial scale experience. The CANFLEX fuel bundle design has been developed by AECL (Canada) and KAERI (Korea) to facilitate the achievement of higher bum-ups and greater fuel performance margins necessary if the full economic potential of advanced CANDU fuel cycles are to be achieved. The manufacture of a CANFLEX fuel bundle containing RU pellets derived from irradiated PWR fuel reprocessed in the THORP plant of BNFL is described. This provided a very practical verification of dose modelling calculations and also demonstrated that the increase of external activity is unlikely to require any

  11. Capital and operating costs of irradiated natural uranium reprocessing plants

    International Nuclear Information System (INIS)

    This paper presents first a method of analysing natural uranium reprocessing plants investment costs (method similar to LANG and BACH well known in the fuel oil industry) and their operating costs (analysed according to their economic type). This method helps establishing standard cost structures for these plants, allowing thus comparisons between existing or planned industrial facilities. It also helps evaluating the foreseeable consequences of technical progress. Some results obtained are given, concerning: the investment costs sensitivity to the various technical parameters defining the fuel and their comparison according to the country or the economic area taken into account. Finally, the influence of the plants size on their investment costs is shown. (author)

  12. Design of a uranium recovery pilot plant

    International Nuclear Information System (INIS)

    The engineering design of a pilot plant of uranium recover, is presented. The diagrams and specifications of the equipments such as pipelines, pumps, values tanks, filters, engines, etc... as well as metallic structure and architetonic design is also presented. (author)

  13. Study of the aqueous chemical treatment of uranium zirconium fuels

    International Nuclear Information System (INIS)

    A dry process has been studied for separating the uranium from the zirconium-either for recovering the enriched uranium from fuel element production waste, or with a view to treating this waste after irradiation. In this process the alloy is treated with hydrochloric acid at 400 deg. C in a fluidized corundum bed which causes the zirconium to volatilize as tetrachloride and the uranium to form the trichloride. This latter is then converted to the hexafluoride by attack with fluorure. After the laboratory tests, a first pilot plant with a capacity of 1 kg of alloy was tried out at the Fontenay-aux-Roses Nuclear Research Centre; this made it possible to fix the operational conditions for the process. An industrial scale plant was then built with the collaboration of the from Kuhlmann, and operated until a satisfactory process had been developed for treating the waste. This installation treats 3 kg/h of alloy with a yield for the hydrochloric acid of about 50 per cent and with a uranium loss in the zirconium tetrachloride of about 0.1 per cent. An active pilot plant capable of treating of treating a few kilos of irradiated alloy is now being studied. (authors)

  14. Corrosion testing of uranium silicide fuel specimens

    International Nuclear Information System (INIS)

    U3Si is the most promising high density natural uranium fuel for water-cooled power reactors. Power reactors fuelled with this material are expected to produce cheaper electricity than those fuelled with uranium dioxide. Corrosion tests in 300oC water preceded extensive in-reactor performance tests of fuel elements and bundles. Proper heat-treatment of U-3.9 wt% Si gives a U35i specimen which corrodes at less than 2 mg/cm2 h in 300oC water. This is an order of magnitude lower than the maximum corrosion rate tolerable in a water-cooled reactor. U3Si in a defected unbonded Zircaloy-2 sheath showed only a slow uniform sheath expansion in 300oC water. All tests were done under isothermal conditions in an out-reactor loop. (author)

  15. Criticality accident in uranium fuel processing plant. Influence of the critical accident seen to consciousness investigation of the public

    International Nuclear Information System (INIS)

    Here was introduced a consciousness investigation result carried out at Fukui prefecture and Osaka city after about two months of the JCO criticality accident. Peoples were disturbed by the accident, and not a little changed their individual estimations on items relating to energy. However, peoples lived in Fukui prefecture did not increase rate of opposition against nuclear energy promotion and nuclear power plant construction to their living area on comparison with a year before the accident. This reason might be understood by that the accident was not an accident of a nuclear power plant directly, and that their living area was much distant from place of the accident and was not suffered any danger. On the other hand, public opinion in Osaka city made worse on comparison with that before a year, and if such worse public opinion was thought to be due to the accident, its effect could be said to be different in each area even with no direct relation to the accident to shown a result dependent upon its various conditions. As a rough tendency on psychological disturbance due to the accident, it could be said that peoples became to have feelings of avoiding hard nuclear energy technology at a chance of the accident and to direct thoughts of soft natural energy and environment respect. (G.K.)

  16. Colloids generation from metallic uranium fuel

    International Nuclear Information System (INIS)

    The possibility of colloid generation from spent fuel in an unsaturated environment has significant implications for storage of these fuels in the proposed repository at Yucca Mountain. Because colloids can act as a transport medium for sparingly soluble radionuclides, it might be possible for colloid-associated radionuclides to migrate large distances underground and present a human health concern. This study examines the nature of colloidal materials produced during corrosion of metallic uranium fuel in simulated groundwater at elevated temperature in an unsaturated environment. Colloidal analyses of the leachates from these corrosion tests were performed using dynamic light scattering and transmission electron microscopy. Results from both techniques indicate a bimodal distribution of small discrete particles and aggregates of the small particles. The average diameters of the small, discrete colloids are ∼3--12 nm, and the large aggregates have average diameters of approximately100--200 nm. X-ray diffraction of the solids from these tests indicates a mineral composition of uranium oxide or uranium oxy-hydroxide

  17. Uranium refining and conversion plant decommissioning project

    International Nuclear Information System (INIS)

    The uranium refining and conversion plant (URCP) at Ningyo-toge was constructed in 1981 for the purpose of demonstrating on refining and conversion process from yellow cake (or uranium trioxide) to uranium hexafluoride by way of uranium tetrafluoride. For 20 years, 385 tons of natural uranium hexafluoride and 336 tons of reprocessed uranium hexafluoride (approximately) was produced. There are two different type of refining processes in the URCP. One is the wet process by converting the natural uranium and the other is the dry conversion process for the reprocessed uranium. The dismantling of the dry process facilities began in March, 2008. It was found the large amount of uranium residuals such as wet slurry and powder uranium inside the vessels and pipes. Therefore, we have to take care of the spread of the contamination during dismantling works. The basic strategy concerning plant dismantling were the optimization of the total labor costs and the minimization of the radioactive wastes generated. The dismantling procedure is shown below; 1) measuring doserate by using high sensitivity surveymeters, and nuclide identification by using gamma ray spectrometry, 2) estimating uranium mass inventory, 3) planning work force distributions with radiological survey staffs, 4) deciding dismantling methods concretely, 5) decontaminating schematically if required, 6) collecting detailed data of working conditions, 7) measuring and classifying contaminated materials, 8) managements of radioactive waste drum and non-contaminated equipment, 9) control for personal exposures. Almost all equipment will be decontaminated except building decontamination it by around 2013FY. In addition, the secondary wastes were also yielded. Few thousands man-days were necessary for this project. The measurement data have not showed the high environmental radiation doserate, generally less than 0.3μSv/h. However, by the trace of the reprocessed uranium, the trans-uranium nuclides such as uranium

  18. Uranium

    International Nuclear Information System (INIS)

    The article includes a historical preface about uranium, discovery of portability of sequential fission of uranium, uranium existence, basic raw materials, secondary raw materials, uranium's physical and chemical properties, uranium extraction, nuclear fuel cycle, logistics and estimation of the amount of uranium reserves, producing countries of concentrated uranium oxides and percentage of the world's total production, civilian and military uses of uranium. The use of depleted uranium in the Gulf War, the Balkans and Iraq has caused political and environmental effects which are complex, raising problems and questions about the effects that nuclear compounds left on human health and environment.

  19. Reprocessed uranium experience and UK options for NDA and Springfields Fuels Limited

    International Nuclear Information System (INIS)

    The Nuclear Decommissioning Authority (NDA) is the owner of over 20 000 t U of uranium arising from the reprocessing of Magnox fuel, known in the United Kingdom (UK) as Magnox Depleted Uranium (MD U). This material is stored in the form of uranium trioxide (UO3) at the NDA’s Capenhurst site. The NDA Strategy, published in March 2006, indicated that solutions to deal with MD U would be sought and that NDA would engage with the UK Government and UK Stakeholders to consider the most appropriate management strategies for uranic material. Springfields Fuels Limited (SFL), currently operated by Westinghouse, has recycled over 15 000 t U of MD U reprocessed uranium though its manufacturing facilities in production campaigns between the 1970s and the early 1990s. UO3 was converted to uranium tetrafluoride (UF4) in reduction and hydrofluorination kilns before being converted to uranium hexafluoride in the now decommissioned UF6 plants. Following enrichment, the UF6 was converted to uranium dioxide (UO2) via the integrated dry route kiln process and manufactured into fuel assemblies for the UK’s advanced gas-cooled reactors (AGR), operated by British Energy (BE). SFL has also demonstrated conversion of limited quantities of oxide reprocessed product. The paper provides details of reprocessed uranium stocks in the UK, NDA’s stakeholder engagement and reviews SFL’s experience from recycling uranium at Springfields which can help contribute to finding optimal solutions for UK reprocessed uranium issues. (author)

  20. Recovery of enriched uranium from waste solution obtained from fuel manufacture laboratories

    International Nuclear Information System (INIS)

    Reversed-phase partition chromatography is shown to be a convenient and applicable method for the quantitative recovery of microgram to gram quantities of Uranium (19.7% enriched with 235U) from highly impure solution. The processing of Uranium compounds for atomic energy project especially in FMPP (fuel manufacture pilot plant) gives rise to a variety of wastes in which the Uranium content is of considerable importance. The recovery of Uranium from concentrated mother liquors produced from ADU (ammonium diuranate) precipitation, as well as those due to ADU washing is studied in this work. Column of Poly-trifluoro-monochloro-ethylene (Kel-F) supporting tri-n-butyl-phosphate (TBP) retains Uranium. Impurities are eluted with 6.5 M HCl, and the Uranium is eluted with water and the recovery of Uranium is better than 94%. (author)

  1. Distribution of equilibrium burnup for an homogeneous core with fuel elements of slightly enriched uranium (0.85% U-235) at Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    At Atucha I, the present fuel management with natural uranium comprises three burnup areas and one irradiation path, sometimes performing four steps in the reactor core, according to the requirements. The discharge burnup is 6.0 Mw d/kg U for a waste reactivity of 6.5 m k and a heavy water purity of 99.75%. This is a preliminary study to obtain the distribution of equilibrium burnup of an homogeneous core with slightly enriched uranium (0.85% by weight U-235), using the time-averaged method implemented in the code PUMA and a representative model of one third of core and fixed rod position. It was found a strategy of three areas and two paths that agrees with the present limits of channel power and specific power in fuel rod. The discharge burnup obtained is 11.6 Mw d/kg U. This strategy is calculated with the same method and a full core representation model is used to verify the obtained results. (Author)

  2. Research on using depleted uranium as nuclear fuel for HWR

    International Nuclear Information System (INIS)

    The purpose of our work is to find a way for application of depleted uranium in CANDU reactor by using MOX nuclear fuel of depleted U and Pu instead of natural uranium. From preliminary evaluation and calculation, it was shown that MOX nuclear fuel consisting of depleted uranium enrichment tailings (0.25% 235U) and plutonium (their ratio 99.5%:0.5%) could replace natural uranium in CANDU reactor to sustain chain reaction. The prospects of application of depleted uranium in nuclear energy field are also discussed

  3. Equipment specifications for an electrochemical fuel reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Hemphill, Kevin P [Los Alamos National Laboratory

    2010-01-01

    Electrochemical reprocessing is a technique used to chemically separate and dissolve the components of spent nuclear fuel, in order to produce new metal fuel. There are several different variations to electrochemical reprocessing. These variations are accounted for by both the production of different types of spent nuclear fuel, as well as different states and organizations doing research in the field. For this electrochemical reprocessing plant, the spent fuel will be in the metallurgical form, a product of fast breeder reactors, which are used in many nuclear power plants. The equipment line for this process is divided into two main categories, the fuel refining equipment and the fuel fabrication equipment. The fuel refining equipment is responsible for separating out the plutonium and uranium together, while getting rid of the minor transuranic elements and fission products. The fuel fabrication equipment will then convert this plutonium and uranium mixture into readily usable metal fuel.

  4. Comparison of the radiological hazard of thorium and uranium spent fuels from VVER-1000 reactor

    Science.gov (United States)

    Frybort, Jan

    2014-11-01

    Thorium fuel is considered as a viable alternative to the uranium fuel used in the current generation of nuclear power plants. Switch from uranium to thorium means a complete change of composition of the spent nuclear fuel produced as a result of the fuel depletion during operation of a reactor. If the Th-U fuel cycle is implemented, production of minor actinides in the spent fuel is negligible. This is favourable for the spent fuel disposal. On the other hand, thorium fuel utilisation is connected with production of 232U, which decays via several alpha decays into a strong gamma emitter 208Tl. Presence of this nuclide might complicate manipulations with the irradiated thorium fuel. Monte-Carlo computation code MCNPX can be used to simulate thorium fuel depletion in a VVER-1000 reactor. The calculated actinide composition will be analysed and dose rate from produced gamma radiation will be calculated. The results will be compared to the reference uranium fuel. Dependence of the dose rate on time of decay after the end of irradiation in the reactor will be analysed. This study will compare the radiological hazard of the spent thorium and uranium fuel handling.

  5. Uranium-plutonium carbide as an LMFBR advanced fuel

    International Nuclear Information System (INIS)

    Uranium-plutonium carbide offers an improved fuel system for advanced breeder reactors. The high thermal conductivity and density of carbide fuels permit superior breeding performance and high specific power operation. These advantages combine to increase plutonium production, reduce fuel cycle and power costs, and lower plant capital costs. The carbide advantages are obtained at conservative fuel sytem design and operating conditions. Carbide fabrication technology has been demonstrated by the production of quality-assured fuel elements for irradiation testing. The carbide irradiation test program has demonstrated that high burnup can be achieved with several designs and that the consequences of postulated off-normal operating events are benign. Design bases to support helium- and sodium-bonded carbide fuel pin test irradiations in the Fast Flux Test Facility have been developed in the Experimental Breeder Reactor-II and the Transient Reactor irradiation experiments. Important issues regarding safety, reprocessing, and commercial-scale fabrication remain to be addressed in the continuing development of carbide fuels. Fiscal and historical circumstances have combined to preclude this development. This report reviews these circumstances and the state of the technology in general and advances a rationale for why development should be continued

  6. Comparison of the radiological hazard of thorium and uranium spent fuels from VVER-1000 reactor

    International Nuclear Information System (INIS)

    Thorium fuel is considered as a viable alternative to the uranium fuel used in the current generation of nuclear power plants. Switch from uranium to thorium means a complete change of composition of the spent nuclear fuel produced as a result of the fuel depletion during operation of a reactor. If the Th–U fuel cycle is implemented, production of minor actinides in the spent fuel is negligible. This is favourable for the spent fuel disposal. On the other hand, thorium fuel utilisation is connected with production of 232U, which decays via several alpha decays into a strong gamma emitter 208Tl. Presence of this nuclide might complicate manipulations with the irradiated thorium fuel. Monte-Carlo computation code MCNPX can be used to simulate thorium fuel depletion in a VVER-1000 reactor. The calculated actinide composition will be analysed and dose rate from produced gamma radiation will be calculated. The results will be compared to the reference uranium fuel. Dependence of the dose rate on time of decay after the end of irradiation in the reactor will be analysed. This study will compare the radiological hazard of the spent thorium and uranium fuel handling. - Highlights: • Spent thorium and uranium fuel composition in VVER-1000 was calculated by MCNPX. • Important nuclide 208Tl is not included in the thorium spent fuel composition. • There are large differences in activity of actinides between spent Th and U fuels. • Dose rate from spent thorium fuel is increasing during 50 years of decay. • 208Tl in spent thorium fuel increased the dose rate by several percent

  7. Uranium-zirconium hydride fuel properties

    Energy Technology Data Exchange (ETDEWEB)

    Olander, D. [Department of Nuclear Engineering, University of California at Berkeley, Berkeley, CA 94720 (United States)], E-mail: fuelpr@nuc.berkeley.edu; Greenspan, Ehud [Department of Nuclear Engineering, University of California at Berkeley, Berkeley, CA 94720 (United States); Garkisch, Hans D. [Westinghouse Electric Company LLC, Pittsburgh, PA 15236 (United States); Petrovic, Bojan [School of Mechanical Engineering, Georgia Institute of Technology, Atlanta, GA (United States)

    2009-08-15

    Properties of the two-phase hydride U{sub 0.3}ZrH{sub 1.6} pertinent to performance as a nuclear fuel for LWRs are reviewed. Much of the available data come from the Space Nuclear Auxiliary Power (SNAP) program of 4 decades ago and from the more restricted data base prepared for the TRIGA research reactors some 3 decades back. Transport, mechanical, thermal and chemical properties are summarized. A principal difference between oxide and hydride fuels is the high thermal conductivity of the latter. This feature greatly decreases the temperature drop over the fuel during operation, thereby reducing the release of fission gases to the fraction due only to recoil. However, very unusual early swelling due to void formation around the uranium particles has been observed in hydride fuels. Avoidance of this source of swelling limits the maximum fuel temperature to {approx}650 deg. C (the design limit recommended by the fuel developer is 750 deg. C). To satisfy this temperature limitation, the fuel-cladding gap needs to be bonded with a liquid metal instead of helium. Because the former has a thermal conductivity {approx}100 times larger than the latter, there is no restriction on gap thickness as there is in helium-bonded fuel rods. This opens the possibility of initial gap sizes large enough to significantly delay the onset of pellet-cladding mechanical interaction (PCMI). The large fission-product swelling rate of hydride fuel (3x that of oxide fuel) requires an initial radial fuel-cladding gap of {approx}300 m if PCMI is to be avoided. The liquid-metal bond permits operation of the fuel at current LWR linear-heat-generation rates without exceeding any design constraint. The behavior of hydrogen in the fuel is the source of phenomena during operation that are absent in oxide fuels. Because of the large heat of transport (thermal diffusivity) of H in ZrH{sub x}, redistribution of hydrogen in the temperature gradient in the fuel pellet changes the initial H/Zr ratio of 1

  8. Uranium-zirconium hydride fuel properties

    International Nuclear Information System (INIS)

    Properties of the two-phase hydride U0.3ZrH1.6 pertinent to performance as a nuclear fuel for LWRs are reviewed. Much of the available data come from the Space Nuclear Auxiliary Power (SNAP) program of 4 decades ago and from the more restricted data base prepared for the TRIGA research reactors some 3 decades back. Transport, mechanical, thermal and chemical properties are summarized. A principal difference between oxide and hydride fuels is the high thermal conductivity of the latter. This feature greatly decreases the temperature drop over the fuel during operation, thereby reducing the release of fission gases to the fraction due only to recoil. However, very unusual early swelling due to void formation around the uranium particles has been observed in hydride fuels. Avoidance of this source of swelling limits the maximum fuel temperature to ∼650 deg. C (the design limit recommended by the fuel developer is 750 deg. C). To satisfy this temperature limitation, the fuel-cladding gap needs to be bonded with a liquid metal instead of helium. Because the former has a thermal conductivity ∼100 times larger than the latter, there is no restriction on gap thickness as there is in helium-bonded fuel rods. This opens the possibility of initial gap sizes large enough to significantly delay the onset of pellet-cladding mechanical interaction (PCMI). The large fission-product swelling rate of hydride fuel (3x that of oxide fuel) requires an initial radial fuel-cladding gap of ∼300 m if PCMI is to be avoided. The liquid-metal bond permits operation of the fuel at current LWR linear-heat-generation rates without exceeding any design constraint. The behavior of hydrogen in the fuel is the source of phenomena during operation that are absent in oxide fuels. Because of the large heat of transport (thermal diffusivity) of H in ZrHx, redistribution of hydrogen in the temperature gradient in the fuel pellet changes the initial H/Zr ratio of 1.6 to ∼1.45 at the center and

  9. Idaho Chemical Processing Plant and Plutonium-Uranium Extraction Plant phaseout/deactivation study

    International Nuclear Information System (INIS)

    The decision to cease all US Department of Energy (DOE) reprocessing of nuclear fuels was made on April 28, 1992. This study provides insight into and a comparison of the management, technical, compliance, and safety strategies for deactivating the Idaho Chemical Processing Plant (ICPP) at Westinghouse Idaho Nuclear Company (WINCO) and the Westinghouse Hanford Company (WHC) Plutonium-Uranium Extraction (PUREX) Plant. The purpose of this study is to ensure that lessons-learned and future plans are coordinated between the two facilities

  10. Fuel rod reprocessing plant

    International Nuclear Information System (INIS)

    A plant for the reprocessing of fuel rods for a nuclear reactor comprises a plurality of rectangular compartments desirably arranged on a rectangular grid. Signal lines, power lines, pipes, conduits for instrumentation, and other communication lines leave a compartment just below its top edges. A vehicle access zone permits overhead and/or mobile cranes to remove covers from compartments. The number of compartments is at least 25% greater than the number of compartments used in the initial design and operation of the plant. Vacant compartments are available in which replacement apparatus can be constructed. At the time of the replacement of a unit, the piping and conduits are altered to utilize the substitute equipment in the formerly vacant compartment, and it is put on stream prior to dismantling old equipment from the previous compartment. Thus the downtime for the reprocessing plant for such a changeover is less than in a traditional reprocessing plant

  11. The Chemwes uranium plant: A case history

    International Nuclear Information System (INIS)

    During the 1970s when the Nuexco exchange value for U3O8 rose from US $6 to $43 per pound, the recovery of uranium from even comparatively low grade deposits appeared to be attractive. Two mines in the Klerksdorp area of the Republic of South Africa, Stilfontein and Buffelsfontein, had been stockpiling uranium bearing tailings material since the early 1960s, and initial sampling of these and other smaller sources of residue in the area suggested that the establishment of a central uranium beneficiation plant to process such material would be economically feasible. Preliminary studies showed that the uranium content of the tailings could not be economically concentrated before leaching, but that the pyrite in the plant tailings could possibly be concentrated by flotation, with subsequent roasting to provide both the acid needed in the uranium dissolution process and a calcine product from which gold could be recovered. A preliminary feasibility study suggested that an operation of 270 kt per month would be the most attractive in economic terms. It was decided that a contract for the expected production should be negotiated so that this security could be used to support the financing of the project. The paper gives a description of the performance of the plant so far. The plant performance is analysed from the processing and the mechanical points of view, with special emphasis on the leaching, solid-liquid separation, recovery, and purification sections. The criteria used in the initial selection of the process are reviewed and compared with the subsequent performance of the plant. (author). 2 refs, 4 figs, 2 tabs

  12. An overview of the regulation of uranium mining, milling, refining and fuel fabrication

    International Nuclear Information System (INIS)

    The mining, milling, refining and fabrication of uranium into nuclear fuel are activities that have in common the handling of natural uranium. The occupational and environmental hazards resulting from these activities vary widely. Uranium presents a radiological hazard throughout, but the principal culprit is radium which creates an occupational hazard in the mine and mill and an environmental hazard in the waste products produced in both the mill and the refinery. The chemicals used in both these latter processes also present hazards. Fuel fabrication presents the least potential for occupational and environmental hazards. The Canadian Atomic Energy Control Board licenses eight plants, and one plant for the extraction of uranium from phosphoric acid. The licensing process is characterised by approval in stages, the placing of the burden of proof on the applicant, inspection at all stages, and joint review by all regulatory agencies involved

  13. Nuclear fuel reprocessing plant

    International Nuclear Information System (INIS)

    The present invention concerns an improvement for corrosion resistance of the welded portion of materials which constitutes a reprocessing plant of spent nuclear fuels. That is, Mo-added austenite stainless steel is used for a plant member at the portion in contact with a nitric acid solution. Then, laser beams are irradiated to the welded portion of the plant member and the surface layer is heated to higher than 1,000degC. If such a heat treatment is applied, the degradation of corrosion resistance of the welded portion can be eliminated at the surface. Further, since laser beams are utilized, heating can be limited only to the surface. Accordingly, undesired thermal deformation of the plant members can be prevented. As a result, the plant member having high pit corrosion resistance against a dissolution solution for spent fuels containing sludges comprising insoluble residue and having resistance to nitric acid solution also in the welded portion substantially equal to that of the matrix can be attained. (I.S.)

  14. Minimization of waste from uranium purification, enrichment and fuel fabrication

    International Nuclear Information System (INIS)

    As any industry, nuclear industry generates a diverse range of waste which has to be managed in a safe manner to be acceptable to the public and the environment. The cost of waste management, the risks to the public and employees, and the detriment to the environment are dependent on the quantity and radioactive content of the waste generated. Waste minimization is a necessary activity needed to reduce the impact from nuclear fuel cycle operations and it is included in the national policy of some countries. In recognition of the importance of the subject, the IAEA has decided to review the current status of the work aimed at waste minimization in the nuclear fuel cycle. The waste minimization issues related to the back end of the nuclear fuel cycle are covered in Technical Reports Series No. 377 'Minimization of Radioactive Waste from Nuclear Power Plants and the Back End of the Nuclear Fuel Cycle' published in 1995. The present report deals with the front end of the nuclear fuel cycle, including existing options, approaches, developments and some specific considerations to be taken into account in decision making on waste minimization. It has been recognized that, in comparison with the back end of the nuclear fuel cycle, much less information is available, and this report should be considered as a first attempt to analyse waste minimization practices and opportunities in uranium purification, conversion, enrichment and fuel fabrication. Although mining and milling is an important part of the front end of the nuclear fuel cycle, these activities are excluded from consideration since relevant activities are covered in other IAEA publications

  15. NRC licensing of uranium enrichment plants

    International Nuclear Information System (INIS)

    The US Nuclear Regulatory Commission (NRC) is preparing a rule making that establishes the licensing requirements for low-enriched uranium enrichment plants. Although implementation of this rule making is timed to correspond with receipt of a license application for the Louisiana Energy Services centrifuge enrichment plant, the rule making is applicable to all uranium enrichment technologies. If ownership of the US gaseous diffusion plants and/or atomic vapor laser isotope separation is transferred to a private or government corporation, these plants also would be licensable under the new rule making. The Safeguards Studies Department was tasked by the NRC to provide technical assistance in support of the rule making and guidance preparation process. The initial and primary effort of this task involved the characterization of the potential safeguards concerns associated with a commercial enrichment plant, and the licensing issues associated with these concerns. The primary safeguards considerations were identified as detection of the loss of special nuclear material, detection of unauthorized production of material of low strategic significance, and detection of production of uranium enriched to >10% 235U. The primary safeguards concerns identified were (1) large absolute limit of error associated with the material balance closing, (2) the inability to shutdown some technologies to perform a cleanout inventory of the process system, and (3) the flexibility of some technologies to produce higher enrichments. Unauthorized production scenarios were identified for some technologies that could prevent conventional material control and accounting programs from detecting the production and removal of 5 kg 235U as highly enriched uranium. Safeguards techniques were identified to mitigate these concerns

  16. Possibilities of using metal uranium fuel in heavy water reactors

    International Nuclear Information System (INIS)

    There are serious economic reasons for using metal uranium in heavy water reactors, because of its high density, i.e. high conversion factor, and low cost of fuel elements production. Most important disadvantages are swelling at high burnup and corrosion risk. Some design concepts and application of improved uranium obtained by alloying are promising for achievement of satisfactory stability of metal uranium under reactor operation conditions

  17. Natural uranium fueled light water moderated breeding hybrid power reactors

    International Nuclear Information System (INIS)

    The feasibility of fission-fusion hybrid reactors based on breeding light water thermal fission systems is investigated. The emphasis is on fuel-self-sufficient (FSS) hybrid power reactors that are fueled with natural uranium. Other LWHRs considered include FSS-LWHRs that are fueled with spent fuel from LWRs, and LWHRs which are to supplement LWRs to provide a tandem LWR-LWHR power economy that is fuel-self-sufficient

  18. Assimilation of uranium by wheat and tomato plants

    International Nuclear Information System (INIS)

    Greenhouse conditions have been used for the study of uptake of uranium by wheat and tomato plants as affected by its concentration in soil and irrigation applied. The highest yield of wheat was obtained at 3.0 ppm of uranium whereas the tomato yield decreased with the increase of uranium in the soil. The analysis shows that Uranium uptake by wheat and tomato not only depends upon the uranium concentration in the soil but also on the amount of irrigation applied. (orig.)

  19. Back-end fuel cycle efficiencies with respect to improved uranium utilization

    International Nuclear Information System (INIS)

    The world-wide nuclear power plant (NPP) capacity is at present 160 GW(e). If one adds the power stations under construction and ordered, a plant capacity of approximately 480 GW(e) is obtained for 1990, with the share of LWRs making up more than 80%. A modern LWR consumes in the open fuel cycle about 4400 metric tonnes of natural uranium per GW(e), assuming a lifetime of 30 years and a load factor of 70%. Considering the natural uranium reserves known at present and exploitable under economic conditions, it can be conveniently estimated that, with the present NPP capacity extension perspective, the natural uranium resources may be exhausted in a few decades. This trend can be counteracted in a flexible manner by various approaches in fuel cycle technology and strategy: (i) by steady further development of the established LWR technology the uranium consumption can be reduced by about 15%; (ii) closing the nuclear fuel cycle on the basis of LWRs (i.e. thermal uranium and plutonium recycling) implies up to 40% savings in natural uranium consumption; (iii) more recent considerations include the advanced pressurized water reactor (APWR). The APWR combines the proven PWR technology with a newly developed tight lattice core with greatly improved conversion characteristics (conversion ratio = 0.90 to 0.95). In terms of uranium utilization, the APWR has an efficiency three to five times higher than a PWR; (iv) Commercial introduction of FBR systems results in an optimal utilization of uranium which, at the same time, guarantees the supply of nuclear fuel well beyond the present century. For a corresponding transition period an energy supply system can be conceived which relies essentially on extended back-end fuel cycle capacities. These would facilitate a symbiosis of PWR, APWR and FBR, characterized by high flexibility with respect to long-term developments on the energy market. (author)

  20. Towards Multi Fuel SOFC Plant

    DEFF Research Database (Denmark)

    Rokni, Masoud; Clausen, Lasse Røngaard; Bang-Møller, Christian

    2011-01-01

    Complete Solid Oxide Fuel Cell (SOFC) plants fed by several different fuels are suggested and analyzed. The plants sizes are about 10 kW which is suitable for single family house with needs for both electricity and heat. Alternative fuels such as, methanol, DME (Di-Methyl Ether) and ethanol...... are also considered and the results will be compared with the base plant fed by Natural Gas (NG). A single plant design will be suggested that can be fed with methanol, DME and ethanol whenever these fuels are available. It will be shown that the plant fed by ethanol will have slightly higher electrical...

  1. Development of Advanced High Uranium Density Fuels for Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Blanchard, James [Univ. of Wisconsin, Madison, WI (United States); Butt, Darryl [Boise State Univ., ID (United States); Meyer, Mitchell [Idaho National Lab. (INL), Idaho Falls, ID (United States); Xu, Peng [Westinghouse Electric Corporation, Pittsburgh, PA (United States)

    2016-02-15

    This work conducts basic materials research (fabrication, radiation resistance, thermal conductivity, and corrosion response) on U3Si2 and UN, two high uranium density fuel forms that have a high potential for success as advanced light water reactor (LWR) fuels. The outcome of this proposed work will serve as the basis for the development of advance LWR fuels, and utilization of such fuel forms can lead to the optimization of the fuel performance related plant operating limits such as power density, power ramp rate and cycle length.

  2. Possibility of using metal uranium fuel in heavy water reactors

    International Nuclear Information System (INIS)

    The review of metal uranium properties including irradiation in the reactor core lead to the following conclusions. Using metal uranium in the heavy water reactors would be favourable from economic point of view for ita high density, i.e. high conversion factor and low cost of fuel elements fabrication. Most important constraint is swelling during burnup and corrosion

  3. Analysis of fuel cycles with natural uranium, Phase I

    International Nuclear Information System (INIS)

    This paper contains analyses of fuel cycles with natural uranium for the following cases: plutonium recycling is not done; recycling of plutonium and irradiated uranium with the condition of equal multiplication factor at the beginning of each cycle; and recycling of plutonium only

  4. Delays hit conversion to low enriched uranium fuel

    Science.gov (United States)

    Allen, Michael

    2016-03-01

    Eliminating highly enriched uranium (HEU) fuel from civilian research reactors around the world will take a lot longer than anticipated, according to a new study by the US National Academies of Sciences, Engineering and Medicine.

  5. Method to manufacture nuclear fuel monocarbide, particularly uranium and uranium-plutonium monocarbide

    International Nuclear Information System (INIS)

    Uranium monocarbide and U-Pu monocarbides are manufactured by converting the oxides with uranium carbide (mol ratio 3:1) at temperatures >16000C in vacuum (-6 bar). Only small quantities of CO are formed, one obtains high-density pellets which are very suitable as nuclear fuels. (IHOE)

  6. DUSCOBS - a depleted-uranium silicate backfill for transport, storage, and disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    A Depleted Uranium Silicate COntainer Backfill System (DUSCOBS) is proposed that would use small, isotopically-depleted uranium silicate glass beads as a backfill material inside storage, transport, and repository waste packages containing spent nuclear fuel (SNF). The uranium silicate glass beads would fill all void space inside the package including the coolant channels inside SNF assemblies. Based on preliminary analysis, the following benefits have been identified. DUSCOBS improves repository waste package performance by three mechanisms. First, it reduces the radionuclide releases from SNF when water enters the waste package by creating a local uranium silicate saturated groundwater environment that suppresses (1) the dissolution and/or transformation of uranium dioxide fuel pellets and, hence, (2) the release of radionuclides incorporated into the SNF pellets. Second, the potential for long-term nuclear criticality is reduced by isotopic exchange of enriched uranium in SNF with the depleted uranium (DU) in the glass. Third, the backfill reduces radiation interactions between SNF and the local environment (package and local geology) and thus reduces generation of hydrogen, acids, and other chemicals that degrade the waste package system. In addition, the DUSCOBS improves the integrity of the package by acting as a packing material and ensures criticality control for the package during SNF storage and transport. Finally, DUSCOBS provides a potential method to dispose of significant quantities of excess DU from uranium enrichment plants at potential economic savings. DUSCOBS is a new concept. Consequently, the concept has not been optimized or demonstrated in laboratory experiments

  7. Test bed for high-uranium-loaded fuel plates

    Energy Technology Data Exchange (ETDEWEB)

    Senn, R.L.; Martin, M.M.

    1979-01-01

    An irradiation test facility has been designed and built to provide a test bed for irradiating a variety of miniature fuel plates. The objective of these tests is to screen various candidate materials as to their suitability for replacing the fully enriched uranium fuel materials currently used by the world's test and research reactors with a lower enrichment fuel material, without significantly degrading reactor operating characteristics and power levels. The use of low-uranium enrichment (up to 45%) in place of highly-enriched fuel for these reactors would reduce the potential for /sup 235/U diversion.

  8. Uranium enrichment plant risk analysis

    International Nuclear Information System (INIS)

    A method for risk analysis of enrichment facilities is presented and applied to a small scale ultracentrifuge plant. Internal events are identified and the consequences of accidental releases of U F6 are quantified in terms of its toxicological and radiological impact. It is shown that releases in the feed and the cascade areas offers no hazards to the public . Releases of liquefied U F6 in the withdrawal areas, associated with failures in the building isolation systems, may cause undesirable consequences. (author). 11 refs, 4 figs, 3 tabs

  9. Towards Multi Fuel SOFC Plant

    OpenAIRE

    Rokni, Masoud; Clausen, Lasse Røngaard; Bang-Møller, Christian

    2011-01-01

    Complete Solid Oxide Fuel Cell (SOFC) plants fed by several different fuels are suggested and analyzed. The plants sizes are about 10 kW which is suitable for single family house with needs for both electricity and heat. Alternative fuels such as, methanol, DME (Di-Methyl Ether) and ethanol are also considered and the results will be compared with the base plant fed by Natural Gas (NG). A single plant design will be suggested that can be fed with methanol, DME and ethanol whenever these fuels...

  10. Towards a comprehensive uranium fuel management policy for Australia

    International Nuclear Information System (INIS)

    A consideration of the problems inherent in the extraction and use of uranium fuels, in particular the long-term radiation hazard posed by wastes, leads to the recognition of several severe geological and social constraints that must govern any comprehensive management policy. In the light of these, and of the need to prevent diversion of nuclear material for the manufacture of weapons, the author outlines a comprehensive uranium fuel management policy for Australia. It is recommended that Australia's uranium should not be mined until the feasibility and cost of such a management policy have been assessed and compared with the added risks entailed in any less stringent management policy. (Author)

  11. Impact of High Burnup Uranium Oxide and Mixed Uranium-Plutonium Oxide Water Reactor Fuel on Spent Fuel Management

    International Nuclear Information System (INIS)

    licensed for maximum burnup and enrichment. Increased decay heat places an additional load on plant cooling systems. Increased neutron activity requires radiometric instruments (used to control criticality) to be recalibrated. Increased alpha activity results in increased heat generation. Increased specific activity in reprocessing phases may result in higher radionuclide discharges to the environment and may result in more HLW. These effects can be managed using blending schemes. As the burnup exceeds a certain level, a new reprocessing facility may be needed. The reprocessing of spent MOX fuel presents additional challenges due to the lower solubility of plutonium. For fuel disposal, higher burnup UOX and MOX fuel means higher source terms of the radionuclides leading to a potentially higher release to the groundwater. Higher heat loads could exceed temperature limits in a repository. This may require significant repository operational changes to accommodate higher burnup UOX and MOX, such as increased repository space (although the reduced volume of higher burnup UOX may counteract the need for additional space), smaller waste containers, longer decay times at the surface prior to loading into the repository, and additional shielding during spent fuel transfer from the transportation cask. If reprocessing is the disposition method of choice the increase in discharge burnup has a significant effect on the isotopic quality of recycled fuel. Increased enrichment of the reprocessed uranium or an increased amount of plutonium in MOX fuel will be required to meet the same burnup target. Increases in shielding may be required for fuel refabrication operations.

  12. Composition and Distribution of Tramp Uranium Contamination on BWR and PWR Fuel Rods

    International Nuclear Information System (INIS)

    In a joint research project of VGB and AREVA NP GmbH the behaviour of alpha nuclides in nuclear power plants with light water reactors has been investigated. Understanding the source and the behaviour of alpha nuclides is of big importance for planning radiation protection measures for outages and upcoming dismantling projects. Previous publications have shown the correlation between plant specific alpha contamination of the core and the so called 'tramp fuel' or 'tramp uranium' level which is linked to the defect history of fuel assemblies and accordingly the amount of previously washed out fuel from defective fuel rods. The methodology of tramp fuel estimation is based on fission product concentrations in reactor coolant but also needs a good knowledge of tramp fuel composition and in-core distribution on the outer surface of fuel rods itself. Sampling campaigns of CRUD deposits of irradiated fuel assemblies in different NPPs were performed. CRUD analyses including nuclide specific alpha analysis have shown systematic differences between BWR and PWR plants. Those data combined with literature results of fuel pellet investigations led to model improvements showing that a main part of fission products is caused by fission of Pu-239 an activation product of U-238. CRUD investigations also gave a better picture of the in-core composition and distribution of the tramp uranium contamination. It was shown that the tramp uranium distribution in PWR plants is time dependent. Even new fuel assemblies will be notably contaminated after only one cycle of operation. For PWR applies the following logic: the higher the local power the higher the contamination. With increasing burnup the local rod power usually decreases leading to decreasing tramp uranium contamination on the fuel rod surface. This is not applicable for tramp uranium contamination in BWR. CRUD contamination (including the tramp fuel deposits) is much more fixed and is constantly increasing

  13. Irradiation behavior of miniature experimental uranium silicide fuel plates

    International Nuclear Information System (INIS)

    Uranium silicides, because of their relatively high uranium density, were selected as candidate dispersion fuels for the higher fuel densities required in the Reduced Enrichment Research and Test Reactor (RERTR) Program. Irradiation experience with this type of fuel, however, was limited to relatively modest fission densities in the bulk from, on the order of 7 x 1020 cm-3, far short of the approximately 20 x 1020 cm-3 goal established for the RERTR program. The purpose of the irradiation experiments on silicide fuels on the ORR, therefore, was to investigate the intrinsic irradiation behavior of uranium silicide as a dispersion fuel. Of particular interest was the interaction between the silicide particles and the aluminum matrix, the swelling behavior of the silicide particles, and the maximum volume fraction of silicide particles that could be contained in the aluminum matrix

  14. Uranium resources and production of nuclear fuel material in the world

    International Nuclear Information System (INIS)

    Data are summed up relating to uranium resources, the technology of nuclear fuel manufacture, the prices of U concentrate and the costs of the basic operations of the fuel cycle. Resources are sufficient for the needs of nuclear power production based on LWR's without closing the fuel cycle for at least up to the year 2010; in the subsequent period fuels of higher price categories would have to be used (> or approx. 130 $/kg U). The manufacture of U concentrate after culmination in the years 1979 to 1980 is declining as is its price. Fuel material on an operation scale (15a t/year and more) is now being manufactured in 11 countries, with the US leading followed by Canada and Japan. From the point of view of technology the following methods are interesting: processing complex ores (USSR), the use of wastes from the production of nuclear purity uranium in agriculture (Canada), single-stage reconversion of UF6 to UO2 (the so-called IDR process in the UK), chemical methods of uranium enrichment (France and Japan), etc. Great attention is being devoted to the design and construction of spent fuel reprocessing plants and thus the introduction of the closed fuel cycle which is the only way towards the full utilization of uranium, to the transition to fast reactors and the extended service life of nuclear power way into the next millenium. (author)

  15. Recycling of plutonium and uranium in water reactor fuels

    International Nuclear Information System (INIS)

    The purpose of the meeting was to make a review of the present knowledge relevant to plutonium and uranium recycling, MOX fuel, on-going programmes, today's industrial capabilities and future plans for development. For countries with commitments to reprocessing, MOX fuel is attractive and will be more so as discharge burnups increase and as the time between discharge and reprocessing optimized. Fabrication experience on MOX fuel has accumulated for many years in several countries and one has been able to cope with the extension of capacities of the plants, as required by MOX fuel implementation, and with the requirements specific to massive use in power reactors. Standards fabrication processes have proven to be adaptable in large quantities and have yielded products satisfying all present specifications. A large body of irradiation experience for some time on various MOX and RepU materials. On the basis of a comparison with UO2, no adverse effect has been observed. Problems like isotopic homogeneity, solubility, alternative processes like gelation deserve further attention. It is encouraging to note that parameters linked to materials obtained by different fabrication routes can be taken into account by existing codes, to an extent similar to various UO2 fuels, provided an adequate data base is available. The fabrication capacities are the limiting factor for MOX penetration in reactors, where a 30 to 50% recycling rate is therefore sufficient. The use of plutonium in 100% MOX reactors or in more advanced reactors deserves more study. The increase of plutonium inventory may influence safety and licensing analysis, but all the safety criteria can be met. On the whole, the experience reported in this meeting pointed to a general consensus of the attractiveness of recycling and the already demonstrated ability of several countries to cope with all questions raised by MOX substitution of UO2 fuel. Refs, figs and tabs

  16. Plant-uptake of uranium: Hydroponic and soil system studies

    Science.gov (United States)

    Ramaswami, A.; Carr, P.; Burkhardt, M.

    2001-01-01

    Limited information is available on screening and selection of terrestrial plants for uptake and translocation of uranium from soil. This article evaluates the removal of uranium from water and soil by selected plants, comparing plant performance in hydroponic systems with that in two soil systems (a sandy-loam soil and an organic-rich soil). Plants selected for this study were Sunflower (Helianthus giganteus), Spring Vetch (Vicia sativa), Hairy Vetch (Vicia villosa), Juniper (Juniperus monosperma), Indian Mustard (Brassica juncea), and Bush Bean (Phaseolus nanus). Plant performance was evaluated both in terms of the percent uranium extracted from the three systems, as well as the biological absorption coefficient (BAC) that normalized uranium uptake to plant biomass. Study results indicate that uranium extraction efficiency decreased sharply across hydroponic, sandy and organic soil systems, indicating that soil organic matter sequestered uranium, rendering it largely unavailable for plant uptake. These results indicate that site-specific soils must be used to screen plants for uranium extraction capability; plant behavior in hydroponic systems does not correlate well with that in soil systems. One plant species, Juniper, exhibited consistent uranium extraction efficiencies and BACs in both sandy and organic soils, suggesting unique uranium extraction capabilities.

  17. Safety of Uranium Fuel Fabrication Facilities. Specific Safety Guide

    International Nuclear Information System (INIS)

    This Safety Guide supplements the Safety Requirements publication Safety of Fuel Cycle Facilities and addresses all the stages in the life cycle of uranium fuel fabrication facilities, with emphasis being placed on design and operation. It describes the actions, conditions and procedures for meeting safety requirements and deals specifically with the handling, processing and storage of low enriched uranium that has a 235U concentration of no more than 6%, derived from natural, highly enriched or reprocessed uranium. The publication is intended to be of use to designers, operating organizations and regulators to ensure the safety of uranium fuel fabrication facilities. Contents: 1. Introduction; 2. General safety recommendations; 3. Site evaluation; 4. Design; 5. Construction; 6. Commissioning; 7. Operation; 8. Decommissioning; Annexes.

  18. A NDT Method For Uranium Fuel Enrichment Verification

    International Nuclear Information System (INIS)

    Nondestructive testing of fresh uranium fuel materials as a part of a QA programme in a nuclear industry is wide accepted today. These inspections of the fuel materials are based on the well known methods. As a participation of the NET Laboratory in the new QA programme at the VINCA Institute, a method for the absolute determination of the enrichment value of fresh high enriched uranium fuel segments, based on a gamma-ray spectroscopy, is developed. It is primary based on the application of a developed ANA computer code for a gamma spectrum analysis, and the initial experience gained in the previous works on the QA at the VINCA Institute. (author)

  19. Uranium in Dartmoor plants of southwest England

    International Nuclear Information System (INIS)

    A number of plants from Dartmoor, SW England, were investigated for their U contents. Uranium concentrations in the substrate ranged from 7.8 μg/g in the granite bedrock to 2.8 μg/g in the A horizon. Heather, gorse, bracken and a number of grasses growing in this substrate were sampled and analysed. Among these heather and gorse contained average U concentrations of 0.14 μg/g and 0.13 μg/g, respectively. Bracken and grasses have levels below the detection threshold of 0.02 μg/g. 9 refs.; 1 figure; 3 tabs

  20. 75 FR 38809 - Southern Turner Cimarron I, LLC; Notice of Filing

    Science.gov (United States)

    2010-07-06

    ... Energy Regulatory Commission Southern Turner Cimarron I, LLC; Notice of Filing June 25, 2010. Take notice that on June 24, 2010, Southern Turner Cimarron I, LLC filed a supplement confirming passive ownership... in accordance with Rules 211 and 214 of the Commission's Rules of Practice and Procedure (18 CFR...

  1. Radioactive decay properties of CANDU fuel. Volume 1: the natural uranium fuel cycle

    International Nuclear Information System (INIS)

    The computer code CANIGEN was used to obtain the mass, activity, decay heat and toxicity of CANDU fuel and its component isotopes. Data are also presented on gamma spectra and neutron emissions. Part 1 presents these data for unirradiated fuel, uranium ore and uranium mill tailings. In Part 2 they have been computed for fuel irradiated to levels of burnup ranging from 140 GJ/kg U to 1150 GJ/kg U. (author)

  2. Detailed analysis of uranium silicide dispersion fuel swelling

    International Nuclear Information System (INIS)

    Swelling of U3Si and U3Si2 is analyzed. The growth of fission gas bubbles appears to be affected by fission rate, fuel loading, and micro structural change taking place in the fuel compounds during irradiation. Several mechanisms are explored to explain the observations. The present work is aimed at a better understanding of the basic swelling phenomenon in order to accurately model irradiation behavior of uranium silicide dispersion fuel. (orig.)

  3. The Ellweiler uranium plant - a demolition and recycling project

    International Nuclear Information System (INIS)

    The uranium plant at Ellweiler, district of Birkenfeld, was used for the production and storage of uranium concentrates. The owner of the Ellweiler uranium plant (UAE), Gewerkschaft Brunhilde GmbH, ceased processing uranium ore and recycling in 1989 and has been in liquidation since September 1991. The State of Rhineland-Palatinate, had safety measures adopted in a first step, getting the plant into a safe state by former plant personnel. The entire plant was demolished in a second step. The contract for demolishing the former uranium plant was awarded to ABB Reaktor as the general contractor in August 1996. Demolition work was carried out between April 1997 and May 1999. A total of approx. 7900 Mg of material was disposed of. At present, recultivation measures are being carried out. (orig.)

  4. Fuel cost analysis of CANDU-PHWR Wolsung Nuclear Power Plant unit 1

    International Nuclear Information System (INIS)

    Being based on the Segal method, calculation was carried out for the natural uranium nuclear fuel cost with Zircaloy-4 cladding having design parameters of Wolsung Nuclear Power Plant, CANDU-PHWR (Unit 1), currently under construction in Korea aiming at its completion in 1982. An attempt was also made for the sensitivity analysis of each fuel component; i.e., depreciation of fuel manufacturing plant caused by its life time, its load factor, production scale expansion of plant facilities, variations of construction and operating costs of fuel manufacturing plant, fluctuation of interest rates, extent of uranium ore price increases and effect of learning factor. (author)

  5. Evaluation of economical at a uranium enrichment demonstration plant

    International Nuclear Information System (INIS)

    In this report, the economy of technical achievement apply in the uranium enrichment demonstration plant is evaluated. From the evaluation, it can be concluded that the expected purpose was achieved because there was a definite economic prospect to commercial plant. The benefit analysis of thirteen years operation of the uranium enrichment demonstration plant also provides a financial aspect of the uranium enrichment business. Therefore, the performance, price and reliability of the centrifuge is an important factor in the uranium enrichment business. And the continuous development of a centrifuge while considering balance with the development cost is necessary for the business in the future. (author)

  6. Uranium

    International Nuclear Information System (INIS)

    With the worldwide revival of nuclear energy comes the question of uranium reserves. For more than 20 years, nuclear energy has been neglected and uranium prospecting has been practically abandoned. Therefore, present day production covers only 70% of needs and stocks are decreasing. Production is to double by 2030 which represents a huge industrial challenge. The FBR-type reactors technology, which allows to consume the whole uranium content of the fuel, is developing in several countries and will ensure the long-term development of nuclear fission. However, the implementation of these reactors (the generation 4) will be progressive during the second half of the 21. century. For this reason an active search for uranium ores will be necessary during the whole 21. century to ensure the fueling of light water reactors which are huge uranium consumers. This dossier covers all the aspects of natural uranium production: mineralogy, geochemistry, types of deposits, world distribution of deposits with a particular attention given to French deposits, the exploitation of which is abandoned today. Finally, exploitation, ore processing and the economical aspects are presented. Contents: 1 - the uranium element and its minerals: from uranium discovery to its industrial utilization, the main uranium minerals (minerals with tetravalent uranium, minerals with hexavalent uranium); 2 - uranium in the Earth's crust and its geochemical properties: distribution (in sedimentary rocks, in magmatic rocks, in metamorphic rocks, in soils and vegetation), geochemistry (uranium solubility and valence in magmas, uranium speciation in aqueous solution, solubility of the main uranium minerals in aqueous solution, uranium mobilization and precipitation); 3 - geology of the main types of uranium deposits: economical criteria for a deposit, structural diversity of deposits, classification, world distribution of deposits, distribution of deposits with time, superficial deposits, uranium

  7. LEU fuel element produced by the Egyptian fuel manufacturing pilot plant

    International Nuclear Information System (INIS)

    The Egyptian Fuel Manufacturing Pilot Plant, FMPP, is a Material Testing Reactor type (MTR) fuel element facility, for producing the specified fuel elements required for the Egyptian Second Research Reactor, ETRR-2. The plant uses uranium hexafluoride (UF6, 19.75% U235 by wt) as a raw material which is processed through a series of the manufacturing, inspection and test plan to produce the final specified fuel elements. Radiological safety aspects during design, construction, operation, and all reasonably accepted steps should be taken to prevent or reduce the chance of accidents occurrence. (author)

  8. Uranium thorium dioxide fuel-cycle and economic analysis

    International Nuclear Information System (INIS)

    The fuel division of Framatome ANP (Advanced Nuclear Power) is performing a fuel-cycle analysis for uranium-thorium dioxide (U/Th) reactor fuel as part of a U.S. Department of Energy Nuclear Energy Research Initiative project titled, ''Advanced Proliferation Resistant, Lower Cost, Uranium-Thorium Dioxide Fuels for Light Water Reactor'', (DE-FC03-99SF21916). The objective is to evaluate the economic viability of the U/Th fuel cycle in commercial nuclear reactors operating in the U.S. This analysis includes formulating the evaluation methodology, validating the methodology via benchmark calculations, and performing a fuel-cycle analysis and corresponding economic evaluation. The APOLLO2-F computer program of Framatome ANP SCIENCE package was modified to incorporate the thorium decay chains and provide cross sections for the SCIENCE fuel-cycle analysis. A comparison and economic evaluation was made between UO2 and UO2/ThO2 fuel cycles in a typical 193-fuel assembly pressurized water reactor using reload batch sizes corresponding to batch average discharge burnups of 50, 70, and 90 GWd/mtHM. Results show an increase in front-end costs for the UO2/ThO2 cycles due primarily to the higher cost in separative work units for enriching the uranium to 19.5 wt% 235U. (author)

  9. Summary of uranium refining and conversion pilot plant at Ningyo-toge works

    International Nuclear Information System (INIS)

    In the Ningyo-toge works, Power Reactor and Nuclear Fuel Development Corp., the construction of the uranium refining and conversion pilot plant was completed, and the operation will be started after the various tests based on the related laws. As for the uranium refining in Japan, the PNC process by wet refining method has been developed since 1958. The history of the development is described. It was decided to construct the refining and conversion pilot plant with 200 t uranium/year capacity as the comprehensive result of the development. This is the amount sufficient to supply UF6 to the uranium enrichment pilot plant in Ningyo-toge. The building for the refining and conversion pilot plant is a three-story ferro-concrete building with the total floor area of about 13,000 m2. The raw materials are the uranium ore produced in Ningyo-toge and the yellow cakes from abroad. Uranyl sulfate solution is obtained by solvent extraction using an extraction tower or a mixer-settler. The following processes are electrolytic reduction, precipitation of uranium tetrafluoride, filtration, drying, dehydration and UF6 conversion. The fluorine for UF6 conversion is produced by the facility in the plant. The operation of the pilot plant will be started in the latter half of the fiscal year 1981, the batch operation is carrried out in 1982, and the continuous operation from 1983. (Kako, I.)

  10. Operative experience in first campaign for reprocessing of uranium-thorium Elk-River fuel elements

    International Nuclear Information System (INIS)

    The main characteristics which differentiates the ITREC pilot plant from the other reprocessing plants are summarized. The report describes: a brief history of the Uranium-Thorium Cycle Program developed in Italy since 1960; the results obtained during the cold tests in the Remote Refabrication Cell until 1974, year in which these activities have been stopped; the operating experience in the Campaign for reprocessing of 20 Uranium-Thorium Elk-River irradiated fuel elements. In particular the results of the following operations are described: dismantling fuel elements and pool water treatment, chopping, dissolution, extraction in a Thorex acid-deficient flow-sheet. Some works of maintenance and modification during hot operations (dismantling machine, replacement of the high level waste evaporators) are also presented. The planned utilization of the ITREC plant is indicated. (author)

  11. Melvin Calvin: Fuels from Plants

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, S.E.; Otvos, J.W.

    1998-11-24

    A logical extension of his early work on the path of carbon during photosynthesis, Calvin's studies on the production of hydrocarbons by plants introduced many in the scientific and agricultural worlds to the potential of renewable fuel and chemical feedstocks. He and his co-workers identified numerous candidate compounds from plants found in tropical and temperate climates from around the world. His travels and lectures concerning the development of alternative fuel supplies inspired laboratories worldwide to take up the investigation of plant-derived energy sources as an alternative to fossil fuels.

  12. Development of metal uranium fuel and testing of construction materials (I-VI); Part I

    International Nuclear Information System (INIS)

    This project includes the following tasks: Study of crystallisation of metal melt and beta-alpha transforms in uranium and uranium alloys; Study of the thermal treatment influence on phase transformations and texture in uranium alloys; Radiation damage of metal uranium; Project related to irradiation of metal uranium in the reactor; Development of fuel element for nuclear reactors

  13. Uranium for Nuclear Power: Resources, Mining and Transformation to Fuel

    International Nuclear Information System (INIS)

    Uranium for Nuclear Power: Resources, Mining and Transformation to Fuel discusses the nuclear industry and its dependence on a steady supply of competitively priced uranium as a key factor in its long-term sustainability. A better understanding of uranium ore geology and advances in exploration and mining methods will facilitate the discovery and exploitation of new uranium deposits. The practice of efficient, safe, environmentally-benign exploration, mining and milling technologies, and effective site decommissioning and remediation are also fundamental to the public image of nuclear power. This book provides a comprehensive review of developments in these areas: • Provides researchers in academia and industry with an authoritative overview of the front end of the nuclear fuel cycle • Presents a comprehensive and systematic coverage of geology, mining, and conversion to fuel, alternative fuel sources, and the environmental and social aspects • Written by leading experts in the field of nuclear power, uranium mining, milling, and geological exploration who highlight the best practices needed to ensure environmental safety

  14. Conceptual design of KALIMER uranium metallic fueled core

    International Nuclear Information System (INIS)

    As a part of the core design development of KALIMER(150 MWe), the KALIMER core design which uses U-Zr binary fuel not in excess of 20% enrichment was performed. Starting from the former uranium metallic fueled core design, a more economic and safer equilibrium core design was first established based on extensive researches for the possible enrichment gains over various design options and in-core fuel management strategies. Further optimization to extend fuel discharge burnup has been achieved by employing strategic loading schemes for initial and transition cycles to reach the equilibrium cycle early. The core performance analysis based on a once-through equilibrium fuel cycle scenario shows that the core has an average breeding ratio of 0.67 and core average discharge burnup of 61.6 MWD/kg. The negative sodium void reactivity over the core shows a beneficial potential to assure inherent safety characteristics. When comparing with conventional plutonium metallic fueled cores of the same power level, the present KALIMER uranium metallic fueled core has an increased physical core size to meet the enrichment restriction, and, as a result, a lower power density to realize the minimum one-year cycle operation. The KALIMER uranium metallic fueled core characterized by its negative sodium void reactivity and low power density can be operated with maximizing its core safety characteristics as a first generation LMR. The present uranium metallic fueled core allows an easy replacement with different fuel compositions by its demands, with the accumulation of operation experience and design data verification. (author). 34 refs., 34 tabs., 12 figs.

  15. The SLOWPOKE-2 reactor with low enrichment uranium oxide fuel

    International Nuclear Information System (INIS)

    A SLOWPOKE-2 reactor core contains less than 1 kg of highly enriched uranium (HEU) and the proliferation risk is very low. However, to overcome proliferation concerns a new low enrichment uranium (LEU) fuelled reactor core has been designed. This core contains approximately 180 fuel elements based on the Zircaloy-4 clad UOsub(2) CANDU fuel element, but with a smaller outside diameter. The physics characteristics of this new reactor core ensure the inherent safety of the reactor under all conceivable conditions and thus the basic SLOWPOKE safety philosophy which permits unattended operation is not affected

  16. Integrated design of SIGMA uranium enrichment plants

    International Nuclear Information System (INIS)

    In the present work, we describe a preliminary analysis of the design feedbacks in a Uranium Enrichment Plant, using the SIGMA concept. Starting from the result of this analysis, a computer code has been generated, which allows finding the optimal configurations of plants, for a fixed production rate. The computer code developed includes the model of the Thermohydraulic loop of a SIGMA module. The model contains numerical calculations of the main components of the circuit. During the calculations, the main components are dimensioned, for a posterior cost compute. The program also makes an estimation of the enrichment gain of the porous membrane, for each separation stage. Once the dimensions of the main components are known, using the enrichment cascade calculation, the capital and operation costs of the plant could be determined. At this point it is simple to calculate a leveled cost of the Separative Work Unit (SWU). A numerical optimizer is also included in the program. This optimizer finds the optimal cascade configuration, for a given set of design parameters. The whole-integrated program permits to investigate in detail the feedback in the component design. Therefore, the sensibility of the more relevant parameters can be computed, with respect of the economical variables of the plant. (author)

  17. Irradiation of the experimental fuel assemblies with uranium-plutonium fuel in the BN-600 reactor

    International Nuclear Information System (INIS)

    Design features of experimental fuel assemblies (EFA) with uranium-plutonium mixed oxide fuel specific aspects of their arrangement within the BN-600 reactor core, conditions and basic results of EFA with the fuel mentioned in the BN-600 reactor are described

  18. Biosolubilization of uranyl ions in uranium ores by hydrophyte plants

    International Nuclear Information System (INIS)

    This paper investigated the bioleaching of uranyl ions from uranium ores, in aqueous medium by hydrophyte plants: Lemna minor, Azolla caroliniana and Elodea canadensis under different experimental conditions. The oxidation of U(IV) to U(VI) species was done by the atomic oxygen generated in the photosynthesis process by the aquatic plants in the solution above uranium ores. Under identical experimental conditions, the capacity of bioleaching of uranium ores decreases according to the following series: Lemna minor > Elodea canadensis > Azolla caroliniana. The results of IR spectra suggest the possible use of Lemna minor and Elodea canadensis as a biological decontaminant of uranium containing wastewaters. (author)

  19. Removal of uranium from uranium plant wastewater using zero-valent iron in an ultrasonic field

    Energy Technology Data Exchange (ETDEWEB)

    Li, Jing; Zhang, Libo; Peng, Jinhui; Ma, Aiyuan; Xia, Hong Ying; Guo, Wen Qian; Yu, Xia [Yunnan Provincial Key Laboratory of Intensification Metallurgy, Kunming (China); Hu, Jinming; Yang, Lifeng [Nuclear Group Two Seven Two Uranium Industry Limited Liability Company, Hengyang (China)

    2016-06-15

    Uranium removal from uranium plant wastewater using zero-valent iron in an ultrasonic field was investigated. Batch experiments designed by the response surface methodology (RSM) were conducted to study the effects of pH, ultrasonic reaction time, and dosage of zero-valent iron on uranium removal efficiency. From the experimental data obtained in this work, it was found that the ultrasonic method employing zero-valent iron powder effectively removes uranium from uranium plant wastewater with a uranium concentration of 2,772.23 μg/L. The pH ranges widely from 3 to 7 in the ultrasonic field, and the prediction model obtained by the RSM has good agreement with the experimental results.

  20. Feasibility of Low Enriched Uranium Fuel for Space Nuclear Propulsion

    International Nuclear Information System (INIS)

    The purpose of this initial study is to create a baseline with which to perform further analysis and to build a solid understanding of the neutronic characteristics of a solid core for the nuclear thermal rocket. Once consistency with work done at Idaho National Laboratory (INL) is established, this paper will provide a study of other fuel types, such as low and medium-enriched uranium fuels. This paper will examine how the implementation of each fuel type affects the multiplication factor of the reactor, and will then explore different possibilities for alterations needed to accommodate their successful usage. The reactor core analysis was done using the MCNP5 code. While this study has not shown that the SNRE can be easily retrofitted for low-enriched U fuel, it has made a detailed study of the SNRE, and identified the difficulties of the implementation of low-enriched fuels in small nuclear rockets. These difficulties are the need for additional moderation and fuel mass in order to achieve a critical mass. Neither of these is insurmountable. Future work includes finding the best method by which to increase the internal moderation of the reactor balanced with appropriate sizing to prevent neutron leakage. Both of these are currently being studied. This paper will present a study of the Small Nuclear Rocket Engine (SNRE) and the feasibility of using low enriched Uranium (LEU) instead of the traditional high enriched Uranium (HEU) fuels

  1. Feasibility of Low Enriched Uranium Fuel for Space Nuclear Propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Venneri, Paolo; Kim, Yonghee [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2013-05-15

    The purpose of this initial study is to create a baseline with which to perform further analysis and to build a solid understanding of the neutronic characteristics of a solid core for the nuclear thermal rocket. Once consistency with work done at Idaho National Laboratory (INL) is established, this paper will provide a study of other fuel types, such as low and medium-enriched uranium fuels. This paper will examine how the implementation of each fuel type affects the multiplication factor of the reactor, and will then explore different possibilities for alterations needed to accommodate their successful usage. The reactor core analysis was done using the MCNP5 code. While this study has not shown that the SNRE can be easily retrofitted for low-enriched U fuel, it has made a detailed study of the SNRE, and identified the difficulties of the implementation of low-enriched fuels in small nuclear rockets. These difficulties are the need for additional moderation and fuel mass in order to achieve a critical mass. Neither of these is insurmountable. Future work includes finding the best method by which to increase the internal moderation of the reactor balanced with appropriate sizing to prevent neutron leakage. Both of these are currently being studied. This paper will present a study of the Small Nuclear Rocket Engine (SNRE) and the feasibility of using low enriched Uranium (LEU) instead of the traditional high enriched Uranium (HEU) fuels.

  2. 10 CFR 51.51 - Uranium fuel cycle environmental data-Table S-3.

    Science.gov (United States)

    2010-01-01

    ... uranium mining and milling, the production of uranium hexafluoride, isotopic enrichment, fuel fabrication... 10 percent of 10 CFR 20 for total processing 26 annual fuel requirements for model LWR. Fission and... 10 Energy 2 2010-01-01 2010-01-01 false Uranium fuel cycle environmental data-Table S-3....

  3. Optimization of fuel cycle strategies with constraints on uranium availability

    International Nuclear Information System (INIS)

    Optimization of nuclear reactor and fuel cycle strategies is studied under the influence of reduced availability of uranium. The analysis is separated in two distinct steps. First, the global situation is considered within given high and low projections of the installed capacity up to the year 2025. Uranium is regarded as an exhaustible resource whose production cost would increase proportionally to increasing cumulative exploitation. Based on the estimates obtained for the uranium cost, a global strategy is derived by splitting the installed capacity between light water reactor (LWR) once-through, LWR recycle, and fast breeder reactor (FBR) alternatives. In the second phase, the nuclear program of an individual utility is optimized within the constraints imposed from the global scenario. Results from the global scenarios indicate that in a reference case the uranium price would triple by the year 2000, and the price escalation would continue throughout the planning period. In a pessimistic growth scenario where the global nuclear capacity would not exceed 600 GW(electric) in 2025, the uranium price would almost double by 2000. In both global scenarios, FBRs would be introduced, in the reference case after 2000 and in the pessimistic case after 2010. In spite of the increases in the uranium prices, the levelized power production cost would increase only by 45% up to 2025 in the utility case provided that the plutonium is incinerated as a substitute fuel

  4. Fuel loading and homogeneity analysis of HFIR design fuel plates loaded with uranium silicide fuel

    International Nuclear Information System (INIS)

    Twelve nuclear reactor fuel plates were analyzed for fuel loading and fuel loading homogeneity by measuring the attenuation of a collimated X-ray beam as it passed through the plates. The plates were identical to those used by the High Flux Isotope Reactor (HFIR) but were loaded with uranium silicide rather than with HFIR's uranium oxide fuel. Systematic deviations from nominal fuel loading were observed as higher loading near the center of the plates and underloading near the radial edges. These deviations were within those allowed by HFIR specifications. The report begins with a brief background on the thermal-hydraulic uncertainty analysis for the Advanced Neutron Source (ANS) Reactor that motivated a statistical description of fuel loading and homogeneity. The body of the report addresses the homogeneity measurement techniques employed, the numerical correction required to account for a difference in fuel types, and the statistical analysis of the resulting data. This statistical analysis pertains to local variation in fuel loading, as well as to ''hot segment'' analysis of narrow axial regions along the plate and ''hot streak'' analysis, the cumulative effect of hot segment loading variation. The data for all twelve plates were compiled and divided into 20 regions for analysis, with each region represented by a mean and a standard deviation to report percent deviation from nominal fuel loading. The central regions of the plates showed mean values of about +3% deviation, while the edge regions showed mean values of about -7% deviation. The data within these regions roughly approximated random samplings from normal distributions, although the chi-square (χ2) test for goodness of fit to normal distributions was not satisfied

  5. Fabrication procedures for manufacturing high uranium concentration dispersion fuel elements

    International Nuclear Information System (INIS)

    IPEN-CNEN/SP developed the technology to produce the dispersion type fuel elements for research reactors and made it available for routine production. Today, the fuel produced in IPEN-CNEN/SP is limited to the uranium concentration of 3.0 gU/cm3 for U3Si2-Al dispersion-based and 2.3 gU/cm3 for U3O8-Al dispersion. The increase of uranium concentration in fuel plates enables the reactivity of the reactor core reactivity to be higher and extends the fuel life. Concerning technology, it is possible to increase the uranium concentration in the fuel meat up to the limit of 4.8 gU/cm3 in U3Si2-Al dispersion and 3.2 gU/cm3 U3O8-Al dispersion. These dispersions are well qualified worldwide. This work aims to develop the manufacturing process of both fuel meats with high uranium concentrations, by redefining the manufacturing procedures currently adopted in the Nuclear Fuel Center of IPEN-CNEN/SP. Based on the results, it was concluded that to achieve the desired concentration, it is necessary to make some changes in the established procedures, such as in the particle size of the fuel powder and in the feeding process inside the matrix, before briquette pressing. These studies have also shown that the fuel plates, with a high concentration of U3Si2-Al, met the used specifications. On the other hand, the appearance of the microstructure obtained from U3O8-Al dispersion fuel plates with 3.2 gU/cm3 showed to be unsatisfactory, due to the considerably significant porosity observed. The developed fabrication procedure was applied to U3Si2 production at 4.8 gU/cm3, with enriched uranium. The produced plates were used to assemble the fuel element IEA-228, which was irradiated in order to check its performance in the IEA-R1 reactor at IPEN-CNEN/SP. These new fuels have potential to be used in the new Brazilian Multipurpose Reactor - RMB. (author)

  6. Estimation of uranium in some edible and commercial plants

    International Nuclear Information System (INIS)

    The trace contents of uranium have been estimated in some edible and commercial plants by PTA (particle track analysis) method. The groups of food plants studied are cereals, pulses, underground vegetables, leafy vegetables, and fruit vegetables. The commercial plants and ingredients taken are betel leaves, tobacco leaves, areca nuts, and lime. Among the different samples studied, the average uranium content, in general, is found to vary from 0.25 to 2.67 ppm. (author). 10 refs., 2 tabs., 1 fig

  7. Estimation of Uranium in Some Edible and Commercial Plants

    Directory of Open Access Journals (Sweden)

    S. Choudhury

    1992-10-01

    Full Text Available The trace contents of uranium have been estimated in some edible and commercial plants by PTA method. The groups of food plants studied are cereals, pulses, underground vegetables, leafy vegetables, and fruit vegetables. The commercial plants and ingredients taken are betel leaves, tobacco leaves, areca nuts, and lime. Among the different samples studied, the average uranium content, in general, is found to vary from 0.25 to 2.67 ppm

  8. Reduction of uranium in disposal conditions of spent nuclear fuel

    International Nuclear Information System (INIS)

    This literature study is a summary of publications, in which the reduction of uranium by iron has been investigated in anaerobic groundwater conditions or in aqueous solution in general. The basics of the reduction phenomena and the oxidation states, complexes and solubilities of uranium and iron in groundwaters are discussed as an introduction to the subject, as well as, the Finnish disposal concept of spent nuclear fuel. The spent fuel itself mainly (∼96 %) consists of a sparingly soluble uranium(IV) dioxide, UO2(s), which is stable phase in the anticipated reducing disposal conditions. If spent fuel gets in contact with groundwater, oxidizing conditions might be induced by the radiolysis of water, or by the intrusion of oxidizing glacial melting water. Under these conditions, the oxidation and dissolution of uranium dioxide to more soluble U(VI) species could occur. This could lead to the mobilization of uranium and other components of spent fuel matrix including fission products and transuranium elements. The reduction of uranium back to oxidation state U(IV) can be considered as a favourable immobilization mechanism in a long-term, leading to precipitation due to the low solubility of U(IV) species. The cast iron insert of the disposal canister and its anaerobic corrosion products are the most important reductants under disposal conditions, but dissolved ferrous iron may also function as reductant. Other iron sources in the buffer or near-field rock, are also considered as possible reductants. The reduction of uranium is a very challenging phenomenon to investigate. The experimental studies need e.g. well-controlled anoxic conditions and measurements of oxidation states. Reduction and other simultaneous phenomena are difficult to distinghuish. The groundwater conditions (pH, Eh and ions) influence on the prevailing complexes of U and Fe and on forming corrosion products of iron and, thus they determine also the redox chemistry. The partial reduction of

  9. Challenges in the front end of the uranium fuel cycle

    International Nuclear Information System (INIS)

    The long-term fundamentals for nuclear remain strong. Climate change and clean air concerns remain high on the agenda of national energy policies, as both developing and developed economies pursue a strategy of energy diversity and energy security. A global industry of 435 reactors is expected to grow to more than 639 reactors within the next 20 years with the potential for even more rapid expansion. This nuclear generating capacity relies on an international fuel cycle that can ensure stable and secure supply for decades to come. As the first step in the fuel cycle, the uranium industry has received various price signals over the past 5 decades, from the birth of an industry with strong demand and stock pile building and the associated robust pricing and new production stimulation, to an industry in decline and a period marked by liquidation of large inventories, to the recent resurgence of nuclear and the associated uranium price signals. In many ways, understanding the current uranium environment and the outlook for the industry requires some understanding of these phases of nuclear. The global nuclear fleet today needs about 65,000 tonnes of uranium per year to meet reactor feed requirements. Primary production meets about two thirds of this requirement while the remainder is drawn from secondary supply. Secondary supply can essentially be described as stockpiles of previously produced uranium. However, secondary supplies are finite and more primary production will be needed. From a long-term perspective, there is no question that there are sufficient uranium resources to support the nuclear industry for many years to come. The IAEA's 'Red Book' estimates that more than 5 million tonnes of known resources could potentially be developed at today's prices. This is enough to supply the global reactor fleet for almost 80 years at current usage rates. Recently higher uranium prices have resulted in some production increases although the rate of growth has been held

  10. Irradiation behavior of experimental miniature uranium silicide fuel plates

    International Nuclear Information System (INIS)

    Uranium silicides, because of their relatively high uranium density, were selected as candidate dispersion fuels for the higher fuel densities required in the Reduced Enrichment Research and Test Reactor (RERTR) Program. Irradiation experience with this type of fuel, however, was limited to relatively modest fission densities in the bulk form, on the order of 7 x 1020 cm-3, far short of he approximately 20 x 1020 cm-3 goal established for the RERTR Program. The purpose of the irradiation experiments on silicide fuels in the ORR, therefore, was to investigate the intrinsic irradiation behavior of uranium silicide as a dispersion fuel. Of particular interest was the interaction between the silicide particles and the aluminum matrix, the swelling behavior of the silicide particles, and the maximum volume fraction of silicide particles that could be contained in the aluminum matrix. The first group of experimental 'mini' fuel plates have recently reached the program's goal burnup and are in various stages of examination. Although the results to date indicate some limitations, it appears that within the range of parameters examined thus far the uranium silicide dispersion holds promise for satisfying most of the needs of the RERTR Program. The twelve experimental silicide dispersion fuel plates that were irradiated to approximately their goal exposure show the 30-vol % U3Si-Al plates to be in a stage of relatively rapid fission-gas-driven swelling at a fission density of 2 x 1020 cm-3. This fuel swelling will likely result in unacceptably large plate-thickness increases. The U3Si plates appear to be superior in this respect; however, they, too, are starting to move into the rapid fuel-swelling stage. Analysis of the currently available post irradiation data indicates that a 40-vol % dispersed fuel may offer an acceptable margin to the onset of unstable thickness changes at exposures of 2 x 1021 fission/cm3. The interdiffusion between fuel and matrix aluminum was found

  11. Evaluation of the uranium immobilization potential of vetiver plants grown on processed solid waste of uranium industry of Jaduguda, India

    International Nuclear Information System (INIS)

    Remediation of contaminated sites using specific plant or plant groups may offer a cheap, renewable and promising technique to minimize the long-term ecological adverse impact of the waste disposal. The major components of process waste of uranium industry are uranium series radionuclides, heavy metals inherently present in the ore, chemical additives and residual uranium. Among the radionuclides quantitative content of residual uranium is highest in the disposed process waste of uranium mill. In view of this fact experiments were conducted to study the uranium immobilization potential of a phytoremediator that can grow and survive in the complex tailings (fine solid process waste) environment. Vetiver grass (Chrysopogon zizanioides (L.) Nash) was selected for translocation and immobilization studies of uranium. The grass was planted in uranium mill tailing ponds at Jaduguda, Jharkhand, India and periodic sampling was carried out to investigate the extent of uranium uptake. The acid aliquot of dry or wet ash samples of plant and soil were subjected to solvent extraction followed by UV-Fluorimetry for uranium estimation. It has been observed that the grass could immobilize up to 8 ppm uranium within 6 months after planting. Uranium is preferably immobilized at the root and translocation of uranium to upper plant parts (shoot) is low compared to roots. The uranium uptake got saturated after a particular concentration range. The increased level of uranium in the soil covering of tailings needs further investigation. (author)

  12. Guidebook on design, construction and operation of pilot plants for uranium ore processing

    International Nuclear Information System (INIS)

    The design, construction and operation of a pilot plant are often important stages in the development of a project for the production of uranium concentrates. Since building and operating a pilot plant is very costly and may not always be required, it is important that such a plant be built only after several prerequisites have been met. The main purpose of this guidebook is to discuss the objectives of a pilot plant and its proper role in the overall project. Given the wide range of conditions under which a pilot plant may be designed and operated, it is not possible to provide specific details. Instead, this book discusses the rationale for a pilot plant and provides guidelines with suggested solutions for a variety of problems that may be encountered. This guidebook is part of a series of Technical Reports on uranium ore processing being prepared by the IAEA's Division of Nuclear Fuel Cycle and Waste Management. 42 refs, 7 figs, 3 tabs

  13. Treatment of uranium bearing waste arising from solvent recovery unit of uranium processing plant

    International Nuclear Information System (INIS)

    During the regeneration of tributyl phosphate in uranium plant, a sizable volume of liquid waste containing about 70 mg/l of uranium, along with high concentrations of nitrates and carbonates, is generated. Laboratory studies revealed that the waste was not amenable to conventional treatment methods, including co-precipitation, owing to high concentration of carbonates, with which uranium forms a stable carbonato complex. Various commmercially available strongly basic anion exchangers were evaluated for the uptake of uranium from the waste under static conditions. Column studies, employing the anion-exchange resin which has shown the highest uptake, were carried out. These studies reveal the application potential of ion-exchange process not only in the treatment of uranium bearing wastes but also in the recovery of uranium. (author). 8 refs., 4 figs., 5 tabs

  14. Status of the atomized uranium silicide fuel development at KAERI

    Energy Technology Data Exchange (ETDEWEB)

    Kim, C.K.; Kim, K.H.; Park, H.D.; Kuk, I.H. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-08-01

    While developing KMRR fuel fabrication technology an atomizing technique has been applied in order to eliminate the difficulties relating to the tough property of U{sub 3}Si and to take advantage of the rapid solidification effect of atomization. The comparison between the conventionally comminuted powder dispersion fuel and the atomized powder dispersion fuel has been made. As the result, the processes, uranium silicide powdering and heat treatment for U{sub 3}Si transformation, become simplified. The workability, the thermal conductivity and the thermal compatibility of fuel meat have been investigated and found to be improved due to the spherical shape of atomized powder. In this presentation the overall developments of atomized U{sub 3}Si dispersion fuel and the planned activities for applying the atomizing technique to the real fuel fabrication are described.

  15. Fuel performance experience with slightly-enriched uranium

    International Nuclear Information System (INIS)

    There is an economic incentive associated with an increase in the average burnup of fuel discharged from CANDU reactors. The slightly enriched uranium fuel cycle provides a path by which increased fuel burnup can be attained in the short term. This paper discusses the effects upon fuel performance of increases in burnup, the existing base of pertinent data and the additional work that is required to verify the technology. Areas of fuel performance that are of particular importance at high burnup are fission gas release and power ramp behaviour. Fuel performance models will also require development and verification. The development program is in two phases, with the first phase leading to intermediate burnups and the second phase providing higher burnups

  16. Phenomenology of uranium-plutonium homogenization in nuclear fuels

    International Nuclear Information System (INIS)

    The uranium and plutonium cations distribution in mixed oxide fuels (U1-y Puy)O2 with y ≤ 0.1 has been studied in laboratory with industrial fabrication methods. Our experiences has showed a slow cations migration. In the substoichiometry (UPu)O2-x the diffusion is in connection with the plutonium valence which is an indicator of the oxidoreduction state of the crystal lattice. The plutonium valence is in connection with the oxygen ion deficit in order to compensate the electrical charge. The oxygen ratio of the solid depends of the oxygen partial pressure prevailing at the time of product elaboration but it can be modified by impurities. These impurities permit to increase or decrease the fuel characteristics and performances. An homogeneity analysis methodology is proposed, its objective is to classify the mixed oxide fuels according to the uranium and plutonium ions distribution

  17. Measurement of thermal conductivity of uranium silicide - aluminum dispersion fuel

    International Nuclear Information System (INIS)

    In conjunction with reducing enrichment program for JMTR, thermal conductivity of uranium silicide - aluminum (U3Si2-Al) dispersion fuel was measured in the temperature range of 25degC ∼ 400degC for the safety evaluation of low enriched uranium fuel. Since thermal conductivity is determined as the product of thermal diffusivity, heat capacity and density, these three properties were individually measured. Thermal diffusivity and heat capacity of the specimen were measured by the laser flash method. Temperature dependence of density was obtained by measuring the thermal linear expansion with differential dilatometer. Obtained results show that conductivity of the U3Si2-Al dispersion fuel slightly increases as temperature increases, and tends to reach the maximum around 300degC. (author)

  18. Chemical treatment of ammonium fluoride solution in uranium reconversion plant

    International Nuclear Information System (INIS)

    A chemical procedure is described for the treatment of the filtrate, produced from the transformation of uranium hexafluoride (U F6) into ammonium uranyl carbonate (AUC). This filtrate is an intermediate product in the U F6 to uranium dioxide (U O2) reconversion process. The described procedure recovers uranium as ammonium peroxide fluoro uranate (APOFU) by precipitation with hydrogen peroxide (H2 O2), and as later step, its calcium fluoride (CaF2) co-precipitation. The recovered uranium is recycled to the AUC production plant. (author)

  19. The first six years of the Chemwes uranium plant

    International Nuclear Information System (INIS)

    The Stilfontein and Buffelsfontein Gold Mines, near Klerksdorp in the Transvaal, had accumulated a large amount of uranium-containing residue and, when the price of uranium rose in the 1970s, consideration was given to the possible recovery of this uranium. Preliminary tests showed that concentration of the uranium prior to leaching would not be economic. However, the pyrite in the residue could be concentrated by flotation, and the flotation concentrate could be roasted to provide both enough acid for leaching the uranium and a calcine from which the gold could be recovered. The feasibility study showed that a uranium operation of 270 kt per month would be most economically attractive, and a plant of that size was accordingly designed and built. In the first six years of its existence, the plant treated over 20 Mt of residue and produced about 3,5 kt of uranium oxide. During that time, the plant was continually being improved to make it more reliable and cost-efficient. This paper analyses the operation of the plant during its first six years from the viewpoints of its mechanical, process, and economic performance. The criteria on which the selection of the process was based are reviewed and compared with the actual performance of the plant, emphasis being placed on the leaching, solid-liquid separation, recovery, and purification stages

  20. Determination of oxygen in mixed uranium-plutonium carbide fuels

    International Nuclear Information System (INIS)

    Determination of oxygen in mixed uranium-plutonium carbide fuels is made by inert gas fusion-coulometry. To minimize oxygen contamination during sample preparation, the sample is crushed, weighted and sealed air-tight in a platinum capsule in an argon gas atmosphere glove box. The true oxygen content is estimated by subtracting the oxygen contamination from the oxygen determined. Routine analysis of 32 samples of mixed uranium-plutonium carbides is performed with a coefficient of variation of 1.6%. (author)

  1. Study of internal exposure to uranium compounds in fuel fabrication plants in Brazil; Estudo da exposicao interna a compostos de uranio na fabricacao do elemento combustivel nuclear no Brasil

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Maristela Souza

    2006-07-01

    The International Commission on Radiological Protection (ICRP) Publication 66 and Supporting Guidance 3) strongly recommends that specific information on lung retention parameters should be used in preference to default values wherever appropriate, for the derivation of effective doses and for bioassay interpretation of monitoring data. A group of 81 workers exposed to UO{sub 2} at the fuel fabrication facility in Brazil was selected to evaluate the committed effective dose. The workers were monitored for determination of uranium content in the urinary and faecal excretion. The contribution of intakes by ingestion and inhalation were assessed on the basis of the ratios of urinary to fecal excretion. For the selected workers it was concluded that inhalation dominated intake. According to ICRP 66, uranium oxide is classified as insoluble Type S compound. The ICRP Supporting Guidance 3 and some recent studies have recommended specific lung retention parameters to UO{sub 2}. The solubility parameters of the uranium oxide compound handled by the workers at the fuel fabrication facility in Brazil was evaluated on the basis of the ratios of urinary to fecal excretion. Excretion data were corrected for dietary intakes. This paper will discuss the application of lung retention parameters recommended by the ICRP models to these data and also the dependence of the effective committed dose on the lung retention parameters. It will also discuss the problems in the interpretation of monitoring results, when the worker is exposed to several uranium compounds of different solubilities. (author)

  2. Dual fuel gradients in uranium silicide plates

    Energy Technology Data Exchange (ETDEWEB)

    Pace, B.W. [Babock and Wilcox, Lynchburg, VA (United States)

    1997-08-01

    Babcock & Wilcox has been able to achieve dual gradient plates with good repeatability in small lots of U{sub 3}Si{sub 2} plates. Improvements in homogeneity and other processing parameters and techniques have allowed the development of contoured fuel within the cladding. The most difficult obstacles to overcome have been the ability to evaluate the bidirectional fuel loadings in comparison to the perfect loading model and the different methods of instilling the gradients in the early compact stage. The overriding conclusion is that to control the contour of the fuel, a known relationship between the compact, the frames and final core gradient must exist. Therefore, further development in the creation and control of dual gradients in fuel plates will involve arriving at a plausible gradient requirement and building the correct model between the compact configuration and the final contoured loading requirements.

  3. Plant Design Nuclear Fuel Element Production Capacity Optimization to Support Nuclear Power Plant in Indonesia

    International Nuclear Information System (INIS)

    The optimization production capacity for designing nuclear fuel element fabrication plant in Indonesia to support the nuclear power plant has been done. From calculation and by assuming that nuclear power plant to be built in Indonesia as much as 12 NPP and having capacity each 1000 MW, the optimum capacity for nuclear fuel element fabrication plant is 710 ton UO2/year. The optimum capacity production selected, has considered some aspects such as fraction batch (cycle, n = 3), length of cycle (18 months), discharge burn-up value (Bd) 35,000 up 50,000 MWD/ton U, enriched uranium to be used in the NPP (3.22 % to 4.51 %), future market development for fuel element, and the trend of capacity production selected by advances country to built nuclear fuel element fabrication plant type of PWR. (author)

  4. Complex approach to modelling of evolution of multi-recycled uranium isotope composition in closed fuel cycle of light water reactors

    International Nuclear Information System (INIS)

    Irradiated uranium fuel contains more then 90% of uranium, but at the present time the level of nuclear science and technology makes us mostly to postpone the using of recycling uranium up to far future or not at all. Only small number of states has the experience in using of recycling uranium and this experience is limited to one recycle. At the same time multiple recycling decreases the need in uranium mining and improves the utilization of uranium resources. Calculations result that in VVER-1000 irradiated fuel residual 235U concentration remains more than in natural uranium up to burnup level ∝ 60 MW.day/kg of heavy metals (h.m.). Utilization of reprocessed uranium as a source is more complicated due to 232U and 236U isotopes presenting in irradiated fuel. Some other uranium isotopes effect on the fuel reprocessing and fabrication is significantly less. 232U effect on neutron physical parameters is negligible due to very small concentration in reprocessed uranium. However introduction of this isotope may lead to an increase radiation dose rate to personnel because the hard gamma rays from its decay daughters. To limit this dose rate at the fuel fabrication plant 232U concentration in reprocessed uranium was restricted at the level 2.10-7 wt. %. Taking into account the enhanced technologies of fuel pin fabrication this restriction can become softer. 236U is a parasitic neutron absorber and to compensate this effect fuel with recycled uranium must be enriched more than that free from 236U. 234U concentration in reprocessed uranium is relatively small in comparison with 236U and need not compensation, but in the future this option can become necessary. (orig.)

  5. Kinetic parameters of a material test research reactor fueled with various low enriched uranium dispersion fuels

    International Nuclear Information System (INIS)

    The effects of using different low enriched uranium fuels, having same uranium density, on the kinetic parameters of a material test research reactor were studied. For this purpose, the original aluminide fuel (UAlx-Al) containing 4.40 gU/cm3 of an MTR was replaced with silicide (U3Si-Al and U3Si2-Al) and oxide (U3O8-Al) dispersion fuels having the same uranium density as of the original fuel. Simulations were carried out to calculate prompt neutron generation time, effective delayed-neutron fraction, core excess reactivity and neutron flux spectrum. Nuclear reactor analysis codes including WIMS-D4 and CITATION were used to carry out these calculations. It was observed that both the silicide fuels had the same prompt neutron generation time 0.02% more than that of the original aluminide fuel, while the oxide fuel had a prompt neutron generation time 0.05% less than that of the original aluminide fuel. The effective delayed-neutron fraction decreased for all the fuels; the decrease was maximum at 0.06% for U3Si2-Al followed by 0.03% for U3Si-Al, and 0.01% for U3O8-Al fuel. The U3O8-Al fueled reactor gave the maximum ρexcess at BOL which was 21.67% more than the original fuel followed by U3Si-Al which was 2.55% more, while that of U3Si2-Al was 2.50% more than the original UAlx-Al fuel. The neutron flux of all the fuels was more thermalized, than in the original fuel, in the active fuel region of the core. The thermalization was maximum for U3O8-Al followed by U3Si-Al and then U3Si2-Al fuel.

  6. Plate-shaped high power nuclear fuel element containing low enrichment uranium and its preparation

    International Nuclear Information System (INIS)

    The present invention provides a plate-shaped high power nuclear fuel element containing low enrichment uranium (5 to 20 percent by weight uranium235 in the uranium component) as the fissionable material, the fuel element essentially comprising a plate of UAl4 provided with a sheath (clad) of aluminum or an aluminum alloy and impurities inherent to the manufacturing process. (DG)

  7. 75 FR 44817 - Notice of Availability of Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding...

    Science.gov (United States)

    2010-07-29

    ... amended. The introduction of uranium hexafluoride into any module of the National Enrichment Facility is... COMMISSION Notice of Availability of Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding...: Ty Naquin, Project Manager, Uranium Enrichment Branch, Division of Fuel Cycle Safety and...

  8. 76 FR 67765 - Notice of Availability of Uranium Enrichment Fuel Cycle Facility's Inspection Reports Regarding...

    Science.gov (United States)

    2011-11-02

    ... Energy Act of 1954, as amended. The introduction of uranium hexafluoride into any module of the National... COMMISSION Notice of Availability of Uranium Enrichment Fuel Cycle Facility's Inspection Reports Regarding... CONTACT: Gregory Chapman, Project Manager, Uranium Enrichment Branch, Division of Fuel Cycle Safety...

  9. 77 FR 65729 - Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services LLC...

    Science.gov (United States)

    2012-10-30

    ... Act of 1954, as amended. The introduction of uranium hexafluoride into any module of the National... COMMISSION Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services LLC... Regulatory Commission Brian W. Smith, Chief, Uranium Enrichment Branch, Division of Fuel Cycle Safety...

  10. 78 FR 23312 - Uranium Enrichment Fuel Cycle Inspection Reports Regarding Louisiana Energy Services, National...

    Science.gov (United States)

    2013-04-18

    ... introduction of uranium hexafluoride (UF 6 ) into cascades numbered 2.9, 2.10, 2.11, 2.12, 3.1, 3.2, 3.3, 3.4... COMMISSION Uranium Enrichment Fuel Cycle Inspection Reports Regarding Louisiana Energy Services, National..., Uranium Enrichment Branch, Division of Fuel Cycle Safety, and Safeguards Office of Nuclear Material...

  11. 77 FR 18272 - Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services LLC...

    Science.gov (United States)

    2012-03-27

    ... 1954, as amended. The introduction of uranium hexafluoride into any module of the National Enrichment... COMMISSION Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services LLC.... Brian W. Smith, Chief, Uranium Enrichment Branch, Division of Fuel Cycle Safety and Safeguards,...

  12. Quantifying Tc-99 contamination in a fuel fabrication plant - 59024

    International Nuclear Information System (INIS)

    The Springfields facility manufactures nuclear fuel products for the UK's nuclear power stations and for international customers. Fuel manufacture is scheduled to continue into the future. In addition to fuel manufacture, Springfields is also undertaking decommissioning activities. Today it is run and operated by Springfields Fuels Limited, under the management of Westinghouse Electric UK Limited. The site has been operating since 1946 manufacturing nuclear fuel. As part of the decommissioning activities, there was a need was to quantify contamination in a large redundant building. This building had been used to process uranium derived from uranium ore concentrate but had also processed a limited quantity of recycled uranium. The major non-uranic contaminant was Tc-99. The aim was to be able to identify any areas where the bulk activity exceeded 0.4 Bq/g Tc-99 as this would preclude the demolition rubble being sent to the local disposal facility. The problems associated with this project were the presence of significant uranium contamination, the realisation that both the Tc-99 and the uranium had diffused into the brickwork to a significant depth and the relatively low beta energy of Tc-99. The uranium was accompanied by Pa-234m, an energetic beta emitter. The concentration/depth profile was determined for several areas on the plant for Tc-99 and for uranium. The radiochemical analysis was performed locally but the performance of the local laboratory was checked during the initial investigation by splitting samples three ways and having confirmation analyses performed by 2 other laboratories. The results showed surprisingly consistent concentration gradients for Tc-99 and for uranium across the samples. Using that information, the instrument response was calculated for Tc-99 using the observed diffusion gradient and averaged through the full 225 mm of brick wall, as agreed by the regulator. The Tc-99 and uranium contributions to the detector signal were separated

  13. Uptake of uranium by plants growing on and around uranium mill tailings pond at Jaduguda, India

    International Nuclear Information System (INIS)

    A field study was conducted in an area where uranium mill tailings are discharged in the form of slurry (mixture of fine sand and effluent). The fine tailings sand is retained there and effluent is decanted for further treatment. Over the years, certain plant species like Typha latifolia, Saccharum spontanium, Ipomoea carnia etc. have covered the major portion of the tailings pond. Concentration and concentration ratio of uranium in different organs of these plants were evaluated. Concentration of uranium in Typha latifolia plant from tailings pond and the CR was found to have inverse relationship with substrate uranium content. Correlation coefficient between CR(R) and soil, CR(St) and soil and CR(L) and soil in Typha latifolia was -0.80, -0.90 and -0.86 respectively. (author)

  14. TAMARA - an uranium extraction pilot plant for demonstration of computerized process-control in reprocessing. Pt. 1

    International Nuclear Information System (INIS)

    An uranium extraction pilot plant with in-line instrumentation is described. The plant was constructed in the course of the development and demonstration of a computer-based control of nuclear fuel reprocessing processes and is connected with the process computer system CALAS. The results gained until now are presented and discussed, and the future work suggested is mentioned. (orig.)

  15. Reactivity feedbacks of a material test research reactor fueled with various low enriched uranium dispersion fuels

    International Nuclear Information System (INIS)

    The reactivity feedbacks of a material test research reactor using various low enriched uranium fuels, having same uranium density were calculated. For this purpose, the original aluminide fuel (UAlx-Al) containing 4.40 gU/cm3 of an MTR was replaced with silicide (U3Si-Al and U3Si2-Al) and oxide (U3O8-Al) dispersion fuels having the same uranium density as of the original fuel. Calculations were carried out to find the fuel temperature reactivity feedback, moderator temperature reactivity feedback, moderator density reactivity feedback and moderator void reactivity feedback. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It was observed that the magnitudes all the respective reactivity feedbacks from 38 deg. C to 50 deg. C and 100 deg. C, at the beginning of life, of all the fuels were very close to each other. The fuel temperature reactivity feedback of the U3O8-Al was about 2% more than the original UAlx-Al fuel. The magnitudes of the moderator temperature, moderator density and moderator void reactivity feedbacks of all the fuels, showed very minor variations from the original aluminide fuel.

  16. Surface harmonics method for burnup calculations of VVER-1000 fuel assemblies with uranium and MOX fuel

    International Nuclear Information System (INIS)

    Development of the SUHAM-U code for burnup calculations of VVER-1000 fuel assemblies with uranium and MOX fuel is described. Developed SUHAM-U code has capacity to calculate burnup in each fuel or poison zone of each cell of VVER-1000 fuel assembly. In so doing Surface Harmonics Method is used for calculation of the detail neutron spectra in fuel assembly at separated burnup values. Verification of SUHAM-U code by burnup calculations of VVER-1000 fuel assemblies with uranium and MOX fuel has been carried out. Comparisons were carried out with calculations by UNK and RECOL codes. UNK code uses the first collisions probabilities method for solution of the neutron transport equation and RECOL code uses Monte-Carlo method with point-wise continues energy presentation of cross-sections. The main conclusion of all comparisons is the SUHAM-U code calculates the fuel burnup of VVER-1000 fuel assemblies with uranium and MOX fuel with enough high accuracy. Time expenditures are adduced. (authors)

  17. Surface harmonics method for burnup calculations of VVER-1000 fuel assemblies with uranium and MOX fuel

    Energy Technology Data Exchange (ETDEWEB)

    Boyarinov, V. F.; Davidenko, V. D.; Polismakov, A. A.; Tsibulsky, V. F. [Russian Research Center Kurchatov Inst., Nuclear Reactor Inst., 123182, Moscow (Russian Federation)

    2006-07-01

    Development of the SUHAM-U code for burnup calculations of VVER-1000 fuel assemblies with uranium and MOX fuel is described. Developed SUHAM-U code has capacity to calculate burnup in each fuel or poison zone of each cell of VVER-1000 fuel assembly. In so doing Surface Harmonics Method is used for calculation of the detail neutron spectra in fuel assembly at separated burnup values. Verification of SUHAM-U code by burnup calculations of VVER-1000 fuel assemblies with uranium and MOX fuel has been carried out. Comparisons were carried out with calculations by UNK and RECOL codes. UNK code uses the first collisions probabilities method for solution of the neutron transport equation and RECOL code uses Monte-Carlo method with point-wise continues energy presentation of cross-sections. The main conclusion of all comparisons is the SUHAM-U code calculates the fuel burnup of VVER-1000 fuel assemblies with uranium and MOX fuel with enough high accuracy. Time expenditures are adduced. (authors)

  18. The utmost ends of the nuclear fuel cycle. How Finns perceive the risks of uranium mining and nuclear waste management

    International Nuclear Information System (INIS)

    The aim of the paper is to analyze how Finns perceive the risks of uranium mining and nuclear waste management. In social science quite much research has been done on the issue of how people perceive the risks of nuclear waste and nuclear waste management, but not much has been done in analyzing the similarities and differences of risk perception (and ethical considerations) of the utmost ends of nuclear fuel cycle. There have been some changes in Finnish nuclear policy during ongoing decade, which make this type of study interesting: decision on the fifth nuclear power plant was done in 2002, the site for spent nuclear fuel has been chosen in 2001 and in 2010 the Parliament will decide which of three competitors will get the permission to construct the sixth nuclear power plant. This national nuclear renaissance was accompanied with the uranium boom, which started in 2005. New international interest in nuclear power had raised the price of uranium. International mining companies started uranium explorations because Finnish bedrock is the oldest in Europe, and it is similar with and also of the same age as is that of the great uranium producers, Canada and Australia. The analysis of risk perceptions between uranium questions and spent nuclear fuel is based on the national survey data (N=1180) gathered in 2007

  19. Advanced fuel cycles: a rationale and strategy for adopting the low-enriched-uranium fuel cycle

    International Nuclear Information System (INIS)

    A two-year study of alternatives to the natural uranium fuel cycle in CANDU reactors is summarized. The possible advanced cycles are briefly described. Selection criteria for choosing a cycle for development include resource utilization, economics, ease of implementaton, and social acceptability. It is recommended that a detailed study should be made with a view to the early implementation of the low-enriched uranium cycle. (LL)

  20. Treatment of uranium-thorium fuel at its production stage

    International Nuclear Information System (INIS)

    The possibility of removing 232U at the stage of obtaining 233U for the convenience of processing and reducing radiation dose has been analyzed in the paper. This problem is solved by the technology of obtaining 233U in extracted neutron beams in cold channels of a reactor. This technology will allow the acceleration of the implementation of the uranium-thorium fuel cycle in the current reactor technologies

  1. Neutronic performance of uranium nitride composite fuels in a PWR

    International Nuclear Information System (INIS)

    Highlights: • Survey and sensitivity assembly level studies for uranium nitride composite fuels. • Composites harden the neutron spectrum and decrease the worth of control rods. • Moderator temperature coefficient is more negative, soluble boron coefficient is less negative. • Similar equilibrium core power peaking and reactivity coefficient when compared to UO2. • Illustrates “do no harm” in evaluation of candidate accident tolerant fuels. - Abstract: Uranium mononitride (UN) based composite nuclear fuels may have potential benefits in light water reactor applications, including enhanced thermal conductivity and increased fuel density. However, uranium nitride reacts chemically when in contact with water, especially at high temperatures. To overcome this challenge, several advanced composite fuels have been proposed with uranium nitride as a primary phase. The primary nitride phase is “shielded” from water by a secondary phase, which would allow the potential benefits of nitride fuels to be realized. This work is an operational assessment of four different candidate composite materials. We considered uranium dioxide (UO2) and UN base cases and compared them with the candidate composite UN-based fuels. The comparison was performed for nominal conditions in a reference PWR with Zr-based cladding. We assessed the impact of UN porosity on the operational performance, because this is a key sensitivity parameter. As composite fuels, we studied UN/U3Si5, UN/U3Si2, UN/UB4, and UN/ZrO2. In the case of UB4, the boron content is 100% enriched in 11B. The proposed zirconium dioxide (ZrO2) phase is cubic and yttria-stabilized. In all cases UN is the primary phase, with small fractions of U3Si5, U3Si5, UB4, or ZrO2 as a secondary phase. In this analysis we showed that two baseline nitride cases at different fractions of theoretical density (0.8 and 0.95) generally bound the neutronic performance of the candidate composite fuels. Performance was comparable with

  2. Characterization of past and present solid waste streams from the Plutonium-Uranium Extraction Plant

    International Nuclear Information System (INIS)

    During the next two decades the transuranic wastes, now stored in the burial trenches and storage facilities at the Hanford Site, are to be retrieved, processed at the Waste Receiving and Processing Facility, and shipped to the Waste Isolation Pilot Plant near Carlsbad, New Mexico for final disposal. Over 7% of the transuranic waste to be retrieved for shipment to the Waste Isolation Pilot Plant has been generated at the Plutonium-Uranium Extraction (PUREX) Plant. The purpose of this report is to characterize the radioactive solid wastes generated by PUREX using process knowledge, existing records, and oral history interviews. The PUREX Plant is currently operated by the Westinghouse Hanford Company for the US Department of Energy and is now in standby status while being prepared for permanent shutdown. The PUREX Plant is a collection of facilities that has been used primarily to separate plutonium for nuclear weapons from spent fuel that had been irradiated in the Hanford Site's defense reactors. Originally designed to reprocess aluminum-clad uranium fuel, the plant was modified to reprocess zirconium alloy clad fuel elements from the Hanford Site's N Reactor. PUREX has provided plutonium for research reactor development, safety programs, and defense. In addition, the PUREX was used to recover slightly enriched uranium for recycling into fuel for use in reactors that generate electricity and plutonium. Section 2.0 provides further details of the PUREX's physical plant and its operations. The PUREX Plant functions that generate solid waste are as follows: processing operations, laboratory analyses and supporting activities. The types and estimated quantities of waste resulting from these activities are discussed in detail

  3. Characterization of past and present solid waste streams from the Plutonium-Uranium Extraction Plant

    Energy Technology Data Exchange (ETDEWEB)

    Pottmeyer, J.A.; Weyns, M.I.; Lorenzo, D.S.; Vejvoda, E.J. [Los Alamos Technical Associates, Inc., NM (US); Duncan, D.R. [Westinghouse Hanford Co., Richland, WA (US)

    1993-04-01

    During the next two decades the transuranic wastes, now stored in the burial trenches and storage facilities at the Hanford Site, are to be retrieved, processed at the Waste Receiving and Processing Facility, and shipped to the Waste Isolation Pilot Plant near Carlsbad, New Mexico for final disposal. Over 7% of the transuranic waste to be retrieved for shipment to the Waste Isolation Pilot Plant has been generated at the Plutonium-Uranium Extraction (PUREX) Plant. The purpose of this report is to characterize the radioactive solid wastes generated by PUREX using process knowledge, existing records, and oral history interviews. The PUREX Plant is currently operated by the Westinghouse Hanford Company for the US Department of Energy and is now in standby status while being prepared for permanent shutdown. The PUREX Plant is a collection of facilities that has been used primarily to separate plutonium for nuclear weapons from spent fuel that had been irradiated in the Hanford Site`s defense reactors. Originally designed to reprocess aluminum-clad uranium fuel, the plant was modified to reprocess zirconium alloy clad fuel elements from the Hanford Site`s N Reactor. PUREX has provided plutonium for research reactor development, safety programs, and defense. In addition, the PUREX was used to recover slightly enriched uranium for recycling into fuel for use in reactors that generate electricity and plutonium. Section 2.0 provides further details of the PUREX`s physical plant and its operations. The PUREX Plant functions that generate solid waste are as follows: processing operations, laboratory analyses and supporting activities. The types and estimated quantities of waste resulting from these activities are discussed in detail.

  4. Postirradiation analysis of experimental uranium-silicide dispersion fuel plates

    International Nuclear Information System (INIS)

    Low-enriched uranium silicide dispersion fuel plates were irradiated to maximum burnups of 96% of 235U. Fuel plates containing 33 v/o U3Si and U3Si2 behaved very well up to this burnup. Plates containing 33 v/o U3 Si-Al pillowed between 90 and 96% burnup of the fissile atoms. More highly loaded U3Si Al plates, up to 50 v/o, were found to pillow at lower burnups. Plates containing 40 v/o U3Si showed an increased swelling rate around 85% burnup. (author)

  5. Postirradiation analysis of experimental uranium-silicide dispersion fuel plates

    International Nuclear Information System (INIS)

    Low-enriched uranium silicide dispersion fuel plates were irradiated to maximum burnups of 96% of 235U. Fuel plates containing 33 v/o U3Si and U3Si2 behaved very well up to this burnup. Plates containing 33 v/o U3Si-Al pillowed between 90 and 96% burnup of the fissile atoms. More highly loaded U3Si-Al plates, up to 50 v/o were found to pillow at lower burnups. Plates containing 40 v/o U3Si showed an increase swelling rate around 85% burnup. 5 refs., 10 figs

  6. Uranium density reduction on fuel element side plates assessment

    Energy Technology Data Exchange (ETDEWEB)

    Rios, Ilka A. [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil); Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Andrade, Delvonei A.; Domingos, Douglas B.; Umbehaun, Pedro E. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    During operation of IEA-R1 research reactor, located at Instituto de Pesquisas Energeticas e Nucleares, IPEN - CNEN/SP, an abnormal oxidation on some fuel elements was noted. It was also verified, among the possible causes of the problem, that the most likely one was insufficient cooling of the elements in the core. One of the propositions to solve or minimize the problem is to reduce uranium density on fuel elements side plates. In this paper, the influence of this change on neutronic and thermal hydraulic parameters for IEA-R1 reactor is verified by simulations with the codes HAMMER and CITATION. Results are presented and discussed. (author)

  7. Dysprosia doped uranium dioxide fuel and the Lambda transition

    International Nuclear Information System (INIS)

    Thermodynamic properties of solid and liquid UO2, at temperatures above 2500 K, are rare and difficult to interpret. Low Void Reactivity Fuel (LVRF) employs Dysprosia in the originally natural Uranium fuel, but sufficient knowledge of solubility at all temperatures is still deficient. This paper describes research into the physical chemistry of UO2-Dy2O3 mixtures, and includes an assessment of Dy2O3 effects on UO2 at high temperatures. Specifically, research was conducted on the Lambda transition occurring at approximately 2670 K. The crystal structure determined by neutron diffraction and its effect on thermodynamic properties is discussed. (author)

  8. Thermo-chemical modelling of uranium-free nitride fuels

    International Nuclear Information System (INIS)

    A production process for americium-bearing, uranium-free nitride fuels was modelled using the newly developed ALCHYMY thermochemical database. The results suggested that the practical difficulties with yield and purity are of a kinetic rather than a thermodynamical nature. We predict that the immediate product of the typical decarburisation step is not methane, but hydrogen cyanide. HCN may then undergo further reactions upon cooling, explaining the difficulty in observing any carbophoric molecules in the gaseous off stream. The thermal stability of nitride fuels in different environments was also estimated. We show that sintering of nitride compounds containing americium should be performed under nitrogen atmosphere in order to the avoid the excessive losses of americium reported from sintering under inert gas. Addition of nitrogen in small amounts to fuel pin filling gas also appears to significantly improve the in-pile stability of transuranium nitride fuels. (author)

  9. RA-3 core with uranium silicide fuel elements

    International Nuclear Information System (INIS)

    Following on with studies on uranium silicide fuel elements, this paper reports some comparisons between the use of standard ECN [U3O8] fuel elements and type P-06 [from U3Si2] fuel elements in the RA-3 core.The first results showed that the calculated overall mean burn up is in agreement with that reported for the facility, which gives more confidence to the successive ones. Comparing the mentioned cores, the silicide one presents several advantages such as: -) a mean burn up increase of 18 %; -) an extraction burn up increase of 20 %; -) 37.4 % increase in full power days, for mean burn up. All this is meritorious for this fuel. Moreover, grouped and homogenized libraries were prepared for CITVAP code that will be used for planning experiments and other bidimensional studies. Preliminary calculations were also performed. (author)

  10. Production of 450 kg uranium metal ingot in Augmented Uranium Metal Plant (AUMP)

    International Nuclear Information System (INIS)

    Magnesio-thermic Reduction (MTR) trials were started in early sixties in Uranium Metal Plant (UMP) and the first uranium ingot weighing about 8 kg came out in April 1963. The trials were continued and about 20 ingots of 45 kg were produced initially. These trials were conducted in a small shed which was situated at the same place where the present MTR Section is functioning in UMP. As calcium metal was available during that time from abroad without much difficulty, the uranium ingot production was continued using calcium. Later, switching over to magnesio-thermic reduction became essential due to non availability of calcium and for cost reduction

  11. Synchronous derivative fluorimetric determination of boron in Uranium fuel samples

    International Nuclear Information System (INIS)

    We report a sensitive and selective method for determination of boron in uranium samples by spectrofluorimetry in synchronous derivative mode. This method is based on the complexation of non-fluorescent boron with fluorescent chromotropic acid to form fluorescent boron–chromotrope complex. The spectrum of native fluorescence of chromotropic acid seriously overlaps with that of the complex and hence, synchronous derivative mode was employed in which physical separation of excess ligand and complex is not necessary. With the optimized experimental and instrumental conditions, limit of detection obtained is 2 ng mL−1. The linear concentration range is 5–100 ng mL−1 with regression coefficient better than 0.997. The precision is better than 5% at 10 ng mL−1 level and 3% at 50 ng mL−1 level (n=9). Fluorescence quenching by residual matrix elements in the final sample solution is corrected by slope-ratio method. The method is validated with reference materials and successfully applied to the uranium nuclear fuels with the accuracy of ±10%. The proposed method reduces sample size requirement; thereby reducing load of uranium recovery from analytical waste in case of enriched uranium based samples. - Highlights: • Selectivity and sensitivity increases in synchronous derivative mode. • Sample size reduction that reduces load of enriched uranium recovery. • Eliminates need of physical separation of excess ligand and complex. • Quenching by residual matrix elements corrected by the slope-ratio method. • Important contribution to quality control of fuel materials in nuclear technology

  12. Researches of experimental uranium-gadolinium oxide fuel

    International Nuclear Information System (INIS)

    Full text: The increase of technical and economic indexes of NPP is related to the increase of burning down of nuclear fuel. The prolongation of campaign of exploitation of fuel load of nuclear reactor from 12 to 18 months (and even to 24 months) requires the application of fuel with enrichment on uranium-235 no less than 4.5 %. For suppression of high initial reactivity in the core of water-moderated reactor it is suggested to insert the burning down absorber of neutrons. Such method has the number of advantages over application of burning down absorber of neutrons as an independent construction element. Introduction of burning down absorber as the oxide of gadolinium (Gd2O3) directly in fuels pellets removes the limitations related with placing of control device, does not diminish the amount of fuel elements in core. Linear power flux of fuel elements is not increased, an useful place is saved in the fuel element assembling, parasite absorption of neutrons goes down, a necessity in transporting and storage of irradiated combustible absorbers is no longer relevant. The results of researches of some descriptions of experimental (U,Gd)O2-fuel are represented in this article

  13. Radionuclide compositions of spent fuel and high level waste for the uranium and plutonium fuelled PWR

    International Nuclear Information System (INIS)

    The activities of a selection of radionuclides are presented for three types of reactor fuel of interest in radioactive waste management. The fuel types are for a uranium 'burning' PWR, a plutonium 'burning' PWR using plutonium recycled from spent uranium fuel and a plutonium 'burning' PWR using plutonium which has undergone multiple recycle. (author)

  14. The feasibility study of small long-life gas cooled fast reactor with mixed natural Uranium/Thorium as fuel cycle input

    Science.gov (United States)

    Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Khairurrijal, Monado, Fiber; Sekimoto, Hiroshi

    2012-06-01

    A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE burn-up scheme has been performed. In this study, design GCFR with Helium coolant which can be continuously operated by supplying mixed Natural Uranium/Thorium without fuel enrichment plant or fuel reprocessing plant. The active reactor cores are divided into two region, Thorium fuel region and Uranium fuel region. Each fuel core regions are subdivided into ten parts (region-1 until region-10) with the same volume in the axial direction. The fresh Natural Uranium and Thorium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh natural Uranium/Thorium fuel. This concept is basically applied to all regions in both cores area, i.e. shifted the core of ith region into i+1 region after the end of 10 years burn-up cycle. For the next cycles, we will add only Natural Uranium and Thorium on each region-1. The calculation results show the reactivity reached by mixed Natural Uranium/Thorium with volume ratio is 4.7:1. This reactor can results power thermal 550 MWth. After reactor start-up the operation, furthermore reactor only needs Natural Uranium/Thorium supply for continue operation along 100 years.

  15. Fuel management for the Beznau nuclear power plant in Switzerland

    International Nuclear Information System (INIS)

    The Beznau nuclear power plant consists of two 350 MW(e) PWRs of Westinghouse design. A number of special features characterize the nuclear industry in Switzerland: there is no fuel cycle industry; nuclear materials must be moved through several countries before they arrive in our country, it is therefore important that agreements are in place between those countries and Switzerland; nearly all of the materials and services required have to be paid in foreign currencies; the interest rate in Switzerland is traditionally low. Aspects of fuel management at the Beznau plant discussed against this background are: the procurement of natural uranium, its conversion and enrichment; fuel fabrication, in-core management, reprocessing and plutonium recycling; and fuel cycle costs. (author)

  16. Historic American Engineering Record, Idaho National Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex

    Energy Technology Data Exchange (ETDEWEB)

    Susan Stacy; Julie Braun

    2006-12-01

    Just as automobiles need fuel to operate, so do nuclear reactors. When fossil fuels such as gasoline are burned to power an automobile, they are consumed immediately and nearly completely in the process. When the fuel is gone, energy production stops. Nuclear reactors are incapable of achieving this near complete burn-up because as the fuel (uranium) that powers them is burned through the process of nuclear fission, a variety of other elements are also created and become intimately associated with the uranium. Because they absorb neutrons, which energize the fission process, these accumulating fission products eventually poison the fuel by stopping the production of energy from it. The fission products may also damage the structural integrity of the fuel elements. Even though the uranium fuel is still present, sometimes in significant quantities, it is unburnable and will not power a reactor unless it is separated from the neutron-absorbing fission products by a method called fuel reprocessing. Construction of the Fuel Reprocessing Complex at the Chem Plant started in 1950 with the Bechtel Corporation serving as construction contractor and American Cyanamid Company as operating contractor. Although the Foster Wheeler Corporation assumed responsibility for the detailed working design of the overall plant, scientists at Oak Ridge designed all of the equipment that would be employed in the uranium separations process. After three years of construction activity and extensive testing, the plant was ready to handle its first load of irradiated fuel.

  17. The analysis of uranium prospecting, mining and extraction plant samples

    International Nuclear Information System (INIS)

    The physical methods for the determination of uranium, such as x-ray fluorescence spectrometry, radiometric analysis, fluorimetric analysis, emission spectrographic analysis, and neutron activation analysis, are discussed as well as the chemical methods for the determination of uranium such as decomposition, spectrophotometric methods and the volumetric determination of U3O8 in ammonium diuranate slurries. The general methods of analysis for plant control is discussed, especially regarding the determination of cobalt, manganese, iron, free acids, chlorides, nitrates, silica, amines, isodecanol, thiocyanates and tetrathionates in uranium solvent extraction and leaching processes

  18. Burnup behavior of FBR fuels sourced in uranium and plutonium recycled in PWRs and its influence on fuel cycle economy

    International Nuclear Information System (INIS)

    Considering the strategic security of uranium resources, authors investigated the effective use of fuel recycling in the current light water reactor system (1.1GWe-class PWRs of standard type), where we supposed the uranium fuel reformed by re-enrichment of the recovery uranium from the PWR spent fuels and the MOX fuel produced with the recovery plutonium as main fissile, and examined three types of MOX fuels prepared by making choice of these matrixes among (A) the recovery uranium, (B) the depleted uranium as the waste from natural uranium enrichment, and (C) the depleted one sourced from recovery uranium re-enrichment. Calculations were carried out of the multi-component uranium isotope separation employing the ideal centrifuge-cascade and of the reactor core burnup analysis using the comprehensive neutronics computation system SRAC-2006. Burnup analysis was performed under the following conditions: The concentration of 235U in the depleted natural uranium is 0.2 mol%. This conditional mass-effect on isotope separation affects the other isotope concentrations in the centrifuge cascade. The integrated burnup of uranium and full-MOX fuels is 45 GWd/t- HM during 3 cycles in one batch burning. The spent original-uranium-fuels are cooled for 10 years before these reprocessing and then fabrication of reformed uranium or MOX fuels. In the reprocessing, the plutonium is recovered as a 50-50 mixture with the spent uranium oxide and the remaining uranium oxide is recovered in isolated form. The result of burnup analysis shows that the uranium recovered from the spent original fuel contains 235U enriched about 20 % more than that of the natural uranium and also 236U transformed from 235U capturing neutrons during fuel burning. The constituent 236U behaves as a neutron absorber. Hence, the reformed uranium fuel containing 236U enriched additionally in the re-enrichment process requires the fissile 235U concentrated 1.154 times more than that in the original fuel, in order

  19. Low content uranium alloys for nuclear fuels

    International Nuclear Information System (INIS)

    A description is given of the structure and the properties of low content alloys containing from 0.1 to 0.5 per cent by weight of Al, Fe, Cr, Si, Mo or a combination of these elements. A study of the kinetics and of the mode of transformation has made it possible to choose the most satisfactory thermal treatment. An attempt has been made to prepare alloys suitable for an economical industrial development having a small α grain structure without marked preferential orientation, with very fine and stable precipitates as well as a high creep-resistance. The physical properties and the mechanical strength of these alloys are given for temperatures of 20 to 600 deg C. These alloys proved very satisfactory when irradiated in the form of normal size fuel elements. (authors)

  20. Main radiation characteristics of spent nuclear uranium and uranium-plutonium fuel at long-term storage

    International Nuclear Information System (INIS)

    Comparison of radiotoxicity and decay heat power of spent uranium and uranium - plutonium nuclear fuel at long-term storage is performed. The contributions of the most important nuclides are determined. Their prime extraction and transmutation permits to reduce decay heat power and radiotoxicity of wastes staying in storage.(author)

  1. Extraction of uranium from seawater: evaluation of uranium resources and plant siting

    International Nuclear Information System (INIS)

    This report deals with the evaluation of U.S. coastal waters as a uranium resource and with the selection of a suitable site for construction of a large-scale plant for uranium extraction. Evaluation of the resource revealed that although the concentration of uranium is quite low, about 3.3 ppB in seawater of average oceanic salinity, the amount present in the total volume of the oceans is very great, some 4.5 billion metric tons. Of this, perhaps only that uranium contained in the upper 100 meters or so of the surface well-mixed layer should be considered accessible for recovery, some 160 million tonnes. The study indicated that open ocean seawater acquired for the purpose of uranium extraction would be a more favorable resource than rivers entering the sea, cooling water of power plants, or the feed or effluent streams of existing plants producing other products such as magnesium, bromine, or potable and/or agricultural water from seawater. Various considerations led to the selection of a site for a pumped seawater coastal plant at a coastal location. Puerto Yabucoa, Puerto Rico was selected. Recommendations are given for further studies. 21 figures, 8 tables

  2. Possibilities for recycling of weapon-grade uranium and plutonium and its peaceful use as reactor fuel

    International Nuclear Information System (INIS)

    At present 90% of the energy production is based on fossil fuels. Since March 1999, however, the peaceful use of weapon-grade uranium as reactor fuel is being discussed politically. Partners of this discussion is a group of some private western companies on one side and a state-owned company of the Russian Federation (GUS) on the other. Main topic of the deal besides the winning of electrical energy is the useful disposal of the surplus on weapon-grade material of both leading nations. According to the deal, about 160,000 t of Russian uranium, expressed as natural uranium U3O8, would be processed during the next 15 years. Proven processes would be applied. Those methods are being already used in Russian facilities at low capacity rates. There are shortages in the production of low enriched uranium (LEU), because of the low capacity rates in the old facilities. The capacity should be increased by a factor of ten, but there is not enough money available in Russia for financing the remodeling of the plants. Financing should therefore probably be provided by the western clients of this deal. The limited amount of uranium produced could be furnised to the uranium market without major difficulties for the present suppliers of natural uranium. The discussions regarding the security of the details of the deal - however - are not yet finalized. (orig.)

  3. The relationship of JNC and JCO in the uranium processing plant criticality accident

    International Nuclear Information System (INIS)

    On September 30th 1999, the criticality accident occurred at JCO's uranium conversion building in Tokai. The accident occurred during reconversion from U3O8 to uranium nitrate solution (UNH) with uranium enriched 18.8% and about 60 kgU. JCO contacted with JNC to supply UNH that is fuel material for the experimental fast breeder reactor 'JOYO'. JNC has contracted with JCO that had started nuclear fuel material processing business following a definite policy of Japanese government and developed SUMITOMO ADU PROCESS'. JNC made the first contract with JCO in 1985 and has made a contact every year. There had never been a problem in their products. JNC inspected products based on contract. JNC discharge our duty as customer inspecting products based on contract. As for safety control, JCO had taken licensing safety review and had been permitted to be 'a processing facility'. Therefore JNC understood that JCO produced following this license. 'The Uranium Processing Plant Criticality Accident Investigation' showed that JCO had been taking a different method from the permit and violating the license. However JNC had never been explained about that and JCO's operation procedures had never described about that. Therefore the Criticality Accident couldn't be avoided. This report describes the relationship of JNC and JCO in the uranium reconversion contract for JOYO, atomic development policy of Japanese government, process to the order and the contents of contract. (author)

  4. Aspects of the contamination with oxygen in obtaining low enriched uranium fuel

    International Nuclear Information System (INIS)

    The manufacturing of TRIGA fuel rods with low enriched uranium follows in principle the same route as high-enriched uranium. The high purity of the primary metals (uranium, zirconium and erbium) is important for determining the equilibrium metal-hydrogen phases. The impurities from the metal, on the surface and from hydrogen may have an important influence on the hydriding process. This paper presents the aspects of the fuel contamination with oxygen during the manufacturing process of the low enriched uranium fuel. The continuous control of the oxygen concentration in the working zone ensures avoidance of the accidental contamination. Key words: manufacturing, fuel, oxygen, contamination. (authors)

  5. Laboratory studies of shear/leach processing of zircaloy clad metallic uranium reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Swanson, J.L.; Bray, L.A.; Kjarmo, H.E.; Ryan, J.L.; Matsuzaki, C.L.; Pitman, S.G.; Haberman, J.H.

    1985-12-01

    The safety aspects addressed centered on understanding and explaining the undesirable reactions, ''fires,'' observed in a few instances during earlier processing of such fuel at the Nuclear Fuels Services (NFS) plant at West Valley, New York. Consideration of the dissolver fires that occurred at NFS leads to the conclusion that they resulted from rapid reactions with uranium metal, rather than with zirconium metal or with sensitized weld beads. The fires observed at NFS during hulls handling operations may have involved sensitized weld beads as suggested by earlier investigators, but current results suggest that these fires also could have been caused by reactions involving uranium metal. Very little pyrophoric activity was observed in leeached cladding hulls, indicating a very low probability for safety problems resulting from the U-Zr intermetallic zone in N-Reactor fuel. Consideration of the potential role of hydrides in the fires observed at NFS indicates that they were also not important factors. Consideration was also given to protective atmospheres to be used during shearing to prevent excessive reaction during that operation. A water deluge during shearing will likely provide adequate safety while meshing well with other process considerations. Studies on the dissolution of metallic uranium in nitric acid show an initial slower reaction followed by a faster reaction that proceeds at a sustained rate for a prolonged period of time. At solution concentrations typical of those encountered in practical uranium dissolver conditions, this sustained rate is governed by an equation such as: Dissolution rate = K (surface area) ((HNO3)+2(U))/sup 2.6/. Little difference was found in dissolution rates of as-fabricated and of irradiated fuel. The transuranic element content of leached cladding hulls was found to be approx. 400 nCi/g. This is too high to allow disposal as low-level waste.

  6. Laboratory studies of shear/leach processing of zircaloy clad metallic uranium reactor fuel

    International Nuclear Information System (INIS)

    The safety aspects addressed centered on understanding and explaining the undesirable reactions, ''fires,'' observed in a few instances during earlier processing of such fuel at the Nuclear Fuels Services (NFS) plant at West Valley, New York. Consideration of the dissolver fires that occurred at NFS leads to the conclusion that they resulted from rapid reactions with uranium metal, rather than with zirconium metal or with sensitized weld beads. The fires observed at NFS during hulls handling operations may have involved sensitized weld beads as suggested by earlier investigators, but current results suggest that these fires also could have been caused by reactions involving uranium metal. Very little pyrophoric activity was observed in leeached cladding hulls, indicating a very low probability for safety problems resulting from the U-Zr intermetallic zone in N-Reactor fuel. Consideration of the potential role of hydrides in the fires observed at NFS indicates that they were also not important factors. Consideration was also given to protective atmospheres to be used during shearing to prevent excessive reaction during that operation. A water deluge during shearing will likely provide adequate safety while meshing well with other process considerations. Studies on the dissolution of metallic uranium in nitric acid show an initial slower reaction followed by a faster reaction that proceeds at a sustained rate for a prolonged period of time. At solution concentrations typical of those encountered in practical uranium dissolver conditions, this sustained rate is governed by an equation such as: Dissolution rate = K (surface area) ([HNO3]+2[U])/sup 2.6/. Little difference was found in dissolution rates of as-fabricated and of irradiated fuel. The transuranic element content of leached cladding hulls was found to be approx. 400 nCi/g. This is too high to allow disposal as low-level waste

  7. Some possibilities for improvement of fuel utilization in nuclear power plants

    International Nuclear Information System (INIS)

    Methods for improving the nuclear fuel utilization with the emphasis on LWRs are being dealt with in this paper. Some basic results concerning tubular fuel pellets of the Krsko nuclear power plants are presented, showing promising possibilities for uranium saving from the neutronics point of view. (author)

  8. Research reactor core conversion from the use of highly enriched uranium to the use of low enriched uranium fuels guidebook

    International Nuclear Information System (INIS)

    In view of the proliferation concerns caused by the use of highly enriched uranium (HEU) and in anticipation that the supply of HEU to research and test reactors will be more restricted in the future, this document has been prepared to assist reactor operators in determining whether conversion to the use of low enriched uranium (LEU) fuel designs is technically feasible for their specific reactor, and to assist in making a smooth transition to the use of LEU fuel designs where appropriate

  9. IAEA Activities on Uranium Resources and Production, and Databases for the Nuclear Fuel Cycle

    International Nuclear Information System (INIS)

    In recent years rising expectation for nuclear power has led to a significant increase in the demand for uranium and in turn dramatic increases in uranium exploration, mining and ore processing activities worldwide. Several new countries, often with limited experience, have also embarked on these activities. The ultimate goal of the uranium raw material industry is to provide an adequate supply of uranium that can be delivered to the market place at a competitive price by environmentally sound, mining and milling practices. The IAEA’s programme on uranium raw material encompass all aspects of uranium geology and deposits, exploration, resources, supply and demand, uranium mining and ore processing, environmental issues in the uranium production cycle and databases for the uranium fuel cycle. Radiological safety and environmental protection are major challenges in uranium mines and mills and their remediation. The IAEA has revived its programme for the Uranium Production Site Appraisal Team (UPSAT) to assist Member States to improve operational and safety performances at uranium mines and mill sites. The present paper summarizes the ongoing activities of IAEA on uranium raw material, highlighting the status of global uranium resources, their supply and demand, the IAEA database on world uranium deposit (UDEPO) and nuclear fuel cycle information system (NFCIS), recent IAEA Technical Meetings (TM) and related ongoing Technical Cooperation (TC) projects. (author)

  10. Uptake of uranium by aquatic plants growing in fresh water ecosystem around uranium mill tailings pond at Jaduguda, India.

    Science.gov (United States)

    Jha, V N; Tripathi, R M; Sethy, N K; Sahoo, S K

    2016-01-01

    Concentration of uranium was determined in aquatic plants and substrate (sediment or water) of fresh water ecosystem on and around uranium mill tailings pond at Jaduguda, India. Aquatic plant/substrate concentration ratios (CRs) of uranium were estimated for different sites on and around the uranium mill tailings disposal area. These sites include upstream and downstream side of surface water sources carrying the treated tailings effluent, a small pond inside tailings disposal area and residual water of this area. Three types of plant groups were investigated namely algae (filamentous and non-filamentous), other free floating & water submerged and sediment rooted plants. Wide variability in concentration ratio was observed for different groups of plants studied. The filamentous algae uranium concentration was significantly correlated with that of water (r=0.86, p<0.003). For sediment rooted plants significant correlation was found between uranium concentration in plant and the substrate (r=0.88, p<0.001). Both for other free floating species and sediment rooted plants, uranium concentration was significantly correlated with Mn, Fe, and Ni concentration of plants (p<0.01). Filamentous algae, Jussiaea and Pistia owing to their high bioproductivity, biomass, uranium accumulation and concentration ratio can be useful for prospecting phytoremediation of stream carrying treated or untreated uranium mill tailings effluent. PMID:26360459

  11. 77 FR 2718 - CPV Cimarron Renewable Energy Company, LLC; Supplemental Notice That Initial Market-Based Rate...

    Science.gov (United States)

    2012-01-19

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission CPV Cimarron Renewable Energy Company, LLC; Supplemental Notice That Initial... notice in the above-referenced proceeding of CPV Cimarron Renewable Energy Company, LLC's application...

  12. The creation of a uranium oxide industry, from the laboratory stage to a pilot plant (1961)

    International Nuclear Information System (INIS)

    The qualities of uranium oxide, in particular its good in-pile characteristics and its resistance to corrosion by the usual heat-exchange fluids, have led to this material being chose at the present time as a nuclear fuel in many power reactors, either planned or under construction. A great effort has been made these last few years in France in studying processes for transforming powdered uranium oxide into a dense material with satisfactory behaviour in a neutron flux. The laboratories at Saclay have studied the physico-chemical features of the phenomena accompanying the calcination of uranium peroxide or ammonium uranate to give uranium trioxide, and the subsequent reduction of the latter to dioxide as well as the sintering of the powders obtained. This work has made it possible on one hand to prepare powder of known specific surface area, and on the other to show the overriding influence of this factor, all other things being equal, on the behaviour of powders during sintering in a hydrogen atmosphere. The work has led to defining two methods for sintering stoichiometric uranium oxide of high density. The technological study of the preparation of the powder and its industrial production are carried out at the plant of Le Bouchet which produces at the moment powders of known characteristics suitable for sintering in hydrogen at 1650 deg. C without prior grinding. The industrial sintering is carried out by the Compagnie industrielle des Combustibles Atomiques Frittes who has set up a pilot plant having a capacity of 25 metric tons/year, for the Commissariat l'Energie Atomique and has been operating this plant since May 1958. This plant is presented by a film entitled 'uranium oxide'. (author)

  13. Mixing of Al into uranium silicides reactor fuels

    International Nuclear Information System (INIS)

    SEM observations have shown that irradiation induced interaction of the aluminum cladding with uranium silicide reactor fuels strongly affects both fission gas and fuel swelling behaviors during fuel burn-up. The authors have used ion beam mixing, by 1.5 MeV Kr, to study this phenomena. RBS and the 27Al(p, γ) 28Si resonance nuclear reaction were used to measure radiation induced mixing of Al into U3Si and U3Si2 after irradiation at 300 C. Initially U mixes into the Al layer and Al mixes into the U3Si. At a low dose, the Al layer is converted into UAl4 type compound while near the interface the phase U(Al.93Si.07)3 grows. Under irradiation, Al diffuses out of the UAl4 surface layer, and the lower density ternary, which is stable under irradiation, is the final product. Al mixing into U3Si2 is slower than in U3Si, but after high dose irradiation the Al concentration extends much farther into the bulk. In both systems Al mixing and diffusion is controlled by phase formation and growth. The Al mixing rates into the two alloys are similar to that of Al into pure uranium where similar aluminide phases are formed

  14. Simulation study for the purification of depleted uranium product in FBTR fuel reprocessing

    International Nuclear Information System (INIS)

    A method is developed for the purification of depleted uranium product obtained after partitioning of uranium and plutonium in the third cycle of FBTR fuel reprocessing. Uranium and plutonium were partitioned and recovered by AUC method. It is observed from the study that plutonium recovery is quantitative (100%) if Pu concentration is 0.6 g/L. (author)

  15. Conceptual design study on very small long-life gas cooled fast reactor using metallic natural Uranium-Zr as fuel cycle input

    Energy Technology Data Exchange (ETDEWEB)

    Monado, Fiber, E-mail: fiber.monado@gmail.com [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung, Indonesia and Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Ariani, Menik [Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Su' ud, Zaki; Waris, Abdul; Basar, Khairul; Permana, Sidik [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung (Indonesia); Aziz, Ferhat [National Nuclear Energy Agency of Indonesia (BATAN) (Indonesia); Sekimoto, Hiroshi [CRINES, Tokyo Institute of Technology, O-okoyama, Meguro-ku, Tokyo 152-8550 (Japan)

    2014-02-12

    A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8% HM. From the neutronic point of view, this design is in compliance with good performance.

  16. Conceptual design study on very small long-life gas cooled fast reactor using metallic natural Uranium-Zr as fuel cycle input

    International Nuclear Information System (INIS)

    Full-text: A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8 % HM. From the neutronic point of view, this design is in compliance with good performance. (author)

  17. Processing and quality control in the front end of uranium fuel cycle

    International Nuclear Information System (INIS)

    Cameco is one of the largest uranium producing companies in the world and is recognized leader in many aspects of the international uranium industry including exploration, mining, extraction and refining of uranium, and uranium fuel manufacturing. The status of current uranium processing activities at Cameco operations are reviewed emphasizing advantages Cameco enjoys and challenges faced. The management of innovation and technology development in key aspects of Cameco's processing and fuel manufacturing activities are highlighted and an approach for ensuring sustainability of the operations through innovation is discussed. (author)

  18. Molten salt extraction of transuranic and reactive fission products from used uranium oxide fuel

    Science.gov (United States)

    Herrmann, Steven Douglas

    2014-05-27

    Used uranium oxide fuel is detoxified by extracting transuranic and reactive fission products into molten salt. By contacting declad and crushed used uranium oxide fuel with a molten halide salt containing a minor fraction of the respective uranium trihalide, transuranic and reactive fission products partition from the fuel to the molten salt phase, while uranium oxide and non-reactive, or noble metal, fission products remain in an insoluble solid phase. The salt is then separated from the fuel via draining and distillation. By this method, the bulk of the decay heat, fission poisoning capacity, and radiotoxicity are removed from the used fuel. The remaining radioactivity from the noble metal fission products in the detoxified fuel is primarily limited to soft beta emitters. The extracted transuranic and reactive fission products are amenable to existing technologies for group uranium/transuranic product recovery and fission product immobilization in engineered waste forms.

  19. Uranium-fuel thermal reactor benchmark testing of CENDL-3

    International Nuclear Information System (INIS)

    CENDL-3, the new version of China Evaluated Nuclear Data Library are being processed, and distributed for thermal reactor benchmark analysis recently. The processing was carried out using the NJOY nuclear data processing system. The calculations and analyses of uranium-fuel thermal assemblies TRX-1,2, BAPL-1,2,3, ZEEP-1,2,3 were done with lattice code WIMSD5A. The results were compared with the experimental results, the results of the '1986'WIMS library and the results based on ENDF/B-VI. (author)

  20. Quantitative Determination of Uranium Homogeneity Distribution in MTR Fuel Type Plates

    International Nuclear Information System (INIS)

    IPEN/CNEN-SP produces the fuel to supply its nuclear research reactor IEA-R1. The fuel is assembled with fuel plates containing a U3Si2-Al composite meat. A good homogeneity in the uranium distribution inside the fuel plate meat is important from the standpoint of irradiation performance. Considering the lower power of reactor IEA- R1, the uranium distribution in the fuel plate has been evaluated only by visual inspection of radiographs. However, with the possibility of IPEN to manufacture the fuel for the new Brazilian Multipurpose Reactor (RMB), with higher power, it urges to develop a methodology to determine quantitatively the uranium distribution into the fuel. This paper presents a methodology based on X-ray attenuation, in order to qualify the uranium concentration distribution in the meat of the fuel plate by using optical densities in radiographs and comparison with standards. (author)

  1. Decommissioning of B and W's fuel conversion plant

    International Nuclear Information System (INIS)

    B and W is managing an ongoing $65 million Project involving the site characterization, decontamination, and deconstruction of its former nuclear fuel fabrication plant in Apollo, Pennsylvania. This 90,000 ft facility was used from the late 1950's until the early 1980's for the conversion of uranium hexafluoride to various fuel forms, including uranium dioxide powder and pellets. Both high- and low-enriched uranium as well as thorium were processed in the facility. Upon discontinuing fuel manufacturing operations, the chemical processing equipment was decontaminated, removed, packaged, and shipped to a licensed low-level radioactive waste (LLRW) burial site. As a result of plant operations, uranium contamination existed within the building and in the soils on the plant site. A detailed site characterization program was completed to establish the extent of contamination and to plan the subsequent soil remediation and building deconstruction efforts. As a result of several factors, B and W made the decision in 1990 to accelerate the final decommissioning of the Apollo site. These factors also became constraints on the completion of the project: Rapidly escalating waste disposal costs, with LLRW burial site surcharges scheduled to increase from $40 to $120 per cubic foot in January 1992; Increasing regulatory confusion on the criteria for the residual radioactivity contamination levels that can remain on an NRC-licensed site being remediated for unlicensed, unrestricted use; The probable loss of burial site alternatives in January 1993 due to the provisions of the Low-Level Radioactive Waste Policy Amendments Act of 1985; Delays in the siting and construction of the Appalachian Compact's burial site which is projected to have a capacity insufficient to handle the large volume of waste produced by a major decommissioning project. This paper presents an overview of this decontamination and decommissioning project with emphasis on the key business issues which

  2. Demand of natural uranium to satisfy the requirements of nuclear fuel of new nuclear power plants in Mexico; Demanda de uranio natural para satisfacer los requerimientos de combustible nuclear de nuevas centrales nucleares en Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J. R.; Rios, M. del C.; Alonso, G.; Palacios H, J. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: jrrs@nuclear.inin.mx

    2008-07-01

    Due to the expectation of that in Mexico new plants of nuclear energy could be installed, turns out from the interest to evaluate the uranium requirements to operate those plants and to also evaluate if the existing reserves in the country could be sufficient to satisfy that demand. Three different scenes from nuclear power plant expansion for the country are postulated here that are desirable for the diversification of generation technologies. The first scene considers a growth in the generation by nuclear means of two reactors of type ABWR that could enter operation by years 2015 and 2020, in the second considers the installation of four reactors but as of 2015 and new every 5 years, in the scene of high growth considers the installation of 6 reactors of the same type that in the other scenes, settling one every three years as of 2015. The results indicate that the uranium reserves could be sufficient to only maintain in operation to one of the reactors proposed by the time of their useful life. (Author)

  3. Uranium uptake by hydroponically cultivated crop plants

    Czech Academy of Sciences Publication Activity Database

    Soudek, Petr; Petrová, Šárka; Benešová, Dagmar; Dvořáková, Marcela; Vaněk, Tomáš

    2011-01-01

    Roč. 102, č. 6 (2011), s. 598-604. ISSN 0265-931X R&D Projects: GA MŠk OC09082; GA MŠk 2B06187; GA MŠk 2B08058 Institutional research plan: CEZ:AV0Z50380511 Keywords : Uranium * Uptake * Sinapis alba Subject RIV: DK - Soil Contamination ; De-contamination incl. Pesticides Impact factor: 1.339, year: 2011

  4. Kinetic and thermodynamic bases to resolve issues regarding conditioning of uranium metal fuels

    International Nuclear Information System (INIS)

    Numerous uranium - bearing fuels are corroding in fuel storage pools in several countries. At facilities where reprocessing is no longer available, dry storage is being evaluated to preclude aqueous corrosion that is ongoing. It is essential that thermodynamic and kinetic factors are accounted for in transitions of corroding uranium-bearing fuels to dry storage. This paper addresses a process that has been proposed to move Hanford N-Reactor fuel from wet storage to dry storage

  5. Determination of uranium in PWR spent fuels by coulometric titration method

    International Nuclear Information System (INIS)

    Controlled-potential coulometric titration method was applied in 0.5M sulphuric acid medium for the determination of uranium content in samples of PWR spent fuel. In this study, we discussed some experimental conditions related to the determination of uranium in PWR spent fuel samples. Accuracy(recovery of uranium) for the coulometric determination of 1∼7mg uranium standard was 99.96∼100.88%. Precision(relative standard deviation, rsd) for the coulometric determination(n=3) of 3∼4mg uranium in PWR spent fuel samples was 0.07∼0.68%. Relative error for the results of the potentiometric and coulometric determination of uranium PWR spent fuel samples was +0.65∼-2.76%

  6. Metallography of pitted aluminum-clad, depleted uranium fuel

    International Nuclear Information System (INIS)

    The storage of aluminum-clad fuel and target materials in the L-Disassembly Basin at the Savannah River Site for more than 5 years has resulted in extensive pitting corrosion of these materials. In many cases the pitting corrosion of the aluminum clad has penetrated in the uranium metal core, resulting in the release of plutonium, uranium, cesium-137, and other fission product activity to the basin water. In an effort to characterize the extent of corrosion of the Mark 31A target slugs, two unirradiated slug assemblies were removed from basin storage and sent to the Savannah River Technology Center for evaluation. This paper presents the results of the metallography and photographic documentation of this evaluation. The metallography confirmed that pitting depths varied, with the deepest pit found to be about 0.12 inches (3.05 nun). Less than 2% of the aluminum cladding was found to be breached resulting in less than 5% of the uranium surface area being affected by corrosion. The overall integrity of the target slug remained intact

  7. Postirradiation examination of high-density uranium alloy dispersion fuels

    International Nuclear Information System (INIS)

    Two irradiation test vehicles, designated RERTR-2, were inserted into the Advanced Test reactor in Idaho in August 1997. These tests were designed to obtain irradiation performance information on a variety of potential new, high-density uranium alloy dispersion fuels, including U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru and U-10Mo-0.05Sn: the intermetallic compounds U2 Mo and U-10Mo-0.-5Sn; the intermetallic compounds U2 Mo and U3 Si2 were also included in the fuel test matrix. These fuels are included in the experiments as microplates (76 mm x 22 mm x 1.3mm outer dimensions) with a nominal fuel volume loading of 25% and irradiated at relatively low temperature (∼100 deg C). RERTR-1 and RERTR-2 were discharged from the reactor in November 1997 and July 1998, respectively at calculated peak fuel burnups of 45 and 71 at %-U235 Both experiments are currently under examination at the Alpha Gamma Hot Cell Facility at Argonne National Laboratory in Chicago. This paper presents the postirradiation examination results available to date from these experiments. (author)

  8. Quantification of damage in uranium dioxide fuel matrices

    International Nuclear Information System (INIS)

    Atomic displacements are accepted as the principal underlying radiation damage mechanism in many materials subjected to fields of energetic neutrons. It is believed that accumulated displacements at the microscopic level (i.e., radiation damage) form the basis for the changes in material properties at the macroscopic level (i.e., radiation effects). Therefore, it is important to quantify the number of displacements caused in such materials by the radiation field corresponding to their intended use. Such a characterization is seldom possible and it is common practice to conduct accelerated irradiation tests in which a comparable, albeit preferably much stronger, irradiation field is used. Theory is then used to correlate the observable experimental effects to the expected behavior in the actual intended operational field. In many materials, atomic displacements may be the principal mechanism of radiation damage. In other classes of materials, such as fuel matrices, volatile (gas) products may be the dominant contributors to macroscopic physical property changes. Some materials, such as graphite, may accumulate significant damage and reflect effects by both mechanisms, and thus require special attention. Previous work has documented a set of modifications to the NJOY99 code to permit more accurate calculation of displacement kerma cross sections for binary materials. The NJOY99 code is widely used for nuclear data processing. It operates on evaluated cross section data, such as available in the ENDF database, and transforms them into pointwise or groupwise cross sections suitable for use in other nuclear analysis codes. The present work extends previous work to permit more complete modeling of damage mechanisms in fissionable materials, including the generation of cross sections for the production of gaseous species are developed. These enhancements allow more accurate estimation of the damage to uranium dioxide fuel matrices. Displacement creation cross sections

  9. Slightly enriched uranium in CANDU: An economic first step towards advanced fuel cycles

    International Nuclear Information System (INIS)

    The natural-uranium fuelled Canada Deuterium-Uranium (CANDU) nuclear reactor system has proven to be a safe, reliable and economical producer of electricity for over a quarter of a century. The CANDU system, however, is not restricted to the use of natural-uranium fuel; a wide range of advanced fuel cycles can be accommodated. In the short term, slightly enriched uranium (SEU) is the most promising of these advanced fuel cycles. SEU offers a reduction in the total fuel cycle cost of between 25 and 50% relative to natural-uranium fuel. Uranium consumption is decreased by 30 to 40%. In addition the volume of spent fuel is reduced by a factor of two to three, depending on the enrichment selected. SEU also offers greater flexibility in the design of future CANDU reactors. A variety of fuel management options can be employed in CANDU with slightly enriched fuels. Fuel performance is expected to be good for the burnups of interest, but further fuel testing is planned and is currently in progress in order to confirm this. Programs in place at Atomic Energy of Canada Limited (AECL) will lead to the demonstration and introduction of slightly enriched uranium in CANDU. Ontario Hydro, a Canadian utility with twenty CANDUs operating or under construction, is considering a program which could lead to the implementation of SEU in its nuclear generating stations. (author). 30 refs, 7 figs

  10. Postirradiation analysis of experimental uranium-silicide dispersion fuel plates

    Energy Technology Data Exchange (ETDEWEB)

    Hofman, G.L.; Neimark, L.A.

    1985-01-01

    Low-enriched uranium silicide dispersion fuel plates were irradiated to maximum burnups of 96% of /sup 235/U. Fuel plates containing 33 v/o U/sub 3/Si and U/sub 3/Si/sub 2/ behaved very well up to this burnup. Plates containing 33 v/o U/sub 3/Si-Al pillowed between 90 and 96% burnup of the fissile atoms. More highly loaded U/sub 3/Si-Al plates, up to 50 v/o were found to pillow at lower burnups. Plates containing 40 v/o U/sub 3/Si showed an increase swelling rate around 85% burnup. 5 refs., 10 figs.

  11. Extraction of Uranium Using Nitrogen Dioxide and Carbon Dioxide for Spent Fuel Reprocessing

    International Nuclear Information System (INIS)

    For the reprocessing of spent nuclear fuels, a new method to extract actinides from spent fuel using highly compressed gases, nitrogen dioxide and carbon dioxide was proposed. Uranium extraction from broken pieces, whose average grain size was 5 mm, of uranium dioxide pellet with nitrogen dioxide and carbon dioxide was demonstrated in the present study. (authors)

  12. Feedback Experience from Decommissioning of Uranium Conversion Plant

    International Nuclear Information System (INIS)

    KAERI has been conducting decommissioning activities of Uranium Conversion Plant (UCP) for the last decade. As a result of all this work KAERI has accumulated significant experience in the field of decommissioning of nuclear facilities. On the basis of the experience gained from decommissioning activities, this paper describes several lessons learned

  13. Selected bibliography for the extraction of uranium from seawater: evaluation of uranium resources and plant siting

    Energy Technology Data Exchange (ETDEWEB)

    Chen, A.C.T.; Gordon, L.I.; Rodman, M.R.; Binney, S.E.

    1979-02-06

    This bibliography contains 471 references pertaining to the evaluation of U.S. territorial ocean waters as a potential uranium resource and to the selection of a site for a plant designed for the large scale extraction of uranium from seawater. This bibliography was prepared using machine literature retrieval, bibliographic, and work processing systems at Oregon State University. The literature cited is listed by author with indices to the author's countries, geographic areas of study, and to a set of keywords to the subject matter.

  14. Selected bibliography for the extraction of uranium from seawater: evaluation of uranium resources and plant siting

    International Nuclear Information System (INIS)

    This bibliography contains 471 references pertaining to the evaluation of U.S. territorial ocean waters as a potential uranium resource and to the selection of a site for a plant designed for the large scale extraction of uranium from seawater. This bibliography was prepared using machine literature retrieval, bibliographic, and work processing systems at Oregon State University. The literature cited is listed by author with indices to the author's countries, geographic areas of study, and to a set of keywords to the subject matter

  15. Uranium production as byproduct from Yarimca (Turkey) phosphoric acid plant

    International Nuclear Information System (INIS)

    Full text: This paper deals with uranium production from the phosphoric acid products of Yarimca Fertilizer Plant. After examination of the phosphate rocks consumed in this plant and the acid products, solvent extraction tests were conducted to determine the effects of acid concentration, solvent concentration in kerosene, contact time and acid solvent ratio on the recoveries of uranium. 98 percent of total uranium in acid was recovered in the organic phase by applying 5 stage extraction. Following the extraction tests, acidic and basic stripping were applied to organic phase and uranium was precipitated as yellow cake from the stripping solutions. In the stripping tests mainly aqueous and organic phase ratio and the stripping time were investigated using HCl and Na2CO3 as stripping agents. Na2CO3 has provided higher uranium recoveries both at the short time and low ratio of the stripping solution. Yellow cakes were produced containing 13-18.4 percent U3o8 from acidic and 30-46.4 percent U3O8 from basic stripping solutions

  16. The relationship between natural uranium and advanced fuel cycles in CANDU reactors

    International Nuclear Information System (INIS)

    CANDU is the most uranium-economic type of thermal power reactor, and is the only type used in Canada. CANDU reactors consume approximately 15% of Canadian uranium production and support a fuel service industry valued at ∼$250 M/a. In addition to their once-through, natural-uranium fuel cycle, CANDU reactors are capable of operating with slightly-enriched uranium (SEU), uranium-plutonium and thorium cycles, more efficiently than other reactors. Only SEU is economically attractive in Canada now, but the other cycles are of interest to countries without indigenous fuel resources. A program is underway to establish the fuel technologies necessary for the use of SEU and the other fuel cycles in CANDU reactors

  17. The relationship between natural uranium and advanced fuel cycles in CANDU reactors

    International Nuclear Information System (INIS)

    CANDU is the most uranium-economic type of thermal power reactor, and is the only type used in Canada. CANDU reactors consume approximately 15% of Canadian uranium production and support a fuel service industry valued at ∼$250 M/a. In addition to their once-through, natural-uranium fuel cycle, CANDU reactors are capable of operating with slightly-enriched uranium (SEU), uranium-plutonium and thorium cycles, more efficiently than other reactors. Only SEU is economically attractive in Canada now, but the other cycles are of interest to countries without indigenous fuel resources. A program is underway to establish the fuel technologies necessary for the use of SEU and the other fuel cycles in CANDU reactors. 22 refs

  18. Final qualification testing results for TRIGA low-enriched uranium fuel

    International Nuclear Information System (INIS)

    Following the adoption of policies by the US government and other fuel supplier nations to limit, with few exceptions, export of fuels to those enriched to 235U, GA Technologies undertook, starting in 1976, a rigorous program to develop uranium zirconium-hydride fuels with high uranium concentrations up to 3.7 g U/cm3, while limiting the enrichment to 235U. By increasing the uranium concentration from 8.5 to 45 wt% and by including erbium as a burnable poison, the long reactivity lifetime of the high-enriched uranium cores has been preserved. The achieved purpose of these low-enriched uranium (LEU) fuels was also to preserve all the unique safety features of the TRIGA fuel system - prompt negative temperature coefficient of reactivity, high fission product retentivity, chemical stability when quenched from high temperatures in water, and dimensional stability over a broad range of operating temperatures

  19. Perspectives of Siberian chemical plant in increasing volumes of uranium concentrates recycling

    OpenAIRE

    Lazarchuk, V. V.; Shikerun, T. G.; Ryabov, A. S.; Shamin, V. I.; Zhiganov, A. N.

    2007-01-01

    The purification technology of uranium concentrate of natural isotopic composition developed at Siberian chemical enterprise is basically universal, allows recycling uranium concentrates with different content of impurities and obtaining uranium nitrate solutions corresponding by quality to the international standards requirements to uranium hexafluoride preparation for isotopes ASTM C 787-03 separation and to ceramic fuel ASTM C 788-02 preparation. Uranium reserves in Russia and abroad were ...

  20. Spent fuel characteristics analysis for thorium-uranium breeding recycle in PWRs

    International Nuclear Information System (INIS)

    Spent fuel characteristics analyses of thorium-based fuel were investigated using ORIGEN-S code compared with uranium-based fuel. Such parameters as radio- activity, radiotoxicity, decay heat, and gamma ray were considered. Relative results in this work could provide some reference information for storage, reprocessing and disposal of thorium-based spent fuel. Four type fuels, thorium-based fuel U3ThOX (mixed reactor grade 233U-thorium oxide), PuThOX (mixed reactor grade plutonium-thorium oxide), uranium-based fuel UOX (uranium oxide) and MOX (mixed reactor grade plutonium-uranium oxide), on the basis of core designs for thorium-uranium breeding recycle in PWRs were investigated. The calculated results show that: 1) Due to extremely low content of transuranic nuclides, the radiotoxicity of U3ThOX is dramatically lower than that of three other types of spent fuel in 1000 years after discharge; 2) In thorium-based spent fuel the intensity of gamma ray near 2.6 MeV mainly generated by 208Tl in 232U decay chain is much stronger than that in uranium-based fuel. The intensity of γ ray near 2.6 MeV reaches a local peak in about 10 years after discharge when the reprocessing should not be performed for thorium-based spent fuel. (authors)

  1. 77 FR 16868 - Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test...

    Science.gov (United States)

    2012-03-22

    ... COMMISSION Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test...-Type Uranium-Aluminum Fuel Elements for Use in Research and Test Reactors,'' is temporarily identified... verifying the quality of plate-type uranium-aluminum fuel elements used in research and test reactors...

  2. 78 FR 33132 - Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test...

    Science.gov (United States)

    2013-06-03

    ... COMMISSION Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test... Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test Reactors.'' This guide... plate-type uranium-aluminum fuel elements used in research and test reactors (RTRs). ADDRESSES:...

  3. Nuclear data needs for uranium-plutonium fuel cycle development

    International Nuclear Information System (INIS)

    Potential needs of nuclear data have been surveyed through the uranium-plutonium fuel cycle. A wide variety of data were used in operation of fuel cycle facilities and its relating activities. Those were decay data, cross section data and fission-product yield data. Although the nuclear data were used mostly as physical constants in analyzing results of measurements and calculations made, the users did not give a first priority to their requirements on the nuclear data. It is necessary to disseminate uniquely evaluated data for securing uniformity of data-treatment basis of the users. A covariance file of the evaluated cross section data was requested by workers involved in reactor dosimetry studies and in actinides incineration studies for waste management. Also, in the various parts of the fuel cycle equally required were some evaluated computer codes including reasonably organized nuclear data set which enable one to predict the amounts of actinides and fission-product nuclides in irradiated nuclear fuels. (Auth.)

  4. Uranium exploration, mining and the nuclear fuel cycle

    International Nuclear Information System (INIS)

    The object of this paper is to summarize the nuclear industry in an understandable and systematic manner. The authors conclude that: (a) Uranium exploration can be carried out in an environmentally safe manner. (b) Uranium mining is being carried on currently in Canada in an environmentally and socially acceptable manner with many benefits accruing to the local population near the mine. (c) Uranium tailings can be properly handled utilizing modern technology both in the short term and the long term. (d) It is generally agreed by the majority of the scientific community that radiation protection standards adequately protect both nuclear workers and the general public. (e) Nuclear and coal-fired electrical generating plants can both supply base load energy supplies in the short and long term. In some jurisdictions it is the nuclear system which can provide the lowest cost energy supply. It is important that this option not be lost, either as a potential source of electrical energy domestically or as an export commodity

  5. Removal of hydrogen fluoride from uranium plant emissions

    International Nuclear Information System (INIS)

    Uranium production technology involves the use of hydrogen fluoride at various stages. It is used in the production of uranium tetrafluoride as well as for the production of fluorine for the conversion of tetrafluoride to hexafluoride in isotopic enrichment plants. The sources of HF pollution in the industry, besides accidental spillages and leakages, are the final off-gases from the UF4 production process or from the hydrogen reduction of hexafluoride (where such process is adopted), venting of tanks and reactors containing HF, safety pressure rupture discs as well as dust collection and ventilation systems

  6. Study on fuel failure behaviour of plutonium-uranium mixed oxide fuel with NSRR, (1)

    International Nuclear Information System (INIS)

    A research programme is being planned to examine fuel behaviour of plutonium-uranium mixed oxide fuel in water under reactivity initiated accident conditions with Nuclear Safety Research Reactor (NSRR). The plutonium content of the mixed oxide fuel is small as in the case of mixed oxide fuel for thermal reactors. In preparation for the programme, the following had been carried out: design and fabrication in trial of a capsule, evaluations of the energy deposition in a test fuel rod and reactivity worth of a capsule, preliminary experiment with UO2 fuel, and fabrication of mixed oxide fuel pellets. Safety evaluation by the Government of the programme and making of a fuel transportation cask are now in progress. Described in this report are plans in detail of the programme, results of the nuclear characteristic evaluation and preliminary experiment, and the development of a capsule and a transportation cask, safety evaluation by a JAERI ad hoc committee and data in this connection. Preliminary experiments showed that an energy deposition of 300 cal/g. fuel could be attained in a doubly-encapsulated 6.33 wt.% PuO2-UO2 fuel rod by a single pulse irradiation in NSRR; the capsule was also adequate. (author)

  7. Fabrication of high-uranium-loaded U3O8-Al developmental fuel plates

    International Nuclear Information System (INIS)

    A common plate-type fuel for research and test reactors is U3O8 dispersed in aluminum and clad with an aluminum alloy. There is an impetus to reduce the 235U enrichment from above 90% to below 20% for these fuels to lessen the risk of diversion of the uranium for nonpeaceful uses. Thus, the uranium content of the fuel plates has to be increased to maintain the performance of the reactors. This paper describes work at ORNL to determine the maximal uranium loading for these fuels that can be fabricated with commercially proven materials and techniques and that can be expected to perform satisfactorily in service

  8. System Studies of Fuel Cell Power Plants

    OpenAIRE

    Kivisaari, Timo

    2001-01-01

    This thesis concerns system studies of power plants wheredifferent types of fuel cells accomplish most of the energyconversion. Ever since William Grove observed the fuel cell effect inthe late 1830s fuel cells have been the subject or more or lessintense research and development. Especially in the USA theseactivities intensified during the second part of the 1950s,resulting in the development of the fuel cells used in theApollo-program. Swedish fuel cell activities started in themid-1960s, w...

  9. Burn-up behavior of FBR fuels sourced in uranium and plutonium recycled in PWRs and its influence on fuel economy

    International Nuclear Information System (INIS)

    Considering the strategic security of uranium resources, the authors investigated the effectual use of fuels recycled in a current LWR system (1.1 GWe-class PWRs of standard type), where they supposed the uranium fuel remade by re-enrichment of the recovery uranium from PWR spent fuels and the MOX fuel produced with the recovery plutonium as main fissile, and examined three cases of MOX fuels prepared by making choice of these matrixes among (A) the recovery uranium, (B) the depleted uranium as waste from natural uranium enrichment and (C) the depleted one sourced from recovery uranium re-enrichment. The result suggests that the multi-recycle of fuels in the LWR system brings the decline in fuel qualities. Particularly, the re-enrichment of recovery uranium brings the issue of an increase of 236U in remanufactured fuel. Thus, in order to investigate the burn-up of fast reactor fuels sourced from the PWR fuel system, they designed a model core of practical-FBR with reference to a concept of FBR reported by JAEA, using the SRAC numerical system. The burn-up behavior of FBR fuels was analyzed which were sourced in the original uranium spent-fuel and the remade uranium spent-fuel. And also, the breeding behavior of blanket materials was investigated which were individually of the depleted natural uranium, the recovery uranium from the original uranium spent-fuel and the recovery uranium from the remade uranium spent-fuel. The fissile 235U in FBR fuels reduces the burden of plutonium while the containment of 236U declines the neutron multiplication in FBRs. (author)

  10. Existing status of uranium refining and conversion plant decommissioning project

    International Nuclear Information System (INIS)

    This technical report shows the situation of the dismantling of the main equipment in the radiation-controlled area of a uranium refining and conversion plant. The dismantling was carried out at the beginning of the uranium refining and conversion plant decommissioning project. We started the dismantling in April 2008 and finished it in 29 September 2011. The dismantled waste and equipment were stored in 200 small drums. All the contaminated devices were sealed and kept in this stage. The radioactivity inventory of the uranium refining and conversion plant did not change in this stage. However, the risk of contamination due to the deterioration of this facility with time became remarkably small. Moreover, we were able to get many information and experience about dismantling. Then, we began decommissioning. We were in a new stage from April 2012. We are going to dismantle or tightly close the fluidization media storage underground tank, the neutralization and precipitation system of a waste fluid with fluorine, and the uranium and ventilation system in about three years from now on. (author)

  11. UFRACAM - a computer code for the evaluation of uranium feed requirements and fresh fuel cost items subject to optimum tails assay

    International Nuclear Information System (INIS)

    The work presented in this paper is a part of an extensive nuclear fuel cycle evaluation program that can serve in fuel cycle optimization and economic evaluation of nuclear power plants. The paper describes the main features of the computer code ''UFRACAM'' developed for the estimation of uranium feed requirements and charges associated with the preirradiation fuel cycle operations for reactors fueled with enriched uranium fuel. The code incorporates a computer routine which calculates the optimum tails assay that yields minimum cost for the enriched uranium program. The program is so flexible that it allows the user to chose the type of calculations and results he needs. A sample problem with input data and output results description is presented. A complete FORTRAN listing of the computer code is also provided

  12. Biometric approach in selecting plants for phytoaccumulation of uranium.

    Science.gov (United States)

    Stojanović, Mirjana; Pezo, Lato; Lačnjevac, Časlav; Mihajlović, Marija; Petrović, Jelena; Milojković, Jelena; Stanojević, Marija

    2016-01-01

    This paper promotes the biometric classification system of plant cultivars, unique characteristics, in terms of the uranium (U) uptake, primarily in the function of the application for phytoremediation. It is known that the degree of adoption of U depends on the plant species and its morphological and physiological properties, but it is less known what impact have plants cultivars, sorts, and hybrids. Therefore, we investigated the U adoption in four cultivars of three plant species (corn, sunflower and soy bean). "Vegetation experiments were carried out in a plastic-house filled with soil (0.66 mgU) and with tailing (15.3 mgU kg(-1)) from closed uranium mine Gabrovnica-Kalna southeast of Serbia". Principal Component Analysis (PCA), Cluster Analysis (CA) and analysis of variance (ANOVA) were used for assessing the effect of different substrates cultivars, plant species and plant organs (root or shoot) on U uptake. Obtained results showed that a difference in U uptake by three investigated plant species depends not only of the type of substrate types and plant organs but also of their cultivars. Biometrics techniques provide a good opportunity for a better understanding the behavior of plants and obtaining much more useful information from the original data. PMID:26606604

  13. The new French uranium refining plant at Narbonne

    International Nuclear Information System (INIS)

    In 1957 the Commissariat l'Energie Atomique in collaboration with two French industrial firms, the Compagnie de Saint-Gobain and the Societe Potasse et Engrais chimique, undertook the construction of a plant for the production of refined uranium on an industrial scale. This plant, which forms part of the French nuclear equipment programme and which works at a capacity of 1000 tons/year, was put into operation in July 1959. First of all the principles on which this under-taking is based are outlined. This is followed by a more detailed account of the construction, including the improvements brought to the process developed at the C.E.A. plant at le Bouchet when it was carried over to the industrial stage by the uranium branch of the Societe d'Etudes et de Travaux. (author)

  14. Impact of arbuscular mycorrhizal fungi on uranium accumulation by plants

    International Nuclear Information System (INIS)

    Contamination by uranium (U) occurs principally at U mining and processing sites. Uranium can have tremendous environmental consequences, as it is highly toxic to a broad range of organisms and can be dispersed in both terrestrial and aquatic environments. Remediation strategies of U-contaminated soils have included physical and chemical procedures, which may be beneficial, but are costly and can lead to further environmental damage. Phytoremediation has been proposed as a promising alternative, which relies on the capacity of plants and their associated microorganisms to stabilize or extract contaminants from soils. In this paper, we review the role of a group of plant symbiotic fungi, i.e. arbuscular mycorrhizal fungi, which constitute an essential link between the soil and the roots. These fungi participate in U immobilization in soils and within plant roots and they can reduce root-to-shoot translocation of U. However, there is a need to evaluate these observations in terms of their importance for phytostabilization strategies

  15. Impact of arbuscular mycorrhizal fungi on uranium accumulation by plants

    Energy Technology Data Exchange (ETDEWEB)

    Dupre de Boulois, H. [Universite catholique de Louvain, Unite de Microbiologie, Croix du Sud 3, 1348 Louvain-la-Neuve (Belgium); Joner, E.J. [Bioforsk Soil and Environment, Fredrik A. Dahls vei 20, N-1432 As (Norway); Leyval, C. [LIMOS, Nancy University, CNRS, Faculte des Sciences, BP239, 54506 Vandoeuvre-les-Nancy, Cedex (France); Jakobsen, I. [Biosystems Department, Riso National Laboratory, Technical University of Denmark, DK-4000 Roskilde (Denmark); Chen, B.D. [Research Centre for Eco-Environmental Sciences, Chinese Academy of Sciences, Beijing 100085 (China); Roos, P. [Radiation Research Department, Riso National Laboratory, Technical University of Denmark, DK-4000 Roskilde (Denmark); Thiry, Y.; Rufyikiri, G. [CEN-SCK, Radiation Protection Research Department, 200 Boeretang, 2400 Mol (Belgium); Delvaux, B. [Universite catholique de Louvain, Unite des Sciences du Sol Croix du Sud 2/10, 1348 Louvain-la-Neuve (Belgium); Declerck, S. [Universite catholique de Louvain, Unite de Microbiologie, Croix du Sud 3, 1348 Louvain-la-Neuve (Belgium)], E-mail: declerck@mbla.ucl.ac.be

    2008-05-15

    Contamination by uranium (U) occurs principally at U mining and processing sites. Uranium can have tremendous environmental consequences, as it is highly toxic to a broad range of organisms and can be dispersed in both terrestrial and aquatic environments. Remediation strategies of U-contaminated soils have included physical and chemical procedures, which may be beneficial, but are costly and can lead to further environmental damage. Phytoremediation has been proposed as a promising alternative, which relies on the capacity of plants and their associated microorganisms to stabilize or extract contaminants from soils. In this paper, we review the role of a group of plant symbiotic fungi, i.e. arbuscular mycorrhizal fungi, which constitute an essential link between the soil and the roots. These fungi participate in U immobilization in soils and within plant roots and they can reduce root-to-shoot translocation of U. However, there is a need to evaluate these observations in terms of their importance for phytostabilization strategies.

  16. Feasibility Study on AFR-100 Fuel Conversion from Uranium-based Fuel to Thorium-based Fuel

    International Nuclear Information System (INIS)

    The feasibility study of converting a fast reactor from uranium-based fuel to thorium-based fuel was studied using the 100 MWe Advanced Fast Reactor (AFR-100). Several fuel conversion scenarios were envisioned in this study. The first scenario is a progressive fuel conversion without fissile support. It consists of progressively replacing the burnt uranium-based fuel with pure thorium-based fuel without fissile material addition. This was found to be impractical because the low excess reactivity of the uranium-fuelled AFR-100 core, resulting in an extremely short cycle length even when only a few assemblies are replaced. A second scenario consists in operating the reference LEU fuelled AFR-100 core for 24 years and then replacing one fuel batch out of four every 7.04 years with thorium-based fuel mixed with transuranics. The transuranics weight fraction required during the transition period is identical to that required at equilibrium and is equal to 18.6%. The original uranium-based fuel is discharged with an average burnup of 120 GWd/t and the Th-TRU fuel with an average burnup of 101 GWd/t. The thermal-hydraulic and passive safety performances of this core are similar to those of the reference AFR-100 design. However, Th-TRU fuel fabrication and performance needs to be demonstrated and TRU separation from the LWR used nuclear fuel is necessary. The third scenario proposed consists of replacing the whole AFR-100 core with fuel assemblies made of several thorium and 20% enriched LEU layers. The mode of operation is similar to that of the reference AFR-100 core with the exception of the cycle length which needs to be reduced from 30 to 18 years. The average LEU and thorium discharge burnups are 79 GWd/t and 23 GWd/t, respectively. The major benefit of this approach is the improved inherent safety of the reactor due to the reduced coolant void worth. (author)

  17. The Fabrication Problem Of U3Si2-Al Fuel With Uranium High Loading

    International Nuclear Information System (INIS)

    The quality of U3Si2-Al dispersion fuel product is the main aim for each fabricator. Low loading of uranium fuel element is easily fabricated, but with the increased, uranium loading, homogeneity of uranium distribution is difficult to achieve and it always formed white spots, blister, and dogboning in the fuel plates. The problem can be eliminated by the increasing treatment of the fuel/Al powder. The precise selection of fuel/Al particles diameter is needed indeed to make easier in the homogeneous process of powder and the porosities arrangement in the fuel plates. The increasing of uranium loading at constant meat thickness will increase the meat hardness, therefore to withdraw the dogboning forming, the use of harder cladding materials is necessity

  18. Lung cancer among workers at a uranium processing plant

    International Nuclear Information System (INIS)

    This study examined the risk of dying from lung cancer among white males who received radiation to the lung as a result of inhaling uranium dust or the dust of uranium compounds. Cases and controls were chosen from a cohort of workers employed in a uranium processing plant during World War II. Cumulative radiation lung dose among study population members ranged from 0 to 75 rads. Relative risk was found to increase with increasing level of exposure even after controlling for age and smoking status, but only for those who were over the age of 45 when first exposed. A statistically significant excess in risk was found for men in this age group with a cumulative lung dose of 20 rads of more. These data suggest that older age groups may be more susceptible to radiation-induced lung cancer than younger age groups

  19. Perimeter safeguards techniques for uranium-enrichment plants

    International Nuclear Information System (INIS)

    In 1972, a working group of the International Atomic Energy Agency identified a goal to develop and evaluate perimeter safeguards for uranium isotope enrichment plants. As part of the United State's response to that goal, Los Alamos Detection and Verification personnel studied gamma-ray and neutron emissions from uranium hexafluoride. They developed instruments that use the emissions to verify uranium enrichment and to monitor perimeter personnel and shipping portals. Unattended perimeter monitors and hand-held verification instruments were evaluated in field measurements and, when possible, were loaned to enrichment facilities for trials. None of the seven package monitoring techniques that were investigated proved entirely satisfactory for an unattended monitor. They either revealed proprietary information about centrifuge design or were subject to interference by shielding materials that could be present in a package. Further evaluation in a centrifuge facility may help in developing an acceptable attended package monitor. 34 figures, 9 tables

  20. Chronicle on the Malargue uranium plant

    International Nuclear Information System (INIS)

    After the Second World War all countries that had industrial development programs, began their works in the nuclear field: if they would achieve satisfactory results depend on their scientific and technological level and on the political choices. Argentina was among those nations that had an industrial development program and very early began a series of actions in the area that at the same time was called 'atomic energy'. Uranium was the basic requirement to design a nuclear program. As a result of the first studies undertaken in the country several uraniferous deposits were discovered. The most important discovery was on May 31, 1952: the Huemul Reservoir located 40 km south of Malargue in the province of Mendoza. The site marks the beginning of the story. (author)

  1. One year of operation of the Belgonucleaire (Dessel) plutonium fuel fabrication plant

    International Nuclear Information System (INIS)

    Based on experience with plutonium since 1958, Belgonucleaire has successively launched a pilot plant and then a fuel fabrication plant for mixed uranium and plutonium oxides in 1968 and 1973 respectively. After describing briefly the plants and the most important stages in the planning, construction and operation of the Dessel plant, the present document describes the principal problems which were met during the course of operation of the plant and their direct incidence on the capacity and quality of the production of fuel elements

  2. SUB-LEU-METAL-THERM-001 SUBCRITICAL MEASUREMENTS OF LOW ENRICHED TUBULAR URANIUM METAL FUEL ELEMENTS BEFORE & AFTER IRRADIATION

    Energy Technology Data Exchange (ETDEWEB)

    SCHWINKENDORF, K.N.

    2006-05-12

    With the shutdown of the Hanford PUREX (Plutonium-Uranium Extraction Plant) reprocessing plant in the 1970s, adequate storage capacity for spent Hanford N Reactor fuel elements in the K and N Reactor pools became a concern. To maximize space utilization in the pools, accounting for fuel burnup was considered. Calculations indicated that at typical fuel exposures for N Reactor, the spent-fuel critical mass would be twice the critical mass for green fuel. A decision was reached to test the calculational result with a definitive experiment. If the results proved positive, storage capacity could be increased and N Reactor operation could be prolonged. An experiment to be conducted in the N Reactor spent-fuel storage pool was designed and assembled and the services of the Battelle Northwest Laboratories (BNWL) (now Pacific Northwest National Laboratory [PNNL]) critical mass laboratory were procured for the measurements. The experiments were performed in April 1975 in the Hanford N Reactor fuel storage pool. The fuel elements were MKIA fuel assemblies, comprising two concentric tubes of low-enriched metallic uranium. Two separate sets of measurements were performed: one with ''green'' (fresh) fuel and one with spent fuel. Both the green and spent fuel, were measured in the same geometry. The spent-fuel MKIA assemblies had an average burnup of 2865 MWd (megawatt days)/t. A constraint was imposed restricting the measurements to a subcritical limit of k{sub eff} = 0.97. Subcritical count rate data was obtained with pulsed-neutron and approach-to-critical measurements. Ten (10) configurations with green fuel and nine (9) configurations with spent fuel are described and evaluated. Of these, 3 green fuel and 4 spent fuel loading configurations were considered to serve as benchmark models. However, shortcomings in experimental data failed to meet the high standards for a benchmark problem. Nevertheless, the data provided by these subcritical measurements can

  3. The logistics and the supply chain in the Juzbado Nuclear Fuel Manufacturing Plant

    International Nuclear Information System (INIS)

    The paper describe the logistics and the supply chain in the Juzbado Nuclear Fuel Manufacturing Plant, located in Juzbado in the province of Salamanca. In the the article are described the principal elements in the supply chain and the difficulties of its management derived from the short period for the manufacturing of the nuclear fuel. It's also given a view in relation to the transportation by land sea of the nuclear components, uranium oxide powder and the manufactured fuel. The characteristics of the supply chain are determined by the plant production forecast, by the origin and high technology of the raw materials and by nuclear fuel delivery site locations. (Author)

  4. Materials Control in the Fabrication of Enriched Uranium Fuels

    International Nuclear Information System (INIS)

    Intense activity in the field of fuel element technology at Oak Ridge National Laboratory during the past 15 years has led to the establishment of sound process and enriched material control procedures that find wide applicability in the commercial fabrication of fuel elements today. Reliable techniques for handling enriched fuel in alloy, dispersion and bulk oxide form were developed and adopted as standards in the course of design and fabrication of prototypic fuel elements for start-up operation of the MTR, Bulk Shielding or ''Swimming Pool'' Reactor, Army Package Power Reactor, Tower Shielding Reactor, Geneva Conference Display Reactor, High Flux Isotope Reactor, and the EGCR. The experience gained serves as background for this paper, which will stress material control problems and their solution during the fabrication of various types of enriched uranium fuel components. The basic objective to be met in the design of a good materials control system are: (1) minimizing the number of material units to be accounted for; (2) designing separate records for each major fabrication step and linking these in a manner that permits isolation of differences with a minimum of effort; (3) integrating the maximum number of controls into the minimum number of records to eliminate duplication; and (4) introducing a sufficient number of cross-checks into the system to ensure reliability. In every fabrication programme, successful control was achieved by establishing a unit procedure in the following areas: (1) starting materials in the as-received form; (2) fabrication of components; (3) component processing; and (4) scrap handling. Consolidation of control records into a master summary was helpful in confirming the materials inventory, evaluating the fabrication process, and preparing management reports. Establishment of sampling methods and examination of results indicated that multiple control is necessary to ensure proper fuel content. Mechanical adjustment and density

  5. Exploring the Response of Plants Grown under Uranium Stress

    International Nuclear Information System (INIS)

    Uranium is a natural element which is mainly redistributed in the environment due to human activity, including accidents and spillages. Plants may be useful in cleaning up after incidents, although little is yet known about the relationship between uranium speciation and plant response. We analyzed the impact of different uranium (U) treatments on three plant species namely sunflower, oilseed rape and wheat. Using inductively coupled plasma mass spectrometry elemental analysis, together with a panel of imaging techniques including scanning electron microscopy coupled with energy dispersive spectroscopy, transmission electron microscopy and particle-induced X-ray emission spectroscopy, we have recently shown how chemical speciation greatly influences the accumulation and distribution of U in plants. Uranyl (UO22+ free ion) is the predominant mobile form in soil surface at low pH in absence of ligands. With the aim to characterize the early plant response to U exposure, complete Arabidopsis transcriptome microarray experiments were conducted on plants exposed to 50 μM uranyl nitrate for 2, 6 and 30 h and highlighted a set of 111 genes with modified expression at these three time points. Quantitative real-time RT-PCR experiments confirmed and completed CATMA micro-arrays results allowing the characterization of biological processes perturbed by U. Functional categorization of deregulated genes emphasizes oxidative stress, cell wall biosynthesis and hormone biosynthesis and signaling. We showed that U stress is perceived by plant cells like a phosphate starvation stress since several phosphate deprivation marker genes were deregulated by U and also highlighted perturbation of iron homeostasis by U. Hypotheses are presented to explain how U perturbs the iron uptake and signaling response. These results give preliminary insights into the pathways affected by uranyl uptake, which will be of interest for engineering plants to help clean areas contaminated with U. (authors)

  6. Exploring the Response of Plants Grown under Uranium Stress

    Energy Technology Data Exchange (ETDEWEB)

    Doustaly, Fany; Berthet, Serge; Bourguignon, Jacques [CEA, iRTSV, Laboratoire de Physiologie Cellulaire Vegetale, UMR 5168 CEA-CNRS-INRA-Univ. Grenoble Alpes (France); Combes, Florence; Vandenbrouck, Yves [CEA, iRTSV, Laboratoire de Biologie a Grande Echelle, EDyP, CEA-Grenoble (France); Carriere, Marie [CEA, INAC, LAN, UMR E3 CEA-Universite Joseph Fourier, Grenoble (France); Vavasseur, Alain [CEA, IBEB, LBDP, Saint Paul lez Durance, CEA Cadarache (France)

    2014-07-01

    Uranium is a natural element which is mainly redistributed in the environment due to human activity, including accidents and spillages. Plants may be useful in cleaning up after incidents, although little is yet known about the relationship between uranium speciation and plant response. We analyzed the impact of different uranium (U) treatments on three plant species namely sunflower, oilseed rape and wheat. Using inductively coupled plasma mass spectrometry elemental analysis, together with a panel of imaging techniques including scanning electron microscopy coupled with energy dispersive spectroscopy, transmission electron microscopy and particle-induced X-ray emission spectroscopy, we have recently shown how chemical speciation greatly influences the accumulation and distribution of U in plants. Uranyl (UO{sub 2}{sup 2+} free ion) is the predominant mobile form in soil surface at low pH in absence of ligands. With the aim to characterize the early plant response to U exposure, complete Arabidopsis transcriptome microarray experiments were conducted on plants exposed to 50 μM uranyl nitrate for 2, 6 and 30 h and highlighted a set of 111 genes with modified expression at these three time points. Quantitative real-time RT-PCR experiments confirmed and completed CATMA micro-arrays results allowing the characterization of biological processes perturbed by U. Functional categorization of deregulated genes emphasizes oxidative stress, cell wall biosynthesis and hormone biosynthesis and signaling. We showed that U stress is perceived by plant cells like a phosphate starvation stress since several phosphate deprivation marker genes were deregulated by U and also highlighted perturbation of iron homeostasis by U. Hypotheses are presented to explain how U perturbs the iron uptake and signaling response. These results give preliminary insights into the pathways affected by uranyl uptake, which will be of interest for engineering plants to help clean areas contaminated with

  7. Design of MOX fuel fabrication plant for Indian PFBR

    International Nuclear Information System (INIS)

    The energy demand is expected to grow rapidly in the coming decades. Fast reactors are expected to be major contributors in sufficing this demand to a great extent. Fast reactors with a co-located fuel cycle facility extend self sufficiency and closing of fuel cycle by recycling of discharged fuel i.e. Plutonium and Uranium. Fuel fabrication facility has a crucial role to play because of limited time period to meet the recurring reload requirements of the reactor. Conventional powder pellet route comprising of various powder metallurgical unit operations is employed for fabrication of fuel. These unit operations are carried out in glove boxes. Automation schemes for handling and transfer of material in the glove box train has been developed. The plant has been analysed for various hazard because of natural and man-made events. Consequence analysis of various postulated events/accidents like criticality, fire, loss of confinement has been carried out. A process layout with branching (hybrid layout) of the glove box train for optimal utilisation of equipment's has been considered. Highly imbalanced time period of different unit operation process necessitates time cycle analysis. Additionally radioactive material inventory limits due to criticality and dose considerations have also been taken into account in the analysis. A code has been developed to simulate this material movement through the optimized process layout. This paper discusses various issues mentioned above for design of a MOX fuel fabrication plant

  8. Uptake of uranium and thorium by native and cultivated plants

    International Nuclear Information System (INIS)

    Large part of available literature on biogeochemistry of uranium and thorium refers to the studies performed either in highly contaminated areas or in nutrient solutions that have been artificially 'spiked' with radionuclides. Effects of background levels of natural radioactivity on soil-grown plants have not been studied to the same extent. In this paper, we summarised results of greenhouse and field experiments performed by the author from 2000 to 2006. We examined some of the factors affecting transfer of U and Th from soil to plants, differences in uptake of these radionuclides by different plants, relationships between U and Th in soil and in plants, and temporal variations of U and Th in different plant species. Concentrations of radionuclides (critical point for experimental studies on biogeochemistry of U and Th - rare trace elements in non-contaminated regions) and essential plant nutrients and trace elements were determined by instrumental neutron activation analysis.

  9. Activities in front-end of uranium fuel cycle in IAEA

    International Nuclear Information System (INIS)

    The Nuclear Fuel Cycle and Materials Section (NFC and MS) in the Division of Nuclear Fuel Cycle and Waste Technology (NEFW) under the Department of Nuclear Energy (NE) of IAEA implements Major Programme 1.B. of the Agency. NFC and MS fosters development of nuclear fuel cycle options that are safe, environment-friendly, economically viable and proliferation-resistant. It promotes information exchange on exploration, mining and processing of uranium and thorium, design, manufacturing, and performance of nuclear fuels, management of spent fuel, including storage and treatment of spent fuel and recycling of plutonium and uranium fuels, and development of advanced and innovative nuclear fuels and fuel cycles through Technical Co-operation, preparation of state-of-theart technical documents, technical meetings, symposia and Coordinated Research Projects (CRP) and databases on nuclear fuels and fuel cycles. The present paper summarizes the portions of the Major Programme 1.B. of IAEA, related to the front-end of uranium fuel cycle, highlighting the activities on uranium supply and demand, exploration, production cycle and environment, the databases, and the technical documents (IAEA/TECDOC) that have been published or under preparation in these areas during the last five years. It reports on the IAEA/OECD-NEA Uranium Red Book, the IAEA database on world distribution of uranium deposits (UDEPO)and the IAEA database on nuclear fuel cycle information system (INFCIS). It discusses the environmental issues in the front-end of the uranium production cycle including mining, milling, chemical purification and long-term management of mine tails, residual materials and radioactive wastes, which are of paramount importance to the uranium industry. The Agency provides guidance on best practices in the planning, operation and closure of uranium production facilities including mine reclamation, from the perspective of changing environmental regulations in mining facilities and growing

  10. IAEA activities on uranium resources and production and databases for the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Full text: Natural uranium (∼99.3% U-238+ ∼ 0.7 % U-235) is the basic raw material for nuclear fuel. The present generation of nuclear power reactors derive energy from the 'fission' of U-235, the only 'fissile' isotope in nature. These reactors also transmute the more abundant U-238 to man-made fissile isotope Pu-239, which could be subjected to multiple recycling, as fuel, in fast reactor for efficient utilization of natural uranium resources and to ensure long term sustainability of nuclear energy. Uranium is mostly mined and produced in countries without a nuclear power programme. On the other hand, uranium is mostly consumed in countries with nuclear power, but having no uranium. In recent years, rising expectation for nuclear power has led to increase in uranium exploration, mining and ore processing activities all over the world and several new countries, with a limited experience, have embarked on uranium exploration, mining and production. Uranium and its daughter products are radioactivity and health hazardous. Radiological safety is a major challenge in uranium production cycle and in uranium mine and mill remediation and reclamation. Another specific challenge being faced currently by uranium raw material industry is the retired or ageing manpower and lack of experienced staff around the world. The IAEA's programme on 'Uranium Resources and Production and Databases for the Nuclear Fuel Cycle' encompass all aspects of uranium geology and deposits, exploration, resources, supply and demand, uranium mining and processing, environmental issues related to uranium production cycle and databases for uranium fuel cycle. The IAEA collaborates with OECD/NEA in producing an authoritative and updated document on uranium resources, production and demand, popularly known as Red Book, which is published biennially by OECD/NEA. As a spin-off from uranium resources activities, two reports titled, 'Analysis of Uranium to 2060' and 'Red Book Retrospective - Country

  11. Uranium hexafluoride - chemistry and technology of a raw material of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Uranium hexafluoride exhibits an unusual combination of properties: UF6 is both a large-scale industrial product, and also one of the most reactive compounds known. Its industrial application arises from the need to use enriched uranium with up to 4% 235U as fuel in light water reactors. Enrichment is performed in isotope separation plants with UF6 as the working gas. Its volatility and thermal stability make UF6 suitable for this application. UF6 handling is difficult because of its high reactivity and its radioactivity, and special experience and equipment are required which are not commonly available in laboratories or industrial facilities. The chemical reactions of UF6 are characterized by its marked fluorination efficiency which is similar to that of F2. Of special importance in connection with the handling of UF6 is its extreme sensitivity to hydrolysis. Because they all use UF6, the isotope separation processes currently in use (gas diffusion, gas centrifuge, separation nozzle process) have a number of common features. For instance, they are all beset by the problem of formation of solid UF6 decomposition products, e.g. by radiolysis of UF6 molecules induced by its own radiation. Reconversion of UF6 into UO2 is achieved by three well-known methods (ADU, AUC, IDP-process). To produce uranium metal, UF6 is first reduced to UF4, which is subsequently reduced by Ca6 or Mg metal. 158 refs

  12. Uranium Oxide Rate Summary for the Spent Nuclear Fuel (SNF) Project (OCRWM)

    International Nuclear Information System (INIS)

    The purpose of this document is to summarize the uranium oxidation reaction rate information developed by the Hanford Spent Nuclear Fuel (SNF) Project and describe the basis for selecting reaction rate correlations used in system design. The selection basis considers the conditions of practical interest to the fuel removal processes and the reaction rate application during design studies. Since the reaction rate correlations are potentially used over a range of conditions, depending of the type of evaluation being performed, a method for transitioning between oxidation reactions is also documented. The document scope is limited to uranium oxidation reactions of primary interest to the SNF Project processes. The reactions influencing fuel removal processes, and supporting accident analyses, are: uranium-water vapor, uranium-liquid water, uranium-moist air, and uranium-dry air. The correlation selection basis will consider input from all available sources that indicate the oxidation rate of uranium fuel, including the literature data, confirmatory experimental studies, and fuel element observations. Trimble (2000) summarizes literature data and the results of laboratory scale experimental studies. This document combines the information in Trimble (2000) with larger scale reaction observations to describe uranium oxidation rate correlations applicable to conditions of interest to the SNF Project

  13. Uranium Oxide Rate Summary for the Spent Nuclear Fuel (SNF) Project (OCRWM)

    Energy Technology Data Exchange (ETDEWEB)

    PAJUNEN, A.L.

    2000-09-20

    The purpose of this document is to summarize the uranium oxidation reaction rate information developed by the Hanford Spent Nuclear Fuel (SNF) Project and describe the basis for selecting reaction rate correlations used in system design. The selection basis considers the conditions of practical interest to the fuel removal processes and the reaction rate application during design studies. Since the reaction rate correlations are potentially used over a range of conditions, depending of the type of evaluation being performed, a method for transitioning between oxidation reactions is also documented. The document scope is limited to uranium oxidation reactions of primary interest to the SNF Project processes. The reactions influencing fuel removal processes, and supporting accident analyses, are: uranium-water vapor, uranium-liquid water, uranium-moist air, and uranium-dry air. The correlation selection basis will consider input from all available sources that indicate the oxidation rate of uranium fuel, including the literature data, confirmatory experimental studies, and fuel element observations. Trimble (2000) summarizes literature data and the results of laboratory scale experimental studies. This document combines the information in Trimble (2000) with larger scale reaction observations to describe uranium oxidation rate correlations applicable to conditions of interest to the SNF Project.

  14. Development of an accident-tolerant fuel composite from uranium mononitride (UN) and uranium sesquisilicide (U3 Si2) with increased uranium loading

    Science.gov (United States)

    Ortega, Luis H.; Blamer, Brandon J.; Evans, Jordan A.; McDeavitt, Sean M.

    2016-04-01

    The processing steps necessary to prepare a potential accident-tolerant fuel composite consisting of uranium mononitride (UN) combined with uranium sesquisilicide (U3 Si2) are described. Liquid phase sintering was performed with U3 Si2 as the liquid phase combined with UN powder or UN μ-spheres. Various UN to U3 Si2 ratios were tested which resulted in up to 94% dense pellets. Composite UN-U3 Si2 samples had greater than 30% more uranium content than UO2.

  15. Occupational safety data and casualty rates for the uranium fuel cycle

    International Nuclear Information System (INIS)

    Occupational casualty (injuries, illnesses, fatalities, and lost workdays) and production data are presented and used to calculate occupational casualty incidence rates for technologies that make up the uranium fuel cycle, including: mining, milling, conversion, and enrichment of uranium; fabrication of reactor fuel; transportation of uranium and fuel elements; generation of electric power; and transmission of electric power. Each technology is treated in a separate chapter. All data sources are referenced. All steps used to calculate normalized occupational casualty incidence rates from the data are presented. Rates given include fatalities, serious cases, and lost workdays per 100 man-years worked, per 1012 Btu of energy output, and per other appropriate units of output

  16. Decommissioning of the Siemens Hanau fuel fabrication plant and hot cells

    International Nuclear Information System (INIS)

    In the early and mid-1990s, a series of decisions had to be made - partially as a result of political requirements but also, in some cases, for economic reasons - to permanently shut down four facilities operated by the Nuclear Fuel Cycle Division of Siemens' Power Generation Group (KWU). 1989 saw the closure of the hot cells in Karlstein in which Germany's most extensive post-irradiation examinations of fuel assemblies had been carried out since 1967. In 1994/95, manufacture of gadolinium-bearing uranium fuel assemblies was abandoned at the Siemens Karlstein Fuel Fabrication Plant which had been in operation since 1963. At the Siemens Hanau Fuel Fabrication Plant, the facilities for manufacturing mixed-oxide (MOX) fuel assemblies and uranium fuel assemblies were permanently shut down in 1991 and 1995, respectively. The uranium processing facility had been in operation since 1969, and the MOX processing facility since 1970. Shutdown and decommissioning of these four facilities have mainly been proceeding in the following stages. First of all the facilities are cleaned out and all process equipment is removed. Then the auxiliary and support systems are dismantled. Finally the buildings are decontaminated and, in some cases, demolished. Possibly contaminated soil will be removed and the site restorated, after which it is released for unrestricted use and is no longer subject to the licensing requirements of the German Atomic Energy Act. Nuclear fuel materials as well as a few of the process components have been given to other nuclear fuel manufacturers. (orig.)

  17. Preparation of uranium oxide powder for nuclear fuel pellet fabrication with uranium peroxide recovered from uranium oxide scraps by using a carbonate-hydrogen peroxide solution

    International Nuclear Information System (INIS)

    This work studied a way to reclaim uranium from contaminated UO2 oxide scraps as a sinterable UO2 powder for UO2 fuel pellet fabrication, which included a dissolution of the uranium oxide scraps in a carbonate solution with hydrogen peroxide and a UO4 precipitation step. Dissolution characteristics of reduced and oxidized uranium oxides were evaluated in a carbonate solution with hydrogen peroxide, and the UO4 precipitation were confirmed by acidification of uranyl peroxo-carbonate complex solution. An agglomerated UO4 powder obtained by the dissolution and precipitation of uranium in the carbonate solution could not be pulverized into fine UO2 powder by the OREOX process, because of submicron-sized individual UO4 particles forming the agglomerated UO4 precipitate. The UO2 powder prepared from the UO4 precipitate could meet the UO2 powder specifications for UO2 fuel pellet fabrication by a series of steps such as dehydration of UO4 precipitate, reduction, and milling. The sinterability of the reclaimed UO2 powder for fuel pellet fabrication was improved by adding virgin UO2 powder in the reclaimed UO2 powder. A process to reclaim the contaminated uranium scraps as UO2 fuel powder using a carbonate solution was finally suggested. (author)

  18. Medium-enriched uranium/thorium fuel cycle parametric studies for the HTGR

    International Nuclear Information System (INIS)

    Operation of HTGRs on proliferation-resistant medium-enriched uranium/thorium fuel cycles is feasible based on the findings of fuel cycle parametric studies conducted for the Department of Energy by General Atomic Company. The analyses performed to evaluate the feasibility and optimization of such fuel cycles are described. Primary variables considered in arriving at optimum designs included cycle length, fuel particle and fuel rod dimensions, the carbon-to-thorium ratio, and the refueling frequency

  19. Atomistic Simulation of High-Density Uranium Fuels

    Directory of Open Access Journals (Sweden)

    Jorge Eduardo Garcés

    2011-01-01

    Full Text Available We apply an atomistic modeling approach to deal with interfacial phenomena in high-density uranium fuels. The effects of Si, as additive to Al or as U-Mo-particles coating, on the behavior of the Al/U-Mo interface is modeled by using the Bozzolo-Ferrante-Smith (BFS method for alloys. The basic experimental features characterizing the real system are identified, via simulations and atom-by-atom analysis. These include (1 the trend indicating formation of interfacial compounds, (2 much reduced diffusion of Al into U-Mo solid solution due to the high Si concentration, (3 Si depletion in the Al matrix, (4 an unexpected interaction between Mo and Si which inhibits Si diffusion to deeper layers in the U-Mo solid solution, and (5 the minimum amount of Si needed to perform as an effective diffusion barrier. Simulation results related to alternatives to Si dispersed in the Al matrix, such as the use of C coating of U-Mo particles or Zr instead of the Al matrix, are also shown. Recent experimental results confirmed early theoretical proposals, along the lines of the results reported in this work, showing that atomistic computational modeling could become a valuable tool to aid the experimental work in the development of nuclear fuels.

  20. Uranium accumulation by aquatic plants from uranium-contaminated water in Central Portugal.

    Science.gov (United States)

    Pratas, João; Favas, Paulo J C; Paulo, Carlos; Rodrigues, Nelson; Prasad, M N V

    2012-03-01

    Several species of plants have developed a tolerance to metal that enables them to survive in metal contaminated and polluted sites. Some of these aquatic plants have been reported to accumulate significant amounts of specific trace elements and are, therefore, useful for phytofiltration. This work focuses the potential of aquatic plants for the phytofiltration of uranium (U) from contaminated water. We observed that Callitriche stagnalis, Lemna minor, and Fontinalis antipyretica, which grow in the uraniferous geochemical province of Central Portugal, have been able to accumulate significant amounts of U. The highest concentration of U was found in Callitriche stagnalis (1948.41 mg/kg DW), Fontinalis antipyretica (234.79 mg/kg DW), and Lemna minor (52.98 mg/kg DW). These results indicate their potential for the phytofiltration of U through constructed treatment wetlands or by introducing these plants into natural water bodies in the uraniferous province of Central Portugal. PMID:22567707

  1. Analysis of the Reuse of Uranium Recovered from the Reprocessing of Commercial LWR Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    DelCul, Guillermo Daniel [ORNL; Trowbridge, Lee D [ORNL; Renier, John-Paul [ORNL; Ellis, Ronald James [ORNL; Williams, Kent Alan [ORNL; Spencer, Barry B [ORNL; Collins, Emory D [ORNL

    2009-02-01

    This report provides an analysis of the factors involved in the reuse of uranium recovered from commercial light-water-reactor (LWR) spent fuels (1) by reenrichment and recycling as fuel to LWRs and/or (2) by recycling directly as fuel to heavy-water-reactors (HWRs), such as the CANDU (registered trade name for the Canadian Deuterium Uranium Reactor). Reuse is an attractive alternative to the current Advanced Fuel Cycle Initiative (AFCI) Global Nuclear Energy Partnership (GNEP) baseline plan, which stores the reprocessed uranium (RU) for an uncertain future or attempts to dispose of it as 'greater-than-Class C' waste. Considering that the open fuel cycle currently deployed in the United States already creates a huge excess quantity of depleted uranium, the closed fuel cycle should enable the recycle of the major components of spent fuel, such as the uranium and the hazardous, long-lived transuranic (TRU) actinides, as well as the managed disposal of fission product wastes. Compared with the GNEP baseline scenario, the reuse of RU in the uranium fuel cycle has a number of potential advantages: (1) avoidance of purchase costs of 11-20% of the natural uranium feed; (2) avoidance of disposal costs for a large majority of the volume of spent fuel that is reprocessed; (3) avoidance of disposal costs for a portion of the depleted uranium from the enrichment step; (4) depending on the {sup 235}U assay of the RU, possible avoidance of separative work costs; and (5) a significant increase in the production of {sup 238}Pu due to the presence of {sup 236}U, which benefits somewhat the transmutation value of the plutonium and also provides some proliferation resistance.

  2. Advanced spent-fuel waste package fill material: Depleted uranium dioxide

    International Nuclear Information System (INIS)

    The use of depleted uranium dioxide (DUO2) particles has been investigated as fill material inside repository waste packages containing light water reactor (LWR) spent nuclear fuel (SNF). The use of DUO2 fill may eliminate repository criticality concerns, reduce radionuclide release rates from the repository, and dispose of excess depleted uranium

  3. 78 FR 63518 - Uranium Enrichment Fuel Cycle Inspection Reports Regarding Louisiana Energy Services, National...

    Science.gov (United States)

    2013-10-24

    ..., New Mexico, and has authorized the introduction of uranium hexafluoride (UF 6 ) into cascades numbered... COMMISSION Uranium Enrichment Fuel Cycle Inspection Reports Regarding Louisiana Energy Services, National... 4th day of October, 2013. For the U.S. Nuclear Regulatory Commission. Brian W. Smith, Chief,...

  4. Comparison of the radiological impacts of thorium and uranium nuclear fuel cycles

    International Nuclear Information System (INIS)

    This report compares the radiological impacts of a fuel cycle in which only uranium is recycled, as presented in the Final Generic Environmental Statement on the Use of Recycle Plutonium in Mixed Oxide Fuel in Light Water Cooled Reactors (GESMO), with those of the light-water breeder reactor (LWBR) thorium/uranium fuel cycle in the Final Environmental Statement, Light Water Breeder Reactor Program. The significant offsite radiological impacts from routine operation of the fuel cycles result from the mining and milling of thorium and uranium ores, reprocessing spent fuel, and reactor operations. The major difference between the impacts from the two fuel cycles is the larger dose commitments associated with current uranium mining and milling operations as compared to thorium mining and milling. Estimated dose commitments from the reprocessing of either fuel type are small and show only moderate variations for specific doses. No significant differences in environmental radiological impact are anticipated for reactors using either of the fuel cycles. Radiological impacts associated with routine releases from the operation of either the thorium or uranium fuel cycles can be held to acceptably low levels by existing regulations

  5. Assessment of the implications of conversion of university research and training reactors to low enrichment uranium fuel

    International Nuclear Information System (INIS)

    The tasks associated with conversion of a research reactor from HEU to LEU fuel are: initial program planning; safety analysis and license amendment; core physics calculations; operating thermal-hydraulics analysis; plant engineering modifications; LEU fuel specifications, procurement of fuel, and calculational confirmation of design; training of staff personnel; HEU core physics measurements and fuel disposal; and experimental verification of reactor behavior with LEU fuel. LEU fuel conversion of the 25 NRC licensed, university-owned reactors considered in this study is based upon the reactor fuel cycle, the type of license modification, and fuel meat technology. Reactors that operate on routine refueling cycles could periodically replace depleted HEU elements with fresh LEU elements. Ultimate full core conversion would depend on the average element residence time in the core. Reactors with lifetime cores would convert by full core replacement as a one-time event. For some reactors, LEU conversion depends upon high density uranium fuel meat technology development. The majority should be able to convert using a direct substitution of current fuel meat technology though some fuel plate or rod internal modifications may be necessary for 16 of the reactors

  6. Criticality Safety of Low-Enriched Uranium and High-Enriched Uranium Fuel Elements in Heavy Water Lattices

    International Nuclear Information System (INIS)

    The RB reactor was designed as a natural-uranium, heavy water, non reflected critical assembly in the Vinca Institute of Nuclear Sciences, Belgrade, Yugoslavia, in 1958. From 1962 until 2002, numerous critical experiments were carried out with low-enriched uranium and high-enriched uranium fuel elements of tubular shape, known as the Russian TVR-S fuel assembly type, placed in various heavy water square lattices within the RB cylindrical aluminum tank. Some of these well-documented experiments were selected, described, evaluated, and accepted for inclusion in the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments', contributing to the preservation of a rather small number of heavy water benchmark critical experiments. (author)

  7. Design of Uranium Isotope Separation Plant by Chemical Exchange

    International Nuclear Information System (INIS)

    The methodology to design a solvent extraction plant for uranium isotope separation by chemical exchange is outlined. This process involves the calculator of the number of stages,the capacity of the plant,the flow rates,and reflux ration in banks of mixer settlers or pulse column used in such a plant. The feed is introduced at the middle of the plant,and the product is withdrawn at one end and the tailings at another. The redox reaction system selected is U(IV)-U(VI) and the equilibrium data of the 40% tri-n-octylamine (TOA) in benzene as the organic phase and 4 M HCI as the aqueous phase are used for the design of the real plant. The resulting analysis for the uranium isotope separation shows that more than 4000 number of stages are required and the reflux ratio is around 700 to produce only 1m3 of product containing 3% of U235 and 0,3% of U235 in the tailings. It is also known that the larger the isotope separation constant the smaller the number of stages needed. The method of design can be used for other systems where the isotope separation constants are more favorable

  8. Environmental sciences: general. 4. Radiological and Chemical Risks in the Canadian Uranium Fuel Cycle

    International Nuclear Information System (INIS)

    management area for disposal. Although there are many aboveground tailings management areas in Canada, the current approach is to place tailings in mined-out open pits that have been re-engineered to receive tailings. While the radioactivity released to the environment is a concern both during operation and following decommissioning, the potential release of arsenic and other ore constituents to the aquatic environment can be as contentious as the release of radioactivity. In refining, uranium concentrate (yellowcake) from uranium mills is converted to uranium trioxide for subsequent processing in the conversion facility to uranium hexafluoride (UF6) or to ceramic-grade uranium dioxide (UO2).The UF6 is sent to the United States or overseas for enrichment and use in nuclear power reactors; the UO2 is sent directly to fuel fabrication facilities where it is made into fuel assemblies for use in Candu reactors. In addition to the potential radiological hazards, there is also some potential for the release of anhydrous ammonia from the refining facility and of anhydrous ammonia (NH3), anhydrous hydrogen fluoride (HF), or UF6 from the conversion facility. It could be argued that the refining and conversion facilities are in fact largely chemical plants that simply happen to have a radioactive feed material. Detailed assessments of the potential for, and consequences of, accidental releases of these chemicals have been carried out. The potential hazards considered in this paper are summarized in Table I. This paper describes the front end of the Canadian fuel cycle and briefly examines the risks arising from the major radiological and chemical hazards noted in Table I and comments on the ways in which the risks are managed. The potential risks and the ways they are managed are illustrated with specific examples for both safety and environmental hazards. The ways in which the radiological and chemical hazards are assessed and managed are compared. The author concludes that

  9. Uranium/fuel cycle 74, New Orleans, Louisiana, 17--20 March 1974. Program report

    International Nuclear Information System (INIS)

    The highlight of papers presented at the conference are summarized. The sessions covered uranium raw material, transportation of spent fuel and radioactive waste, plutonium recycle, waste management, and safeguards. (U.S.)

  10. Construction and operation program of reprocessing plant of the Power Reactor and Nuclear Fuel Development Corporation

    International Nuclear Information System (INIS)

    The reprocessing plant of the Power Reactor and Nuclear Fuel Development Corporation which is constructed in Tokai is operated according to the purex system, and the following design parameters; the capacity is 0.7 t/day, the burn-up of fuel is about 28,000 MWd/t, the cooling period is more than 180 days, the uranium enrichment is less than 4%, the maximum dimensions of fuel assembly is 470cm x 26cm x 26cm and the maximum weight of the fuel assembly is 1.2 t. The flow diagram of the reprocessing system in this plant, the plant layout at the site, the main specifications of the civil engineering, the main instrumentation cask specifications, the functional test of these facilities especially for the cleaning of the components and pipings by compressed air, water and steam, the inspection of the volumes of tanks, the characteristic test of components, the water flow test of piping system, the chemical characteristic test, waste disposal test, the test using uranium, the organization and the training of operators in this plant and the safety considerations are described in detail in this paper. The hot test will be conducted with the spent fuel which is taken from JPDR at first and then from BWR and PWR plants in Japan step by step. The hot test with the OTL facility using uranium is particularly explained in detail. (Nakai, Y.)

  11. Investigation of different scenarios of thorium–uranium fuel distribution in the VVER-1200 first core

    International Nuclear Information System (INIS)

    Highlights: • Insertion of thorium on neutronic parameters of VVER-1200 first core is studied. • Two patterns, mixed thorium uranium fuel and seed-blanket fuel, were compared. • Position of thorium assemblies in core is important to determine the cycle length. • It was concluded that the best location of thorium is in the periphery of the core. - Abstract: Thorium fuel gives some superiority from safeguards point of view and is now being considered as an option for nuclear fuel. It is found to achieve high proliferation resistance because it has significant lower production of plutonium and minor actinides as compared with uranium fuel. The effect of the insertion of thorium, as a part of the nuclear fuel, on the neutronic parameters of the VVER-1200 first core under normal operation was studied. Two different patterns, namely mixed thorium uranium fuel and seed-blanket fuel, were compared. In addition to the amount of the inserted thorium, it was found that the position of the thorium assemblies inside the reactor core plays an important role in determining the effective multiplication factor and hence the core cycle length. It was concluded that the best location of thorium is in the periphery of the reactor core. As the preparation of thorium fuel does not involve all the requirements imposed by the uranium fuel, it is therefore expected that the economical feedback of both uranium–thorium fuels will be also positive, especially for a country with high thorium abundance

  12. Uranium separation from phosphates and the fuel cycle process

    International Nuclear Information System (INIS)

    A short introduction on the recycle of uranium and plutonium is presented. The uranium world market at present, the prices during the last few years, the actual requirements and those for the years 1978-1983 are given. In a special paragraph the present resources of uranium in Israel as well as the extraction possibilities are discussed. (B.G.)

  13. Optimal sizes and siting of nuclear fuel reprocessing plants

    International Nuclear Information System (INIS)

    The expansion of a nuclear economy entails the development of fuel process and reprocessing plant programmes. The model proposed makes it possible to select the size, the site and the start-up schedule of the plants in such a way as to minimize the total freight and reprocessing costs. As an illustration, we have approached the problem of burnt natural uranium processing plants related to natural uranium-graphite as nuclear power stations. The sites and annual output of the reactors, the possible plant sites and cost functions (freight and reprocessing) are supposed to be known. The method consists in first approaching the process plant problem as a Dynamic Programming problem, increasing programme slices (total reactor output) being explored sequentially. When the quantities of burnt natural uranium to be reprocessed are fixed, the minimization of the transport cost is then also carried out as a dynamic programming problem. The neighbourhood of the optimum process cost is explored in order to find the minimum summation of a suboptimal processing cost and corresponding optimal transport cost. As the reprocessing problem can be represented on a sequential graph, in order to compute the sub-optima, we developed and used a 'reflexion algorithm'. The method can be interpreted as a general mechanism for determining the optimum when to a sequential dynamic problem (for example an equipment programme) is added a complementary problem (transport, for instance). It also makes it possible to estimate the economic losses which result from the choice of a non optimal policy for other than economic reasons. (author)

  14. Natural uranium fueled light water moderated breeding hybrid power reactors: a feasibility study

    International Nuclear Information System (INIS)

    The first part of the study consists of a thorough investigation of the properties of subcritical thermal lattices for hybrid reactor applications. Light water is found to be the best moderator for (fuel-self-sufficient) FSS hybrid reactors for power generation. Several lattice geometries and compositions of particular promise for LWHRs are identified. Using one of these lattices, fueled with natural uranium, the performance of several concepts of LWHR blankets is investigated, and optimal blanket designs are identified. The effect of blanket coverage efficiency and the feasibility of separating the functions of tritium breeding and of power generation to different blankets are investigated. Optimal iron-water shields for LWHRs are also determined. The performance of generic types of LWHRs is evaluated. The evolution of the blanket properties with burnup is evaluated and fuel management schemes are briefly examined. The feasibility of using the lithium system of the blanket to control the blanket power amplitude and shape is also investigated. A parametric study of the energy balance of LWHR power plants is carried out, and performance parameters expected from LWHRs are estimated. Discussions are given of special features of LWHRs and their fuel cycle

  15. The nuclear fuel cycle

    International Nuclear Information System (INIS)

    After a short introduction about nuclear power in the world, fission physics and the French nuclear power plants, this brochure describes in a digest way the different steps of the nuclear fuel cycle: uranium prospecting, mining activity, processing of uranium ores and production of uranium concentrates (yellow cake), uranium chemistry (conversion of the yellow cake into uranium hexafluoride), fabrication of nuclear fuels, use of fuels, reprocessing of spent fuels (uranium, plutonium and fission products), recycling of energetic materials, and storage of radioactive wastes. (J.S.)

  16. Feeding the nuclear fuel cycle with a long term view; AREVA's front-end business units, uranium mining, UF6 conversion and isotopic enrichment

    International Nuclear Information System (INIS)

    As a leading provider of technological solutions for nuclear power generation and electricity transmission, the AREVA group has the unique capability of offering a fully integrated fuel supply, when requested by its customers. At the core of the AREVA group, COGEMA Front End Division is an essential part of the overall fuel supply chain. Composed of three Business Units and gathering several subsidiaries and joint 'ventures, this division enjoys several leading positions as shown by its market shares and historical production records. Current Uranium market evolutions put the natural uranium supply under focus. The uranium conversion segment also recently revealed some concerning evolutions. And no doubt, the market pressure will soon be directed also at the enrichment segment. Looking towards the long term, AREVA strongly believes that a nuclear power renewal is needed, especially to help limiting green house effect gas release. Therefore, to address future supplies needed to fuel the existing fleet of nuclear power plants, but also new ones, the AREVA group is planning very significant investments to build new facilities in all the three front-end market segments. As far as uranium mining is concerned, these new mines will be based upon uranium reserves of outstanding quality. As for uranium conversion and enrichment, two large projects will be based on the most advanced technologies. This paper is aimed at recalling COGEMA Front End Division experience, the current status of its plants and operating entities and will provide a detailed overview of its major projects. (authors)

  17. Cold rolling technique for U3Si2-Al dispersion fuel plates with high uranium loading

    International Nuclear Information System (INIS)

    A new rolling procedure alternated between cold rolling and annealing is described The high uranium-loading U3Si2-Al dispersion fuel pates with uniform thickness of cladding ad meat are fabricated by this procedure. The fuel plates have good metallurgical bond. The minimum thickness of cladding is not less then 0. 25 mm, 7g/cm-3 equivalent uranium loading or even more can be achieved. (author)

  18. The decommissioning of the Barnwell nuclear fuel plant

    International Nuclear Information System (INIS)

    The decommissioning of the Barnwell Nuclear Fuel Plant is nearing completion. The owner's objective is to terminate the plant radioactive material license associated with natural uranium and transuranic contamination at the plant. The property is being released for commercial-industrial uses, with radiation exposure from residual radioactivity not to exceed 0.15 millisieverts per year. Historical site assessments have been performed and the plant characterized for residual radioactivity. The decommissioning of the uranium hexafluoride building was completed in April, 1999. Most challenging from a radiological control standpoint is the laboratory building that contained sixteen labs with a total of 37 glove boxes, many of which had seen transuranics. Other facilities being decommissioned include the separations building and the 300,000-gallon underground high-level waste tanks. This decommissioning in many ways is the most significant project of this type yet undertaken in South Carolina. Many innovations have been made to reduce the time and costs associated with the project. (author)

  19. Model of a Generic Natural Uranium Conversion Plant ? Suggested Measures to Strengthen International Safeguards

    Energy Technology Data Exchange (ETDEWEB)

    Raffo-Caiado, Ana Claudia [ORNL; Begovich, John M [ORNL; Ferrada, Juan J [ORNL

    2009-11-01

    This is the final report that closed a joint collaboration effort between DOE and the National Nuclear Energy Commission of Brazil (CNEN). In 2005, DOE and CNEN started a collaborative effort to evaluate measures that can strengthen the effectiveness of international safeguards at a natural uranium conversion plant (NUCP). The work was performed by DOE s Oak Ridge National Laboratory and CNEN. A generic model of a NUCP was developed and typical processing steps were defined. Advanced instrumentation and techniques for verification purposes were identified and investigated. The scope of the work was triggered by the International Atomic Energy Agency s 2003 revised policy concerning the starting point of safeguards at uranium conversion facilities. Prior to this policy only the final products of the uranium conversion plant were considered to be of composition and purity suitable for use in the nuclear fuel cycle and therefore, subject to the IAEA safeguards control. DOE and CNEN have explored options for implementing the IAEA policy, although Brazil understands that the new policy established by the IAEA is beyond the framework of the Quadripartite Agreement of which it is one of the parties, together with Argentina, the Brazilian-Argentine Agency for Accounting and Control of Nuclear Materials (ABACC) and the IAEA. Two technical papers on this subject were published at the 2005 and 2008 INMM Annual Meetings.

  20. Development and Evaluation of Mixed Uranium-Refractory Carbide/Refractory Carbide Cer-Cer Fuels Project

    Data.gov (United States)

    National Aeronautics and Space Administration — In this proposal a new carbide-based fuel is introduced with outstanding potential to eliminate the loss of uranium, minimizes the loss of uranium, and retains...

  1. Development of high uranium-density fuels for use in research reactors

    International Nuclear Information System (INIS)

    The uranium silicide U3Si2 possesses uranium density 11.3 gU/cm3 with a congruent melting point of 1665degC, and is now successfully in use as a research reactor fuel. Another uranium silicide U3Si and U6Me-type uranium alloys (Me=Fe,Mn,Ni) have been chosen as new fuel materials because of the higher uranium densities 14.9 and 17.0 gU/cm3, respectively. Experiments were carried out to fabricate miniature aluminum-dispersion plate-type and aluminum-clad disk-type fuels by using the conventional picture-frame method and a hot-pressing technique, respectively. These included the above-mentioned new fuel materials as well as U3Si2. Totally 14 miniplates with uranium densities from 4.0 to 6.3 gU/cm3 of fuel meat were prepared together with 28 disk-type fuel containing structurally-modified U3Si, and subjected to the neutron irradiation in JMTR (Japan Materials Testing Reactor). Some results of postirradiation examinations are presented. (author)

  2. Safety aspects of the FMPP (Fuel Manufacturing Pilot Plant) setup constructed by INVAP in the Arabic Republic of Egypt

    International Nuclear Information System (INIS)

    The FMPP is a fuel plates manufacturing plant for test reactors. This facility was designed, constructed in El Cairo and turned-key handled by INVAP SE to the Arabian Republic of Egypt. In this project, CNEA participated in the transference of technology, elaboration of documents, training of Egyptian personnel and technical services during the setup of the facility in El Cairo. These tasks were undertaken by UPMP (Uranium Powder Manufacturing Plant) and ECRI (Research Reactors Fuel Elements Plant) personnel. Both plants in CNEA served as a FMPP design basis. During the setup of the facility a fuel element with natural uranium was firstly manufactured and then another one using uranium with 20% enrichment. In this paper the responses of the system regarding safety, after finishing the first two stages of manufacturing, are analyzed and evaluated. (author)

  3. Plant and soil relationships of uranium and thorium decay series radionuclides - a review

    International Nuclear Information System (INIS)

    The behavior of radionuclides of the uranium (U) and thorium (Th) decay series in terrestrial systems is of interest because of environmental effects of mining and disposal activities related to nuclear power plant fuels. The soil-plant relationships of U, Th, and polonium (Pb), and some other daughter radionuclides, notably radium (226Ra), are not well understood. Most studies have been concerned with relative uptake of these radionuclides by various plant species. Plant concentrations have been related to total contents of these radionuclides in the soil as a plant/soil concentration ratio (CR), even though the fraction of these radionuclides, which may be available to plants, is not well known. These CR values have been used to predict transport of radionuclides and other elements of interest through the food chain as well as for other purpose including biogeochemical exploration for U. Little information is available on uptake and transport mechanisms of radionuclides in plants. However, the mechanisms relating to Ca uptake and translocation in plants may be similar to those of some radionuclides, especially 226Ra. Son chemical reactions of these radionuclides also have not been studied as well as those of plant nutrients, although knowledge of the effects of soil pH, soil texture, and organic matter content on uptake, as well as mobility in soil of these radionuclides, has been gained in recent years. 45 refs., 13 tabs

  4. Uranium-236 as an indicator of fuel-cycle uranium in ground water

    International Nuclear Information System (INIS)

    Environmental monitoring on and around the Hanford Site includes regular sampling of onsite monitoring wells and offsite farm wells. Uranium has been identified in the ground water onsite and also in water from farm wells located on the east side of the Columbia River, across from the Hanford Site. Information on the hydrology of the area indicates that the source of the offsite uranium is not the Hanford Site. This study evaluated the isotopic composition of the uranium in water from the various wells to differentiate the onsite uranium contamination from natural uranium offsite. 5 refs., 2 figs., 2 tabs

  5. Novovoronezh Unit 5 WWER-1000 reactor transfer to uranium-gadolinium fuel -operating experience of fuel cycles 23 and 24

    International Nuclear Information System (INIS)

    Development of Uranium-Gadolinium Fuel (UGF) cycle and safety justification for the Novovoronezh NPP Unit 5 were performed by Kurchatov Institute, Gidropress, VNIINM, VNIIAES, while preparation of the new FAs for fabrication was performed by NZKhK. The purpose of the project was to support fuel cycle duration of 300 effective days and longer with the average burn-up of about 50 MWday/kgU in spent FAs. The work was coordinated by concerns TVEL and Rosenergoatom. Special UGF FAs were developed for the Novovoronezh Unit 5 with initial enrichment of 3.9% and 4.3%. Core maps with UGF FAs are provided. During pilot operation (fuel loads 23 to 26) installation of UGF FAs was based on full scope of the core make-up. Fuel load 23 operated from 29 August 2005 to 25 July 2006. The fuel cycle duration was 312.5 effective days. The second batch of UGF FAs operated in fuel load 24 from 4 September 2006 to 14 July 2007. 30 UGF FAs were loaded in the core with the average initial enrichment of 4.3%, and 12 UGF FAs - with the average initial enrichment of 3.9%. This fuel cycle lasted for 302 effective days. For both cycles reactor power behavior curve and the maximum relative FA power as well as the curves of calculated and measured boric acid concentrations by chemical analysis during the cycles are shown. Comparison of FA relative power reconstructed by thermocouple (ThC) readings with calculation in maximum-density FA, reconstructed power root mean-square deviation (RMSD) from calculation in self-powered neutron detectors (SPND) locations during 23rd and 24rd fuel charge operation are presented. Changes in reactor plant thermal power and FA maximum relative power during the both fuel campaigns are also given. Based on the presented results authors concluded that: 1) During UGF operation all parameters monitored according to the Novovoronezh NPP Unit 5 technical specification of safe operation were within operational limits; 2) Introduction of UGF FAs did not result in increased

  6. Estimates of future demand for uranium and nuclear fuel cycle services

    International Nuclear Information System (INIS)

    As a review of forecasts made over the last few years amply demonstrates, projections of nuclear power capacity on a country, regional or world basis are subject to uncertainties. It summarizes the evolution of estimates made in the recent past, should provide a sobering reminder of the advisability of relying on ranges rather than on single figures. Although they are derived from a relatively narrow range of assumptions for nuclear power capacity, the alternative estimates of demands for uranium and nuclear-fuel-cycle services differ by about 50%. If plausible variations in breeder penetration, load factors, tails assays and fuel performance were taken into account, a ratio of 2 between maximum and minimum possible demands for the 2000 could easily be approached. Thus, for instance, a 15% (instead of 5%) breeder penetration by the year 2000 would decrease annual natural uranium demand by about 10%, a drop of load factor from 0.7 to 0.6 would drop the demand by another 10%, a decrease in tail assay from 0.25% to 0.2% would drop the demand by 8%. These momentous uncertainties, characteristic of medium- and long-term demand projections, offer a sharp contrast to the inflexibility of short-term requirements. Once a nuclear plant is ordered, the demand for the fuel services required for its core and for its replacement loadings is practically fixed (subject to minor trade-offs) and it can only be delayed in time by accepting exceedingly heavy additional costs. The demand for uranium can be characterized as being uncertain in the future and inelastic in the present. It faces sources of supply which, with the exception of fabrication and conversion facilities, are characterized by long planning times, lengthy prospecting and construction times, and above all by heavy capital investments. This combination offers an almost ideal framework for instability and wild price fluctuations if consumers and suppliers operate independently seeking temporary guidance in their

  7. The ''RB'' reactor uranium fuel enrichment verification by gamma-ray spectroscopy

    International Nuclear Information System (INIS)

    Gamma spectrometry analysis of natural and 2% enriched uranium metal fuel at the RB reactor was performed by germanium gamma spectrometer applying developed computer code ANA. Different samples of the RB reactor uranium fuel, placed at various distances from the Ge detector, were used during measurements. Gamma-ray self-absorption in the fuel material and the geometrical corrections were included in the calculation performed by computer code EFI based on a Monte Carlo method. Evaluated experimental data were used to determine branching ratio for the 1001 keV gamma line of 234mPa which is in equilibrium with 238U. Obtained results were in good agreement with the results of other authors. Applied gamma spectrometry method is used for examination of the fresh fuel composition and validation of isotopic enrichment of the 2% enriched uranium fuel at the RB reactor. (author)

  8. Economy of uranium resources in a three-component reactor fleet with mixed thorium/uranium fuel cycles

    International Nuclear Information System (INIS)

    The potential for minimizing uranium consumption by using a reactor fleet with three different components and mixed thorium/uranium cycles has been investigated with a view to making nuclear power a more sustainable and cleaner means of generating energy. Mass flows of fissile material have been calculated from burnup simulations at the core-equivalent assembly level for each of the three components of the proposed reactor fleet: plutonium extracted from the spent fuel of a standard pressurised water reactor (first component) is converted to 233U in an advanced boiling water reactor (second component) to feed a deficit of multi-recycled 233U needed for the Th/233U fuel of the light/heavy water reactor (third component) which has a high breeding ratio. Although the proposed fleet cannot breed its own fuel, we show that it offers the possibility for substantial economy of uranium resources without the need to resort to innovative (and costly) reactor designs. A very high fleet breeding ratio is achieved by using only currently existing water-based reactor technology and we show that such three-component systems will become economically competitive if the uranium price becomes sufficiently high (> 300 $/kg). Another major advantage of such systems is a corresponding substantial decrease in production of minor actinide waste. (authors)

  9. Plant-scale anodic dissolution of unirradiated N-Reactor fuel

    International Nuclear Information System (INIS)

    Anodic dissolution tests were made with unirradiated N-Reactor fuel to determine the fuel segment length, diameter, and shape required for high throughput electrorefiner treatment for ultimate disposal in a geologic repository. Based on these tests, a conceptual design was produced of an electrorefiner for a full-scale plant to process N-Reactor spent fuel. In this design, the diameter of an electrode assembly is about 0.6 m (25 in.). Eight of these assemblies in an electrorefiner would accommodate a 1.333-metric-ton batch of N-Reactor fuel. Electrorefining would proceed at a rate of 40 kg uranium per hour

  10. Micromechanical approach of behavior of uranium dioxide nuclear fuel

    International Nuclear Information System (INIS)

    Uranium dioxide (UO2) is the reference fuel for pressurized water nuclear reactors. Our study deals with understanding and modeling of mechanical behavior at the microstructure scale at low temperatures (brittle fracture) and high temperature (viscoplastic strain). We have first studied the geometrical properties of polycrystals at large and of UO2 polycrystal more specifically. As of now, knowledge of this behavior in the brittle fracture range is limited. Consequently, we developed an experimental method which allows better understanding of brittle fracture phenomenon at grain scale. We show that fracture is fully intra-granular and {100} planes seem to be the most preferential cleavage planes. Experimental results are directly used to deduce constitutive equations of intra-granular brittle fracture at crystal scale. This behavior is then used in 3D polycrystal simulation of brittle fracture. The full field calculation gives access to the initiation of fracture and propagation of the crack through the grains. Finally, we developed a mechanical behavior model of UO2 in the viscoplastic range. We first present constitutive equations at macroscopic scale which accounts for an ageing process caused by migration of defects towards dislocations. Secondly, we have developed a crystal plasticity model which was fitted to UO2. This model includes the rotation of the crystal lattice. We present examples of polycrystalline simulations. (author)

  11. Radiation shielding calculation for the MOX fuel fabrication plant Melox

    International Nuclear Information System (INIS)

    Radiation shielding calculation is an important engineering work in the design of the MOX fuel fabrication plant MELOX. Due to the recycle of plutonium and uranium from UO2 spent fuel reprocessing and the large capacity of production (120t HM/yr.), the shielding design requires more attention in this LWR fuel plant. In MELOX, besides several temporary storage facilities of massive fissile material, about one thousand radioactive sources with different geometries, forms, densities, quantities and Pu concentrations, are distributed through different workshops from the PuO2 powder reception unit to the fuel assembly packing room. These sources, with or without close shield, stay temporarily in different locations, containers and glove boxes. In order to optimize the dimensions, the material and the cost of shield as well as to limit the calculation work in a reasonable engineer-hours, a calculation scheme for shielding design of MELOX is developed. This calculation scheme has been proved to be useful in consideration of the feedback from the evolutionary design and construction. The validated shielding calculations give a predictive but reliable radiation doses information. (authors). 2 figs., 10 refs

  12. Fuel Gas Demonstration Plant Program. Volume I. Demonstration plant

    Energy Technology Data Exchange (ETDEWEB)

    1979-01-01

    The objective of this project is for Babcock Contractors Inc. (BCI) to provide process designs, and gasifier retort design for a fuel gas demonstration plant for Erie Mining Company at Hoyt Lake, Minnesota. The fuel gas produced will be used to supplement natural gas and fuel oil for iron ore pellet induration. The fuel gas demonstration plant will consist of five stirred, two-stage fixed-bed gasifier retorts capable of handling caking and non-caking coals, and provisions for the installation of a sixth retort. The process and unit design has been based on operation with caking coals; however, the retorts have been designed for easy conversion to handle non-caking coals. The demonstration unit has been designed to provide for expansion to a commercial plant (described in Commercial Plant Package) in an economical manner.

  13. Imitators of plutonium and americium in a mixed uranium- plutonium nitride fuel

    Science.gov (United States)

    Nikitin, S. N.; Shornikov, D. P.; Tarasov, B. A.; Baranov, V. G.; Burlakova, M. A.

    2016-04-01

    Uranium nitride and mix uranium nitride (U-Pu)N is most popular nuclear fuel for Russian Fast Breeder Reactor. The works in hot cells associated with the radiation exposure of personnel and methodological difficulties. To know the main physical-chemical properties of uranium-plutonium nitride it necessary research to hot cells. In this paper, based on an assessment of physicochemical and thermodynamic properties of selected simulators Pu and Am. Analogues of Pu is are Ce and Y, and analogues Am - Dy. The technique of obtaining a model nitride fuel based on lanthanides nitrides and UN. Hydrogenation-dehydrogenation- nitration method of derived powders nitrides uranium, cerium, yttrium and dysprosium, held their mixing, pressing and sintering, the samples obtained model nitride fuel with plutonium and americium imitation. According to the results of structural studies have shown that all the samples are solid solution nitrides rare earth (REE) elements in UN.

  14. Quantitative determination of uranium distribution homogeneity in MTR fuel type plates

    International Nuclear Information System (INIS)

    IPEN/CNEN-SP produces the fuel to supply its nuclear research reactor IEA-R1. The fuel is assembled with fuel plates containing an U3Si2-Al composite meat. A good homogeneity in the uranium distribution inside the fuel plate meat is important from the standpoint of irradiation performance. Considering the lower power of reactor IEA-R1, the uranium distribution in the fuel plate has been evaluated only by visual inspection of radiographs. However, with the possibility of IPEN to manufacture the fuel for the new Brazilian Multipurpose Reactor (RMB), with higher power, it urges to develop a methodology to determine quantitatively the uranium distribution into the fuel. This paper presents a methodology based on X-ray attenuation, in order to quantify the uranium concentration distribution in the meat of the fuel plate by using optical densities in radiographs and comparison with standards. The results demonstrated the inapplicability of the method, considering the current specification for the fuel plates due to the high intrinsic error to the method. However, the study of the errors involved in the methodology, seeking to increase their accuracy and precision, can enable the application of the method to qualify the final product. (author)

  15. Impact of arbuscular mycorrhizal fungi on uranium accumulation by plants.

    Science.gov (United States)

    de Boulois, H Dupré; Joner, E J; Leyval, C; Jakobsen, I; Chen, B D; Roos, P; Thiry, Y; Rufyikiri, G; Delvaux, B; Declerck, S

    2008-05-01

    Contamination by uranium (U) occurs principally at U mining and processing sites. Uranium can have tremendous environmental consequences, as it is highly toxic to a broad range of organisms and can be dispersed in both terrestrial and aquatic environments. Remediation strategies of U-contaminated soils have included physical and chemical procedures, which may be beneficial, but are costly and can lead to further environmental damage. Phytoremediation has been proposed as a promising alternative, which relies on the capacity of plants and their associated microorganisms to stabilize or extract contaminants from soils. In this paper, we review the role of a group of plant symbiotic fungi, i.e. arbuscular mycorrhizal fungi, which constitute an essential link between the soil and the roots. These fungi participate in U immobilization in soils and within plant roots and they can reduce root-to-shoot translocation of U. However, there is a need to evaluate these observations in terms of their importance for phytostabilization strategies. PMID:18069098

  16. Biogeochemistry of uranium in plants associated to phosphatic rocks in the coastal region of Syria

    International Nuclear Information System (INIS)

    Investigation studies in general, demonstrate that background levels of U in plant ash are less than 2 ppm and plant materials which contain more in excess of this amount are indicative either of local uranium mineralization, or the presence of high background levels of uranium in the substrate. Uranium concentrations in different plant parts grown on decomposite phosphate rocks in the mountain coast region of Syria was investigated. Mean uranium concentrations in the soil ranged between 0.44 - 3.91 ppm in the reference area and 22 - 92 ppm in the area of outcrop in phosphate rocks. The results showed that low-order plant forms (Fuaria, Lycopodium, and Pteridium) readily accumulate uranium, whereas high-order forms accumulate uranium in certain parts only. The greatest amount of uranium in flowering parts is concentrated in the plant roots, followed by leaves, twigs and fruits. In addition, results showed that there is a good correlation between uranium in soil and uranium in plant roots. the study demonstrate that Galium Canum could be considered as a good uranium indicator plant for two reason: It was distributed on decomposite phosphate rocks only, and the high concentration of uranium in aerial part similar to the concentration in soil (89.9 ppm). Lagurus Ovatus may be considered as uranium indicator plant, because it was highly dense on the outcrop phosphate rocks, and has a high uranium concentration in its roots (up to 93 ppm) and aerial parts (up to 33 ppm) compared to concentrations in roots and aerial parts in the reference area (10.2 and 0.37 ppm) respectively. (Author)

  17. Uranium dioxide caramel fuel. An alternative fuel cycle for research and test reactors

    International Nuclear Information System (INIS)

    The work performed in France on Caramel fuels for research reactors reflects the reality of a program based on non proliferation criteria, as they have already appeared several years ago. This work actually includes the following different aspects: - identification of the non proliferation criterion defining this action; - determination of the economical and technical goals to be reached; - realization of research and development studies finalized in a full scale demonstration; - transposition to an industrial and commercial level. The Caramel fuel goals have been defined by comparison with existing reactors: to keep the same performance level in the same safety and reliability conditions, without substantial increase in the fuel cycle cost. Taking into account the wide range of the reactors in operation, these goals will be reached, totally or partially, by assemblies with various geometries. The Caramel fuels utilize slightly enriched uranium, because of the high density of the uranium dioxide 10.25 g/cm.3 The reactivity control capacity of the core is consistent with the behaviour under irradiation so as to keep the operation cycle lengths with the same values as the present ones; the average burn-up being limited to about 30,000 MWd/t, the enrichment is maintained lower than 10%. A study of the Caramel behaviour under irradiation has been undertaken. It started with individual Caramel, and followed successfully with fuel assemblies, irradiated in the reactor Osiris within a significant environment: maximum specific power higher than 3000 W/cm3, and maximum burn up about 30 000 MWd/t. Safety experiments have led to creation of deliberate defects such as clad failure in order to test the irradiation behaviour to study its evolution. The results are positive, the kinetics being rather slow. The full change of a fuel cycle connected with nonproliferation goals appears to be a very wide program with political, technical and industrial implications. The development

  18. Uranium in the Nuclear Fuel Cycle: Creation of Plutonium (Invited)

    Science.gov (United States)

    Ewing, R. C.

    2009-12-01

    One of the important properties of uranium is that it can be used to “breed” higher actinides, particularly plutonium. During the past sixty years, more than 1,800 metric tonnes of Pu, and substantial quantities of the “minor” actinides, such as Np, Am and Cm, have been generated in nuclear reactors - a permanent record of nuclear power. Some of these transuranium elements can be a source of energy in fission reactions (e.g., 239Pu), a source of fissile material for nuclear weapons (e.g., 239Pu and 237Np), and of environmental concern because of their long-half lives and radiotoxicity (e.g., 239Pu and 237Np). In fact, the new strategies of the Advance Fuel Cycle Initiative (AFCI) are, in part, motivated by an effort to mitigate some of the challenges of the disposal of these long-lived actinides. There are two basic strategies for the disposition of these heavy elements: 1.) to “burn” or transmute the actinides using nuclear reactors or accelerators; 2.) to “sequester” the actinides in chemically durable, radiation-resistant materials that are suitable for geologic disposal. There has been substantial interest in the use of actinide-bearing minerals, such as zircon or isometric pyrochlore, A2B2O7 (A= rare earths; B = Ti, Zr, Sn, Hf), for the immobilization of actinides, particularly plutonium, both as inert matrix fuels and nuclear waste forms. Systematic studies of rare-earth pyrochlores have led to the discovery that certain compositions (B = Zr, Hf) are stable to very high doses of alpha-decay event damage1. The radiation stability of these compositions is closely related to the structural distortions that can be accommodated for specific pyrochlore compositions and the electronic structure of the B-site cation. Recent developments in the understanding of the properties of heavy element solids have opened up new possibilities for the design of advanced nuclear fuels and waste forms.

  19. Proposal to recirculate glove box and fabrication area air in a plutonium fuel fabrication plant

    International Nuclear Information System (INIS)

    Recirculating glove box and fabrication area ventilation systems are proposed for a 40 Te/yr mixed plutonium--uranium oxide fuel fabrication plant. The ventilation design criteria are outlined, features of the fabricating plant relating to the ventilation system are shown and the recirculating systems are described. A method of operating and recirculating systems during unusual situations, energy conservation and system advantages are discussed. (U.S.)

  20. Thermal-hydraulic calculations for KUHFR with reduced enrichment uranium fuel

    International Nuclear Information System (INIS)

    This report provides the preliminary results of the thermal-hydraulic calculations to study the safety aspects in fueling the KUHFR with reduced enrichment uranium. The calculations were based on what was outlined in the Safety Analysis Report for the KUHFR and the guidebook for research reactor core conversion, IAEA-TECDOC-233, published by the International Atomic Energy Agency. No significant differences in the thermal-hydraulic operating conditions have been found between HEU and MEU fuels. However, in LEU cases, the combination of three factors - larger power peaking with LEU fuel, smaller thermal conductivity of U3O8-Al fuel with high uranium densities, and thicker fuel meat - resulted in higher maximum fuel and surface temperatures with the LEU oxide fuel. (author)

  1. The effect of the uranium loading on the U3Si2-Al fuel plates

    International Nuclear Information System (INIS)

    The experiment of producing U3Si2-Al fuel plates with uranium loading 3.60; 4.80; and 5.20 gU/cm3 had been done. Each loading was made in two fuel plates, following the picture and frame technique. Test of the fuel plates comprising destructive and non destructive test showed that there ware no blisters, white spots and those plates had good dimension of fuel zone. Higher uranium loadings resulted higher porosity. All the tests showed that fuel plates with the loading 4.80 gU/cm3 no defect while the loading of 5.2 gU/cm3 there was dogboning in the end of fuel plate that gave the cladding thickness less then minimum (0.25 mm) according to the RSG-GAS fuel plate specification. (author)

  2. Spent fuel disposal impact on plant decommissioning

    International Nuclear Information System (INIS)

    Regardless of the decommissioning option selected (DECON, SAFSTOR, or ENTOMB), a 10 CFR 50 license cannot be terminated until the spent fuel is either removed from the site or stored in a separately 10 CFR 72 licensed Independent Spent Fuel Storage Installation (ISFSI). Humboldt Bay is an example of a plant which has selected the SAFSTOR option. Its spent fuel is currently in wet storage in the plant's spent fuel pool. When it completes its dormant period and proceeds with dismantlement, it will have to dispose of its fuel or license an ISFSI. Shoreham is an example of a plant which has selected the DECON option. Fuel disposal is currently critical path for license termination. In the event an ISFSI is proposed to resolve the spent fuel removal issue, whether wet or dry, utilities need to properly determine the installation, maintenance, and decommissioning costs for such a facility. In considering alternatives for spent fuel removal, it is important for a utility to properly account for ISFSI decommissioning costs. A brief discussion is presented on one method for estimating ISFSI decommissioning costs

  3. Production of nuclear ceramic fuel for nuclear power plants at 'Ulba metallurgical plant' OSC

    International Nuclear Information System (INIS)

    The paper describes the flow-sheet of production of uranium dioxide powders and nuclear ceramic fuel pellets of them existing at the facility. 'UMP' OSC applies ADU extraction process of UO2 powders production. An indisputable success of the process is the possibility of use of the wide range of raw materials. Uranium hexafluoride, uranium oxides, uranium metal, uranium tetrafluoride, uranyl salts, uranium ore concentrates, all possible types of uranium-containing materials the processing of which by routine methods is difficult (ashes, scraps, etc.) are used as the raw materials. In addition, a reprocessed nuclear fuel can be used for fuel production. The quality of uranium dioxide powder produced does not depend on the type of uranium raw material used. High selectivity of extraction refining makes possible to obtain material with rather low impurities content that meets practically all specifications for uranium dioxide known to us. Ceramic and process features of uranium dioxide powders, namely, specific surface, bulk density, grain size and sinterability make possible to produce nuclear ceramic fuel with specified features. Quality of uranium dioxide powders produced by 'UMP' OSC was highly rated by General Electric company that is one of the leading companies from fuel manufactures in the USA market . It has certified 'UMP' OSC as its supplier. Currently, our company makes great efforts on establishing production of uranium dioxide powders with natural isotopes content for production of fuel for CANDU reactors. Trial lots of such powders are under tests at some companies manufacturing fuel for this type reactors in Canada, USA and Corea

  4. Experience of MOX-fuel operation in the Gundremmingen BWR plant: Nuclear characteristics and in-core fuel management

    International Nuclear Information System (INIS)

    After 4 years of good experience with MOX-fuel operation in the BWR plants Gundremmingen units B and C the number of inserted MOX-FAs will be increased in the future continuously. Until now all MOX-FAs are in good condition. Furthermore calculations and measurements concerning zero power tests and tip measurements are in good agreement as expected: all results lead to the conclusion that MOX-FAs can be calculated with the same precision as uranium-FAs. (author)

  5. Human health impacts avoided by blending highly enriched uranium to low-enriched uranium for commercial nuclear fuel

    International Nuclear Information System (INIS)

    The end of the Cold War and subsequent Strategic Arms Reduction Treaties have resulted in surplus stockpiles of weapons-usable fissile materials in the United States. If not managed properly, these excess stockpiles could pose a danger to national and international security with potential for environmental, safety, and health consequences. The United States has declared 200 tonnes of fissile materials surplus, of which 165 tonnes is highly enriched uranium (HEU). Uranium with 235U enrichments of 20% or greater is considered HEU. The U.S. Department of Energy proposes to blend the surplus HEU to low-enriched uranium (LEU) to eliminate the risk of diversion for nuclear proliferation purposes and, where practical, to reuse the resulting LEU in ways that recover its commercial value. This paper presents the human health risk assessment results for each proposed blending alternative and compares the health impact to that of the commercial nuclear fuel cycle

  6. Fuel burnup extension effect on the fuel utilization and economical impact for a typical PWR plant

    International Nuclear Information System (INIS)

    Currently in Japan, fuel assembly average burn-up is limited to 48GWd/t and is going to be extended to 55GWd/t in these years. Moreover, R and D programs for further extension are under operation. Simultaneous extension of fuel burn-up limitation and cycle length reduces the number of fuel required to produce a given amount of energy reducing the radioactive waste generation, the occupational radiation exposure and the electricity generation cost. In this paper, the effect of fuel burn-up and operation cycle length extension is estimated from the view point of electricity generation cost and amount of discharged fuel assemblies, and the desirable burn-up extension in the future is studied. The present 5wt% uranium-235 enrichment restriction for commercial reactors divides the burn-up extension implementation in two steps. The fuel burn-up achievable with the present 5wt% enrichment limitation and without it is analyzed. A standard 3 loop PWR plant loading 17x17 fuel assemblies has been chosen for the feasibility study of operation cycle longer than 15 months and up to 24 months under extended fuel burn-up limitation. With the 5wt% enrichment limitation, the maximum assembly average burn-up is between 60GWd/t and 70GWd/t. Three batches reload fuel strategy and 18 months operation cycle allow the electricity generation cost reduction in about 4% and the number of fuel assemblies discharged per year is reduced in approximately 15% compared with the current 48GWd/t fuel. Relaxing the enrichment limitation, for the 24 months operation cycle with 3 batches reload fuel strategy, the maximum assembly average burn-up become 80GWd/t. The electricity generation cost reduction is about 8% and the number of fuel assemblies discharged per year is reduced in approximately 35% compared with the current condition. This study shows the contribution of simultaneous extension of fuel burn-up limitation and operation cycle length to reduce the electricity generation cost and the number

  7. High-Uranium-Loaded U3O8-Al fuel element development program. Part 1

    International Nuclear Information System (INIS)

    The High-Uranium-Loaded U3O8-Al Fuel Element Development Program supports Argonne National Laboratory efforts to develop high-uranium-density research and test reactor fuel to accommodate use of low-uranium enrichment. The goal is to fuel most research and test reactors with uranium of less than 20% enrichment for the purpose of lowering the potential for diversion of highly-enriched material for nonpeaceful usages. The specific objective of the program is to develop the technological and engineering data base for U3O8-Al plate-type fuel elements of maximal uranium content to the point of vendor qualification for full scale fabrication on a production basis. A program and management plan that details the organization, supporting objectives, schedule, and budget is in place and preparation for fuel and irradiation studies is under way. The current programming envisions a program of about four years duration for an estimated cost of about two million dollars. During the decades of the fifties and sixties, developments at Oak Ridge National Laboratory led to the use of U3O8-Al plate-type fuel elements in the High Flux Isotope Reactor, Oak Ridge Research Reactor, Puerto Rico Nuclear Center Reactor, and the High Flux Beam Reactor. Most of the developmental information however applies only up to a uranium concentration of about 55 wt % (about 35 vol % U3O8). The technical issues that must be addressed to further increase the uranium loading beyond 55 wt % U involve plate fabrication phenomena of voids and dogboning, fuel behavior under long irradiation, and potential for the thermite reaction between U3O8 and aluminum

  8. 2010 Status of Uranium Conversion Plant Decommissioning Project

    International Nuclear Information System (INIS)

    The Uranium Conversion Plant (UCP) was used to manufacture 100 tons of UO2. This paper introduced briefly decommissioning activities in the first half year of 2010. powder for the Wolsong-1 CANDU reactor. The conversion plant has been shut down and minimally maintained for the prevention of a contamination by a deterioration of the equipment. The conversion plant has building area of 2916 m2 and two main conversion processes. ADU (Ammonium Di-Uranate) and AUC (Ammonium Uranyl Carbonate) process are installed in the backside and the front side of the building, respectively. Conversion plant has two lagoons, which is to store all wastes generated from the plant operation. Sludge wastes stored 150m3 and 100m3 in Lagoon 1 and 2, respectively. Main compounds of sludge are ammonium nitrate, sodium nitrate, calcium nitrate, and calcium carbonate. In 2000, the decommissioning of the plant was finally decided upon and a decommissioning program was launched to complete by 2010. In the middle of 2004, decommissioning program obtained the approval of regulatory body and decommissioning activities started. The scope of the project includes the removal of all equipment and the release of the building for re-use. The project is scheduled to be completed at the end of 2010 with a total budget of 10.9 billion This paper introduced briefly decommissioning activities in the first half year of 2010

  9. Seismic vulnerability study of the nation's uranium enrichment plants

    International Nuclear Information System (INIS)

    The US Department of Energy's uranium enrichment production is accomplished with three gaseous diffusion plants located at Oak Ridge, Tennessee; Paducah, Kentucky; and Portsmouth, Ohio. The plants were built in the 1940's and 1950's with no seismic design requirements and are located in three different seismic zones. Paducah is in the New Madrid seismic zone (UBC-Zone 3), Oak Ridge is in the Southern Appalachian seismic zone (UBC-Zone 2) and Portsmouth is near the Anna, Ohio seismic zone (UBC-Zone 1). This paper discusses the approach that was used to determine the seismic vulnerability of each of the plants in response to safety and operability analysis studies. Using state-of-the-art seismic evaluation methods, the study showed that the plants are more resistant to seismic excitation than previously thought. However, the study also showed that small seismic excitations could cause any one of the plants to shut down because of weak links in the process systems. It was determined that for about $6 million each, the Oak Ridge and Paducah plants could be upgraded to provide continuity of operation and operational safety at the evaluation basis earthquake levels. At Portsmouth the upgrade costs were determined to be about $1 million, much less than Paducah or Oak Ridge because of process equipment uniqueness

  10. Study of Tower Reactor Fuel Elements Based on Sintered Uranium Dioxide

    International Nuclear Information System (INIS)

    The paper gives the results of loop tests on a large batch of experimental fuel elements based on sintered uranium dioxide. Generalized data on the operation of fuel elements used in the reactors of the icebreaker ''Lenin'' are also included. (author)

  11. The improvement of technology for high-uranium-density Al-base dispersion fuel plates

    International Nuclear Information System (INIS)

    An improved rolling process was developed for manufacturing Al-base dispersion fuel plates. When the fuel content in the meat increased up to 50 vol%, the non-uniformity of uranium is not more than ± 7.2%, and the minimum cladding thickness is not less than 0.32 mm. (Author)

  12. Safety analysis report of uranium dioxide fuel laboratory, Nuclear Research Centre Inchas, Egypt

    International Nuclear Information System (INIS)

    In the Nuclear Research Center Inchas a uranium dioxide fuel laboratory is planned and built by the AEA Cairo (Atomic Energy Authority). The layout of this fuel lab and the programmatical contents are subject to the bilaterial cooperation between Egypt and the Federal Republic of Germany. In this report the safety analysis as basic items for the approval procedure are started in detail. (orig.)

  13. Development, preparation and characterization of uranium molybdenum alloys for dispersion fuel application

    International Nuclear Information System (INIS)

    Most of the research and test reactors worldwide have undergone core conversion from high enriched uranium base fuel to low enriched uranium base fuel under the Reduced Enrichment for Research and Test Reactor (RERTR) program, which was launched in the late 1970s to reduce the risk of nuclear proliferation. To realize this goal, high density uranium compounds and γ-stabilized uranium alloy powder were identified. In Metallic Fuels Division of BARC, R and D efforts are on to develop these high density uranium base alloys. This paper describes the preparation flow sheet for different compositions of Uranium and molybdenum alloys by an innovative powder processing route with uranium and molybdenum metal powders as starting materials. The same composition of U-Mo alloys were also fabricated by conventional method i.e. ingot metallurgy route. The U-Mo alloys prepared by both the methods were then characterized by XRD for phase analysis. The photomicrographs of alloys with different compositions prepared by powder metallurgy and ingot metallurgy routes are also included in the paper. The paper also covers the comparison of properties of the alloys prepared by powder metallurgy and ingot metallurgy routes

  14. Cost of transporting irradiated fuels and maintenance costs of a chemical treatment plant for irradiated fuels

    International Nuclear Information System (INIS)

    Numerous studies have been made of the cost of a fuel cycle, but many of them are based on a priori studies and are therefore to be treated with reserve. Thus, in the part dealing with the treatment of irradiated fuels, some important factors in the cost have only rarely been given on the basis of practical experience: the cost of transporting the fuels themselves and the plant maintenance costs. Investigations relating to transport costs are generally based on calculations made from somewhat arbitrary data. The studies carried out in France on the transport of irradiated uranium between the EDF reactors at Chinon and the retreatment plant at La Hague of the irradiated uranium from research reactors to foreign retreatment plants, are reported; they show that by a suitable choice of transport containers and details of expedition it has been possible to reduce the costs very considerably. This has been achieved either by combining rail and road transport or by increasing the writ capacities of the transport containers: an example is given of a container for swimming-pool pile elements which can transport a complete pile core at one time, thus substantially reducing the cost. Studies concerning the maintenance costs of retreatment plants are rarer still, although in direct maintenance plants these figures represent an appreciable fraction of the total treatment cost. An attempt has been made, on the basis of operational experience of a plant, to obtain some idea of these costs. Only maintenance proper has been considered, excluding subsidiary operations such as the final decontamination of apparatus, the burial of contaminated material and radioprotection operations Maintenance has been divided into three sections: mechanical maintenance, maintenance of electrical equipment and maintenance of control and adjustment apparatus. In each of these sections the distinction has been made between manpower and the material side. In order to allow comparisons to be made with

  15. SUB-LEU-METAL-THERM-001 SUBCRITICAL MEASUREMENTS OF LOW ENRICHED TUBULAR URANIUM METAL FUEL ELEMENTS BEFORE & AFTER IRRADIATION

    Energy Technology Data Exchange (ETDEWEB)

    TOFFER, H.

    2006-07-18

    With the shutdown of the Hanford PUREX (Plutonium-Uranium Extraction Plant) reprocessing plant in the 1970s, adequate storage capacity for spent Hanford N Reactor fuel elements in the K and N Reactor pools became a concern. To maximize space utilization in the pools, accounting for fuel burnup was considered. Fuel that had experienced a neutron environment in a reactor is known as spent, exposed, or irradiated fuel. In contrast fuel that has not yet been placed in a reactor is known as green, unexposed, or unirradiated fuel. Calculations indicated that at typical fuel exposures for N Reactor, the spent-fuel critical mass would be twice the critical mass for green fuel. A decision was reached to test the calculational result with a definitive experiment. If the results proved positive, storage capacity could be increased and N Reactor operation could be prolonged. An experiment to be conducted in the N Reactor spent-fuel storage pool was designed and assembled (References 1 and 2) and the services of the Battelle Northwest Laboratories (BNWL) (now Pacific Northwest National Laboratory [PNNL]) critical mass laboratory were procured for the measurements (Reference 3). The experiments were performed in April 1975 in the Hanford N Reactor fuel storage pool. The fuel elements were MKIA fuel assemblies, comprised of two concentric tubes of low-enriched metallic uranium. Two separate sets of measurements were performed: one with unirradiated fuel and one with irradiated fuel. Both the unirradiated and irradiated fuel, were measured in the same geometry. The spent-fuel MKIA assemblies had an average burnup of 2865 MWd (megawatt days)/t. A constraint was imposed restricting the measurements to a subcritical limit of k{sub eff} = 0.97. Subcritical count rate data was obtained with pulsed-neutron and approach-to-critical measurements. Ten (10) configurations with green fuel and nine (9) configurations with spent fuel are described and evaluated. Of these, three (3) green fuel

  16. Lung Cancer Mortality among Uranium Gaseous Diffusion Plant Workers: A Cohort Study 1952–2004

    OpenAIRE

    LW Figgs

    2013-01-01

    Background: 9%–15% of all lung cancers are attributable to occupational exposures. Reports are disparate regarding elevated lung cancer mortality risk among workers employed at uranium gaseous diffusion plants.Objective: To investigate whether external radiation exposure is associated with lung cancer mortality risk among uranium gaseous diffusion workers.Methods: A cohort of 6820 nuclear industry workers employed from 1952 to 2003 at the Paducah uranium gaseous diffusion plant (PGDP) was ass...

  17. Transition to integrated safeguards at nuclear fuel fabrication plants

    International Nuclear Information System (INIS)

    Full-text: This paper presents the improvements in efficiency and effectiveness of the safeguards approach applied at depleted, natural and low enriched uranium (DNLEU) conversion and fuel fabrication plants (FFPs) following a transition from traditional safeguards (TS) to integrated safeguards1 (IS). The paper will explain the relationship between the Agency drawing a broader conclusion for a State and the implementation of IS, highlight how the transition from TS to IS can be accomplished, and describe a process by which IS can be introduced through a combination of good IAEA-National Authority cooperation and field trials. For demonstration purposes, a comparison of the onsite safeguards activities before and after the introduction of IS at a fuel fabrication plant which manufactures both PWR (Pressurized Water Reactors) fuel assemblies and CANDU fuel bundles will be examined. It is expected that this will provide a good reference baseline on the safeguards improvements made possible by the introduction of IS at fuel fabrication facilities without mixed oxide fuels (MOX). The paper will also show the operational and logistical differences between the TS and IS regimes and highlight some of the advantages to a State for which a broader conclusion has been drawn by the Agency. The paper will offer technical insight to safeguards coverage of nuclear material borrowing scenarios and implementation of other safeguards measures such as the introduction of short notice random inspections (SNRI), use of a secure mailbox system at the FFP, and implementation of random interim inspections (RIIs) at LWRs (Light Water Reactors) and CANDU power reactors. (author)

  18. Uranium isotope separation by gaseous diffusion and plant safety

    International Nuclear Information System (INIS)

    This report constitutes a safety guide for operators of uranium isotope separation plants, and includes both aspects of safety and protection. Taking into account the complexity of safety problems raised at design and during operation of plants which require specialized guides, this report mainly considers both the protection of man, the environment and goods, and the principles of occupational safety. It does not claim to be comprehensive, but intends to state the general principles, the particular points related to the characteristics of the basic materials and processes, and to set forth a number of typical solutions suitable for various human and technical environments. It is based on the French experience gained during the last fifteen years

  19. 226Ra bioavailability of plants at urgeirica uranium mill tailings

    International Nuclear Information System (INIS)

    Large amounts of solid wastes (tailings) resulting from the exploitation and treatment of uranium ore at the Urgeirica mine (north of Portugal) have been accumulated in dams (tailing ponds). To reduce the dispersion of natural radionuclides into the environment some dams were revegetated with eucalyptus (Eucalyptus globolus) and pines (Pinus pinea). Besides, some shrubs (Cytisus s.p.) are growing at some of the dams. The objective of this study is to determine the 226Ra bioavailability from uranium mill tailings through the quantification of the total and available fraction of radium in the solid wastes and to estimate its transfer to the plants growing on the tailing piles. Plants and solid waste samples were randomly collected at dams. Activity concentration of 226Ra in plants (aerial part and roots) and solid wastes were measured by gamma spectrometry. The exchangeable fraction of radium in solid wastes was quantified using one single step extraction with 1 mol dm-3 ammonium acetate (pH=7) or 1 mol dm-3 calcium chloride solutions. The results obtained for the 226Ra uptake by plants show that 226Ra concentration ratios for eucalyptus and pines decrease at low 226Ra concentration in the solid wastes and appear relatively constant at higher radium concentrations. For shrubs, the concentration ratios increase at higher 226Ra solid waste concentrations approaching a saturation value. Percentage values of 16.0±8.3 and 12.9±8.9, for the fraction of radium extracted from the solid wastes, using 1 mol dm-3 ammonium acetate or calcium chloride solutions respectively, were obtained. The 226Ra concentration ratios determined on the basis of exchangeable radium are one order of magnitude higher than those based on total radium. It can be concluded that, within the standard error values, more consistent 226Ra concentration ratios were obtained when calculated on the basis of available radium than when total radium was considered, for all the dams. (author)

  20. Transport mechanisms of uranium released to the coolant from fuel defects

    International Nuclear Information System (INIS)

    Fuel performance at domestic CANDU-600s, Point Lepreau and Gentilly, has been very good, with only a small number of fuel defects releasing uranium to the coolant. The in-core monitoring on these early fuel defects using the delayed neutron system, provides some insight into uranium transport mechanisms and how they influence signal trends. Better understanding of these mechanisms, will assist the station operator in responding to trend changes and will ultimately provide guidance in assigning removal priorities should several fuel defects occur simultaneously. The average delayed neutron signal of all channels is the key parameter for monitoring fuel performance in-core, and should be regarded as an early warning indicator of fuel performance problems

  1. Establishing Specifications for Low Enriched Uranium Fuel Operations Conducted Outside the High Flux Isotope Reactor Site

    Energy Technology Data Exchange (ETDEWEB)

    Pinkston, Daniel [ORNL; Primm, Trent [ORNL; Renfro, David G [ORNL; Sease, John D [ORNL

    2010-10-01

    The National Nuclear Security Administration (NNSA) has funded staff at Oak Ridge National Laboratory (ORNL) to study the conversion of the High Flux Isotope Reactor (HFIR) from the current, high enriched uranium fuel to low enriched uranium fuel. The LEU fuel form is a metal alloy that has never been used in HFIR or any HFIR-like reactor. This report provides documentation of a process for the creation of a fuel specification that will meet all applicable regulations and guidelines to which UT-Battelle, LLC (UTB) the operating contractor for ORNL - must adhere. This process will allow UTB to purchase LEU fuel for HFIR and be assured of the quality of the fuel being procured.

  2. Implementation of the MARSSIM to Evaluate the Final Status After Decommissioning Uranium Conversion Plant

    International Nuclear Information System (INIS)

    The decommissioning project of Uranium Conversion Plant was launched in 2001 and completed at the first half of 2011. The final stage of decommissioning process was the release of a site and building from regulatory control. KAERI carried out a final status survey based on the guidance provided in the MARSSIM (Multi-Agency Radiation Survey and Site Investigation Manual). The Uranium Conversion Plant was used to manufacture UO2 powder for CANDU fuel, the plant was contaminated only natural uranium. In this study, plans for the final status survey and release criteria for a site were established by applying the MARSSIM procedures. The survey design for the final status survey of the UCP site and buildings was carried out based on the statistical test of the results form scoping and characterization survey. The site and buildings were classified based on the potential contamination by using measured and calculated results. The results of the final status survey were satisfied the release criteria based on the measured data from a site and building. The summarized final status survey results are given in Figure 1 shows detail residual contamination level in Class 1. The results of the final status survey are sufficiently lower than the release criteria. The MARSSIM procedures were proved to be flexible, scientifically rigorous and cost effective for final status survey of decommissioning site and building. For the effective plan for the final status survey of the UCP site and its implementation, KAERI and its the regulation body are continuously discussing way to ensure the validation of the final status survey.

  3. MEU/Th fuel cycle optimization for the Lead Plant

    International Nuclear Information System (INIS)

    The reference equilibrium cycle fuel composition for the Lead Plant was specified previously by a C/Th ratio of 850 and a fuel rod diameter of 1.17 cm, which is optimal for non-recycle operation and close to optimal for recycle of bred U-233. Subsequent work has emphasized the importance of full recycle of all discharged uranium to maintain the competitive advantage of the MEU/Th cycle. Cycles with full recycle optimize at higher thorium loadings and larger rod diameters. This is an additional benefit for core design and reduces fabrication problems. New optimization studies based on full recycle lead to an equilibrium cycle composition characterized by a C/Th ratio of 600 and a rod diameter of 1.35 cm. The average packing fraction of fuel particles in the rod is 0.43. The C/Th ratio for the initial core is 350, which can also be accommodated with the 1.35 cm rod diameter. Mass flow data for 30 year operation and fuel cycle cost data have been obtained for this cycle and for several other thorium loadings

  4. Transport of high enriched uranium fresh fuel from Yugoslavia to the Russian federation

    Directory of Open Access Journals (Sweden)

    Pešić Milan P.

    2002-01-01

    Full Text Available This paper presents the relevant data related to the recent shipment (August 2002 of fresh highly enriched uranium fuel elements from Yugoslavia back to the Russian Federation for uranium down blending. In this way, Yugoslavia gave its contribution to the Reduced Enrichment for Research and Test Reactors (RERTR Program and to the world's joint efforts to prevent possible terrorist actions against nuclear material potentially usable for the production of nuclear weapons.

  5. Neutronics Studies of Uranium-bearing Fully Ceramic Micro-encapsulated Fuel for PWRs

    International Nuclear Information System (INIS)

    Our study evaluated the neutronics and some of the fuel cycle characteristics of using uranium-based fully ceramic microencapsulated (FCM) fuel in a pressurized water reactor (PWR). Specific PWR lattice designs with FCM fuel have been developed that are expected to achieve higher specific burnup levels in the fuel while also increasing the tolerance to reactor accidents. The SCALE software system was the primary analysis tool used to model the lattice designs. A parametric study was performed by varying tristructural isotropic particle design features (e.g., kernel diameter, coating layer thicknesses, and packing fraction) to understand the impact on reactivity and resulting operating cycle length. Moreover, to match the lifetime of an 18-month PWR cycle, the FCM particle fuel design required roughly 10% additional fissile material at beginning of life compared with that of a standard uranium dioxide (UO2) rod. Uranium mononitride proved to be a favorable fuel for the fuel kernel due to its higher heavy metal loading density compared with UO2. The FCM fuel designs evaluated maintain acceptable neutronics design features for fuel lifetime, lattice peaking factors, and nonproliferation figure of merit

  6. Thorium - denatured uranium fuel cycles in PHWR-pressure tube type using low enriched uranium as annual externally supplied fissile material

    International Nuclear Information System (INIS)

    The use of denatured uranium as the initial fissile inventory of the thorium-uranium cycles could be straight-forward. The use of denatured uranium as annual externally supplied fissile material could be not applicable in the case of homogenous HWR fuel bundle concept if it is intended to avoid the shift from the Th/U cycle to U/Pu cycle or the reenrichment of the recycled uranium containing U-232. The paper presents a heterogenous fuel concept for HWR which permits the use of denatured uranium without the above-mentioned shift. According to this concept the annual externally supplied fissile material is introduced in distinct fuel rods separable at the front end of the reprocessing or as distinct fuel bundles. In these cases the normal reenrichment could be applied, this part of the fuel being free of U-233, and therefore free of U-232. The resource utilization penalties in addition to those introduced by the denaturing of the initial core were evaluated. At 3% enrichment these penalties rise with about 40% the annual natural uranium requirements. It is concluded that for these Th/U cycles in HWR, it is possible to avoid the presence of highly enriched uranium at the fuel fabrication step

  7. CRITICALITY CONTROL DURING THE DISMANTLING OF A URANIUM CONVERSION PLANT

    International Nuclear Information System (INIS)

    Within the Commissariat a l'Energie Atomique, in the Cadarache Research Center in southern France, the production at the Enriched Uranium Treatment Workshops started in 1965 and ended in 1995. The dismantling is in progress and will last until 2006. The decommissioning is planned in 2007. Since the authorized enrichment in 235U was 10% in some parts of the plant, and unlimited in others, the equipment and procedures were designed for criticality control during the operating period. Despite the best previous removing of the uranium in the inner parts of the equipment, evaluation of the mass of remaining fissile material by in site gamma spectrometry measurement shows that the safety of the ''clean up'' operations requires specific criticality control procedures, this mass being higher than the safe mass. The chosen method is therefore based on the mapping of fissile material in the contaminated parts of the equipment and on the respect of particular rules set for meeting the criticality control standards through mass control. The process equipment is partitioned in separated campaign, and for each campaign the equipment dismantling is conducted with a precise traceability of the pieces, from the equipment to the drum of waste, and the best final evaluation of the mass of fissile material in the drum. The first results show that the mass of uranium found in the dismantled equipment is less than the previous evaluation, and they enable us to confirm that the criticality was safely controlled during the operations. The mass of fissile material remaining in the equipment can be then carefully calculated, when it is lower than the minimal critical mass, and on the basis of a safety analysis, we will be free of any constraints regarding criticality control, this allowing to make procedures easier, and to speed up the operations

  8. Safety aspects of fuel reprocessing plants

    International Nuclear Information System (INIS)

    The reprocessing of irradiated reactor fuels is an important activity in the total fuel cycle. The reprocessing step has special radiological, technological and operational problems associated with it since it renders the high integrity solid reactor fuel into highly dispersible forms resulting in problems of containment and confinement of toxic materials like plutonium and fission products. This operation also makes plutonium available in a concentrated and pure form. The design and operation of a reprocessing plant must result in very low environmental releases under these conditions. The irradiated fuel must be reprocessed periodically for one or more of the following reasons : i) For separation of accumulated fission product poisons which adversely affects the reactivity in the reactor; ii) For recovery of unspent fissile and fertile materials; iii) For recovery of the reactor produced plutonium and iv) Due to possible physical changes in the fuel rendering reactor operation difficult

  9. Equipment for manufacture of uranium-plutonium mixed carbide fuel pins

    International Nuclear Information System (INIS)

    The equipment for manufacturing fuel pins, which is neccesary for irradiation tests of uranium-plutonium mixed carbide fuels, has been provided. This equipment is composed of a centerless grinder, an apparatus for loading fuel pellets and endplugs into cladding tubes, a TIG-welder, an apparatus for decontamination of welded fuel pins, and so on. Most of them are installed into gloveboxes, in order to prevent the workers from plutonium contamination. The maximum size of the pins manufactured by the equipment is 15 mm in radius and 600 mm in length. In this report, design, construction, and ability of the equipment for the manufacture of carbide fuel pins are described. (author)

  10. SUB-LEU-METAL-THERM-001 SUBCRITICAL MEASUREMENTS OF LOW-ENRICHED TUBULAR URANIUM METAL FUEL ELEMENTS BEFORE and AFTER IRRADIATION

    International Nuclear Information System (INIS)

    With the shutdown of the Hanford PUREX (Plutonium-Uranium Extraction Plant) reprocessing plant in the 1970s, adequate storage capacity for spent Hanford N Reactor fuel elements in the K and N Reactor pools became a concern. To maximize space utilization in the pools, accounting for fuel burnup was considered. Fuel that had experienced a neutron environment in a reactor is known as spent, exposed, or irradiated fuel. In contrast fuel that has not yet been placed in a reactor is known as green, unexposed, or unirradiated fuel. Calculations indicated that at typical fuel exposures for N Reactor, the spent-fuel critical mass would be twice the critical mass for green fuel. A decision was reached to test the calculational result with a definitive experiment. If the results proved positive, storage capacity could be increased and N Reactor operation could be prolonged. An experiment to be conducted in the N Reactor spent-fuel storage pool was designed and assembled (References 1 and 2) and the services of the Battelle Northwest Laboratories (BNWL) (now Pacific Northwest National Laboratory [PNNL]) critical mass laboratory were procured for the measurements (Reference 3). The experiments were performed in April 1975 in the Hanford N Reactor fuel storage pool. The fuel elements were MKIA fuel assemblies, comprised of two concentric tubes of low-enriched metallic uranium. Two separate sets of measurements were performed: one with unirradiated fuel and one with irradiated fuel. Both the unirradiated and irradiated fuel, were measured in the same geometry. The spent-fuel MKIA assemblies had an average burnup of 2865 MWd (megawatt days)/t. A constraint was imposed restricting the measurements to a subcritical limit of keff = 0.97. Subcritical count rate data was obtained with pulsed-neutron and approach-to-critical measurements. Ten (10) configurations with green fuel and nine (9) configurations with spent fuel are described and evaluated. Of these, three (3) green fuel and

  11. Soil-to-plant transfer of uranium and its distribution between plant parts in four boreal forest species

    International Nuclear Information System (INIS)

    Uranium (U) can be released to the environment through the entire nuclear fuel cycle. U uptake by plants is an important process for possible adverse effects in ecosystems. The soil-to-plant transfer of natural U and its distribution across plant parts were investigated in May lily (Maianthemum bifolium), narrow buckler fern (Dryopteris carthusiana), rowan (Sorbus aucuparia) and Norway spruce (Picea abies). Concentration ratios (CR) between plant and soil were calculated. The CRs for roots were higher than those for the above-ground parts of the plants. Soil pH was the only soil parameter showing an effect on CRs. No significant differences were noticed between species. The CRs observed were consistent with those reported previously in other forest types. The pooled values of 0.06 for roots and 0.005 for stems/petioles and leaves/needles can be considered as good estimates of CR values to be used in modelling the U uptake in boreal forest species. (orig.)

  12. Biomass-fueled power plants in Finland

    Energy Technology Data Exchange (ETDEWEB)

    Raiko, M. [IVO Power Engineering Ltd., Vantaa (Finland); Hulkkonen, S. [Imatran Voima Oy, Vantaa (Finland)

    1997-07-01

    Combined heat and power production (CHP) from biomass is a commercially viable alternative when district heat or process steam is needed in small towns or in a process industry. The high nominal investment cost of a small power plant that uses local biomass fuels is compensated by the revenues from the heat. The price of the district heat or the steam generated in the CHP-plant can be valued at the same price level as the heat from a mere steam boiler. Also, the price of heat produced by a small-generation-capacity plant is local and higher, whereas electricity has a more general market price. A typical small Finnish CHP-plant consists of a bubbling fluidized bed boiler and a simplified steam turbine cycle generating 4 to 10 MW of electricity and 10 to 30 MW of district heat or process steam. There are about 10 power plants of this type in commercial operation in Finland. As a whole, biomass, which is used in more than 200 plants, provides about 20% of the primary energy consumption in Finland. Roughly half of these produce only heat but the rest are combined heat and power plants. The majority of the plants is in pulp and paper industry applications. Imatran Voima Oy (IVO) is the biggest energy producer in Finland. IVO builds, owns and operates several biomass-fired power plants and carries out active R and D work to further develop the biomass-fueled small power plant. This paper discusses the experiences of the biomass-fueled power plants. (author)

  13. The specific determination of uranium in nuclear fuel

    International Nuclear Information System (INIS)

    This titration method is satisfactory for routine determination of uranium in nitric acid or tributyl phosphate-kerosen solution within the range 2-300 mg and a precision of +-0.1-0.005 percent. The method involves reduction of uranium (VI) to (IV) by ferrous sulphate in concentrated phosphoric acid medium. The excess iron (II) is oxidised with nitric acid using molybdenum catalyst. After addition of sulphuric acid and dilution with water, the iron (II) which is equivalent to uranium (IV) is titrated with dichromate by potentiometric end point using a platinum-reference electrode (Ag/AgCl) pair. IAEA Intercomparison sample SR-40 has been analysed and uranium concentration has been found as 85.898+-0.203 percent. (author)

  14. Biogeochemistry of uranium in the soil-plant and water-plant systems in an old uranium mine.

    Science.gov (United States)

    Favas, Paulo J C; Pratas, João; Mitra, Soumita; Sarkar, Santosh Kumar; Venkatachalam, Perumal

    2016-10-15

    The present study highlights the uranium (U) concentrations in water-soil-plant matrices and the efficiency considering a heterogeneous assemblage of terrestrial and aquatic native plant species to act as the biomonitor and phytoremediator for environmental U-contamination in the Sevilha mine (uraniferous region of Beiras, Central Portugal). A total of 53 plant species belonging to 22 families was collected from 24 study sites along with ambient soil and/or water samples. The concentration of U showed wide range of variations in the ambient medium: 7.5 to 557mgkg(-1) for soil and 0.4 to 113μgL(-1) for water. The maximum potential of U accumulation was recorded in roots of the following terrestrial plants: Juncus squarrosus (450mgkg(-1) DW), Carlina corymbosa (181mgkg(-1) DW) and Juncus bufonius (39.9mgkg(-1) DW), followed by the aquatic macrophytes, namely Callitriche stagnalis (55.6mgkg(-1) DW) Lemna minor (53.0mgkg(-1) DW) and Riccia fluitans (50.6mgkg(-1) DW). Accumulation of U in plant tissues exhibited the following decreasing trend: root>leaves>stem>flowers/fruits and this confirms the unique efficiency of roots in accumulating this radionuclide from host soil/sediment (phytostabilization). Overall, the accumulation pattern in the studied aquatic plants (L. minor, R. fluitans, C. stagnalis and Lythrum portula) dominated over most of the terrestrial counterpart. Among terrestrial plants, the higher mean bioconcentration factor (≈1 in roots/rhizomes of C. corymbosa and J. squarrosus) and translocation factor (31 in Andryala integrifolia) were encountered in the representing families Asteraceae and Juncaceae. Hence, these terrestrial plants can be treated as the promising candidates for the development of the phytostabilization or phytoextraction methodologies based on the accumulation, abundance and biomass production. PMID:27314898

  15. Chemical Decontamination of Metallic Waste from Uranium Conversion Plant Dismantling

    International Nuclear Information System (INIS)

    Korea Atomic Energy Research Institute (KAERI) started a decommissioning program of the uranium conversion plant. Pre-work was carried as follows; installation of the access control facility, installation of a changing room and shower room, designation of an emergency exit way and indicating signs, installation of a radiation management facility, preparation of a storage area for tools and equipments, inspection and load test of crane, distribution and packaging of existing waste, and pre-decontamination of the equipment surface and the interior. First, decommissioning work was performed in kiln room, which will be used for temporary radioactive waste storage room. Kiln room housed hydro fluorination rotary kiln for production of uranium tetra-fluoride. The kiln is about 0.8 m in diameter and 5.5 m long. The total dismantled waste was 6,690 kg, 73 % of which was metallic waste and 27 % the others such as cable, asbestos, concrete, secondary waste, etc. And effluent treatment room and filtration room were dismantled for installation of decontamination equipment and lagoon sludge treatment equipment. There were tanks and square mixer in these rooms. The total dismantled waste was 17,250 kg, 67% of which was metallic waste and 33% the others. These dismantled metallic wastes consist of stainless and carbon steel. In this paper, the stainless steel plate and pipe were decontaminated by the chemical decontamination with ultrasonic

  16. Considerations in the assessment of the consequences of effluents from mixed oxide fuel fabrication plants

    International Nuclear Information System (INIS)

    The purpose of this study was to provide information and identify parameters relevant to assessing the consequences to man and his environment of large scale mixed plutonium-uranium oxide fuel fabrication plants which will be needed in the next 10 to 15 years. The report identifies the pertinent parameters, values, factors and methods which may be used in evaluating the environmental consequences of routine plant operation as well as postulated accidents. This study provides a base for the development of siting criteria and safety analyses for mixed oxide fuel fabrication facilities. (auth)

  17. Development of a computerized nuclear materials control and accounting system for a fuel reprocessing plant

    International Nuclear Information System (INIS)

    A computerized nuclear materials control and accounting system (CNMCAS) for a fuel reprocessing plant is being developed by Allied-General Nuclear Services at the Barnwell Nuclear Fuel Plant. Development work includes on-line demonstration of near real-time measurement, measurement control, accounting, and processing monitoring/process surveillance activities during test process runs using natural uranium. A technique for estimating in-process inventory is also being developed. This paper describes development work performed and planned, plus significant design features required to integrate CNMCAS into an advanced safeguards system

  18. Determination of defective SiC fraction and free uranium fraction in the HTGR fuel compacts

    International Nuclear Information System (INIS)

    Free uranium fraction and defective SiC fraction in fuel compacts and coated fuel particles were determined. For determination of the free uranium in the fuel compacts, the nitric acid leaching after electric disintegration of the compacts was adopted. Defective SiC layer was detected by a burn-leach method where the particles were burnt in air and leached with nitric acid. The free uranium fractions were in order of 10-8 to 10-4. The defective SiC fractions were in order of 10-4 to 10-3. We found a correlation between the value of defective SiC fraction and the times of burn and leaching. (author)

  19. Occupational safety data and casualty rates for the uranium fuel cycle. [Glossaries

    Energy Technology Data Exchange (ETDEWEB)

    O' Donnell, F.R.; Hoy, H.C.

    1981-10-01

    Occupational casualty (injuries, illnesses, fatalities, and lost workdays) and production data are presented and used to calculate occupational casualty incidence rates for technologies that make up the uranium fuel cycle, including: mining, milling, conversion, and enrichment of uranium; fabrication of reactor fuel; transportation of uranium and fuel elements; generation of electric power; and transmission of electric power. Each technology is treated in a separate chapter. All data sources are referenced. All steps used to calculate normalized occupational casualty incidence rates from the data are presented. Rates given include fatalities, serious cases, and lost workdays per 100 man-years worked, per 10/sup 12/ Btu of energy output, and per other appropriate units of output.

  20. Out-of-pile researches of modernized Uranium-Gadolinium fuel for WWER reactors

    International Nuclear Information System (INIS)

    Thermal conductivity and creeping characteristics (creep rate) of modernized uranium-gadolinium fuel as a composite - solid solution granules (U,Gd)O2 dispersed into UO2 matrix have been researched in wide temperature range. The main peculiarity of the fuel which is under study is a thin fine-grained zone surrounding each granule (U,Gd)O2. Presence of this zone blocks intergranular diffusion of components preventing decrease of concentration of solid solution (U,Gd)O2. In addition fine-grained zone is the original buffer relaxing both interphase tensions and tensions that can be resulted from radiation-thermal loads during irradiation of fuel. Higher thermal and physical characteristics of describable composite uranium-gadolinium fuel (creep rate, thermal conductivity, peculiarity of microstructure), in comparison with conventional obtained by dry blending of UO2 and Gd2O3 powders, allow consider this fuel as candidate for using in WWER and PWR reactors. (authors)

  1. The production, characterization, and neutronic performance of boron nitride coated uranium dioxide fuel

    International Nuclear Information System (INIS)

    The fuel pellets produced by sol-gel technique were coated with boron nitride (BN). This was achieved through chemical vapor deposition (CVD) using boron trichloride and ammonia. Mixing and chemical reaction take place at a temperature around 875 K. The coated samples were then sintered at 1600 K. Thermal reactor physics lattice-cell code WIMS-D/4 was used in the neutronic analysis of CANDU fuel bundle to observe the neutronic performance of the coated fuel. Three types of fuel were considered; fuel made of natural uranium, slightly enriched uranium (SEU, enrichment: 0.82 % U-235), and SEU with various BN coatings. The burnup calculations showed that feasible coating thickness is between 1 to 2 μm. (author)

  2. Babcock and Wilcox plate fabrication experience with uranium silicide spherical fuel

    International Nuclear Information System (INIS)

    This report is written to present the fuel fabrication experience of Babcock and Wilcox using atomized spherical uranium silicide powder. The intent is to demonstrate the ability to fabricate fuel plates using spherical powder and to provide useful information proceeding into the next phase of work using this type of fuel. The limited quantity of resources- spherical powder and time, did not allow for much process optimizing in this work scope. However, the information contained within provides optimism for the future of spherical uranium silicide fuel plate fabrication at Babcock and Wilcox.The success of assembling fuel elements with spherical powder will enable Babcock and Wilcox to reduce overall costs to its customers while still maintaining our reputation for providing high quality research and test reactor products. (author)

  3. Irradiated uranium reprocessing, Final report I-VI, IV Deo IV - Separation of uranium, plutonium and fission products from the irradiated fuel of the reactor in Vinca

    International Nuclear Information System (INIS)

    This study describes the technology for separation of uranium, plutonium and fission products from the radioactive water solution which is obtained by dissolving the spent uranium fuel from the reactor in Vinca. The procedure should be completed in a hot cell, with the maximum permitted activity of 10 Ci

  4. Evaluation of health effects in Sequoyah Fuels Corporation workers from accidental exposure to uranium hexafluoride

    International Nuclear Information System (INIS)

    Urine bioassay measurements for uranium and medical laboratory results were studied to determine whether there were any health effects from uranium intake among a group of 31 workers exposed to uranium hexafluoride (UF6) and hydrolysis products following the accidental rupture of a 14-ton shipping cylinder in early 1986 at the Sequoyah Fuels Corporation uranium conversion facility in Gore, Oklahoma. Physiological indicators studied to detect kidney tissue damage included tests for urinary protein, casts and cells, blood, specific gravity, and urine pH, blood urea nitrogen, and blood creatinine. We concluded after reviewing two years of follow-up medical data that none of the 31 workers sustained any observable health effects from exposure to uranium. The early excretion of uranium in urine showed more rapid systemic uptake of uranium from the lung than is assumed using the International Commission on Radiological Protection (ICRP) Publication 30 and Publication 54 models. The urinary excretion data from these workers were used to develop an improved systemic recycling model for inhaled soluble uranium. We estimated initial intakes, clearance rates, kidney burdens, and resulting radiation doses to lungs, kidneys, and bone surfaces. 38 refs., 10 figs., 7 tabs

  5. Experiments of JRR-4 low-enriched-uranium-silicied fuel core

    International Nuclear Information System (INIS)

    JRR-4, a light-water-moderated and cooled, swimming pool type research reactor using high-enriched uranium plate-type fuels had been operated from 1965 to 1996. In order to convert to low-enriched-uranium-silicied fuels, modification work had been carried out for 2 years, from 1996 to 1998. After the modification, start-up experiments were carried out to obtain characteristics of the low-enriched-uranium-silicied fuel core. The measured excess reactivity, reactor shutdown margin and the maximum reactivity addition rate satisfied the nuclear limitation of the safety report for licensing. It was confirmed that conversion to low-enriched-uranium-silicied fuels was carried out properly. Besides, the necessary data for reactor operation were obtained, such as nuclear, thermal hydraulic and reactor control characteristics. This report describes the results of start-up experiments and burnup experiments. The first criticality of low-enriched-uranium-silicied core was achieved on 14th July 1998, and the operation for joint-use has been carried out since 6th October 1998. (author)

  6. Waste management in MOX fuel fabrication plants

    International Nuclear Information System (INIS)

    After a short description of a MOX fuel fabrication plant's activities the waste arisings in such a plant are discussed according to nature, composition, Pu-content. Experience has shown that proper recording leads to a reduction of waste arisings by waste awareness. Aspects of the treatment of α-waste are given and a number of treatment processes are reviewed. Finally, the current waste management practice and the α-waste treatment facility under construction at ALKEM are outlined. (orig./RW)

  7. Holdup measurement for nuclear fuel manufacturing plants

    Energy Technology Data Exchange (ETDEWEB)

    Zucker, M.S.; Degen, M.; Cohen, I.; Gody, A.; Summers, R.; Bisset, P.; Shaub, E.; Holody, D.

    1981-07-13

    The assay of nuclear material holdup in fuel manufacturing plants is a laborious but often necessary part of completing the material balance. A range of instruments, standards, and a methodology for assaying holdup has been developed. The objectives of holdup measurement are ascertaining the amount, distribution, and how firmly fixed the SNM is. The purposes are reconciliation of material unbalance during or after a manufacturing campaign or plant decommissioning, to decide security requirements, or whether further recovery efforts are justified.

  8. The credit analysis of recycling beryllium and uranium in BeO-UO2 nuclear fuel

    International Nuclear Information System (INIS)

    This study quantifies the credits of beryllium and uranium which are used as the raw materials for BeO-UO2 nuclear fuel by analyzing the influence of their credits on the nuclear fuel cycle cost was analyzed, where the credit was defined as the value of raw materials recovered from spent fuel and the raw materials that were re-cycled. The credits of beryllium and uranium at 60 MWD/kg burn-up were -0.22 Mills/kWh and -0.14 Mills/kWh, respectively. These findings were based on the assumption that the optimal mixing proportion of beryllium in the BeO-UO2 nuclear fuel is 4.8 wt%. In sum, the present study verified that the credits of beryllium and uranium in relation to BeO-UO2 nuclear fuel are significant cost drivers in the cost of the nuclear fuel cycle and in estimating the nuclear fuel cycle of the reprocessing option for spent nuclear fuels. (author)

  9. Uranium to Electricity: The Chemistry of the Nuclear Fuel Cycle

    Science.gov (United States)

    Settle, Frank A.

    2009-01-01

    The nuclear fuel cycle consists of a series of industrial processes that produce fuel for the production of electricity in nuclear reactors, use the fuel to generate electricity, and subsequently manage the spent reactor fuel. While the physics and engineering of controlled fission are central to the generation of nuclear power, chemistry…

  10. Direct FuelCell/Turbine Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Hossein Ghezel-Ayagh

    2008-09-30

    This report summarizes the progress made in development of Direct FuelCell/Turbine (DFC/T{reg_sign}) power plants for generation of clean power at very high efficiencies. The DFC/T system employs an indirectly heated Turbine Generator to supplement fuel cell generated power. The concept extends the high efficiency of the fuel cell by utilizing the fuel cell's byproduct heat in a Brayton cycle. Features of the DFC/T system include: electrical efficiencies of up to 75% on natural gas, minimal emissions, reduced carbon dioxide release to the environment, simplicity in design, direct reforming internal to the fuel cell, and potential cost competitiveness with existing combined cycle power plants. Proof-of-concept tests using a sub-MW-class DFC/T power plant at FuelCell Energy's (FCE) Danbury facility were conducted to validate the feasibility of the concept and to measure its potential for electric power production. A 400 kW-class power plant test facility was designed and retrofitted to conduct the tests. The initial series of tests involved integration of a full-size (250 kW) Direct FuelCell stack with a 30 kW Capstone microturbine. The operational aspects of the hybrid system in relation to the integration of the microturbine with the fuel cell, process flow and thermal balances, and control strategies for power cycling of the system, were investigated. A subsequent series of tests included operation of the sub-MW Direct FuelCell/Turbine power plant with a Capstone C60 microturbine. The C60 microturbine extended the range of operation of the hybrid power plant to higher current densities (higher power) than achieved in initial tests using the 30kW microturbine. The proof-of-concept test results confirmed the stability and controllability of operating a fullsize (250 kW) fuel cell stack in combination with a microturbine. Thermal management of the system was confirmed and power plant operation, using the microturbine as the only source of fresh air supply

  11. Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies

    International Nuclear Information System (INIS)

    This thesis examines and confirms the feasibility of using non-uranium fuel in a pressurized water reactor (PWR) radial blanket to eliminate plutonium of both weapons and civilian origin. In the equilibrium cycle, the periphery of the PWR is loaded with alternating fresh and once burned non-uranium fuel assemblies, with the interior of the core comprised of conventional three batch UO2 assemblies. Plutonium throughput is such that there is no net plutonium production: production in the interior is offset by destruction in the periphery. Using this approach a 50 MT WGPu inventory could be eliminated in approximately 400 reactor years of operation. Assuming all other existing constraints were removed, the 72 operating US PWRs could disposition 50 MT of WGPu in 5.6 years. Use of a low fissile loading plutonium-erbium inert-oxide-matrix composition in the peripheral assemblies essentially destroys 100% of the 239Pu and ≥90% totalPu over two 18 month fuel cycles. Core radial power peaking, reactivity vs EFPD profiles and core average reactivity coefficients were found to be comparable to standard PWR values. Hence, minimal impact on reload licensing is anticipated. Examination of potential candidate fuel matrices based on the existing experience base and thermo-physical properties resulted in the recommendation of three inert fuel matrix compositions for further study: zirconia, alumina and TRISO particle fuels. Objective metrics for quantifying the inherent proliferation resistance of plutonium host waste and fuel forms are proposed and were applied to compare the proposed spent WGPu non-uranium fuel to spent WGPu MOX fuels and WGPu borosilicate glass logs. The elimination disposition option spent non-uranium fuel product was found to present significantly greater barriers to proliferation than other plutonium disposal products

  12. Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Chodak, P. III

    1996-05-01

    This thesis examines and confirms the feasibility of using non-uranium fuel in a pressurized water reactor (PWR) radial blanket to eliminate plutonium of both weapons and civilian origin. In the equilibrium cycle, the periphery of the PWR is loaded with alternating fresh and once burned non-uranium fuel assemblies, with the interior of the core comprised of conventional three batch UO{sub 2} assemblies. Plutonium throughput is such that there is no net plutonium production: production in the interior is offset by destruction in the periphery. Using this approach a 50 MT WGPu inventory could be eliminated in approximately 400 reactor years of operation. Assuming all other existing constraints were removed, the 72 operating US PWRs could disposition 50 MT of WGPu in 5.6 years. Use of a low fissile loading plutonium-erbium inert-oxide-matrix composition in the peripheral assemblies essentially destroys 100% of the {sup 239}Pu and {ge}90% {sub total}Pu over two 18 month fuel cycles. Core radial power peaking, reactivity vs EFPD profiles and core average reactivity coefficients were found to be comparable to standard PWR values. Hence, minimal impact on reload licensing is anticipated. Examination of potential candidate fuel matrices based on the existing experience base and thermo-physical properties resulted in the recommendation of three inert fuel matrix compositions for further study: zirconia, alumina and TRISO particle fuels. Objective metrics for quantifying the inherent proliferation resistance of plutonium host waste and fuel forms are proposed and were applied to compare the proposed spent WGPu non-uranium fuel to spent WGPu MOX fuels and WGPu borosilicate glass logs. The elimination disposition option spent non-uranium fuel product was found to present significantly greater barriers to proliferation than other plutonium disposal products.

  13. Reactivity change measurements on plutonium-uranium fuel elements in hector experimental techniques and results

    International Nuclear Information System (INIS)

    The techniques used in making reactivity change measurements on HECTOR are described and discussed. Pile period measurements were used in the majority of oases, though the pile oscillator technique was used occasionally. These two methods are compared. Flux determinations were made in the vicinity of the fuel element samples using manganese foils, and the techniques used are described and an error assessment made. Results of both reactivity change and flux measurements on 1.2 in. diameter uranium and plutonium-uranium alloy fuel elements are presented, these measurements being carried out in a variety of graphite moderated lattices at temperatures up to 450 deg. C. (author)

  14. The relationship of JNC and JCO in the uranium processing plant criticality accident. The second revision edition

    International Nuclear Information System (INIS)

    On September 30th 1999, the criticality accident occurred at JCO's uranium conversion building in Tokai. The accident occurred during reconversion from U3O8 to uranium nitrate solution (UNH) with uranium enriched 18.8% and about 60 kgU. JCO contracted with JNC to supply UNH that is fuel material for the experimental fast breeder reactor 'JOYO'. JNC has contracted with JCO that had started nuclear fuel material processing business following a definite policy of Japanese government and developed SUMITOMO ADU PROCESS'. JNC made the first contract with JCO in 1985 and has made a contact every year. There had never been a problem in their products. JNC inspected products based on contract. JNC discharge our duty as customer inspecting products based on contract. As for safety control, JCO had taken licensing safety review and had been permitted to be 'a processing facility'. Therefore JNC understood that JCO produced following this license. 'The Uranium Processing Plant Criticality Accident Investigation' showed that JCO had been taking a different method from the permit and violating the license. However JNC had never been explained about that and JCO's operation procedures had never described about that. Therefore the Criticality Accident couldn't be avoided. The reports is the revision of former JNC TN8420 2003-003. (author)

  15. Sequoyah Uranium Hexafluoride Plant (Docket No. 40-8027): Final environmental statement

    International Nuclear Information System (INIS)

    The proposed action is the continuation of Source Material License SUB-1010 issued to Kerr-McGee Nuclear Corporation authorizing the operation of a uranium hexafluoride manufacturing facility located in Sequoyah County, Oklahoma, close to the confluence of the Illinois and Arkansas Rivers. The plant produces high purity uranium hexafluoride using uranium concentrates (yellowcake) as the starting material. It is currently designed to produce 5000 tons of uranium per year as uranium hexafluoride and has been in operation since February 1970 without significant environmental incident or discernible offsite effect. The manufacturing process being used includes wet chemical purification to convert yellowcake to pure uranium trioxide followed by dry chemical reduction, hydrofluorination, and fluorination technique to produce uranium hexafluoride. 8 figs, 12 tabs

  16. Accumulation of uranium in plant roots absorbed from aqueous solutions

    International Nuclear Information System (INIS)

    In order to study accumulation mechanisms of uranium (U) in terrestrial plants, uptake experiments for U have been carried out by using Indian mustard (Brassica juncea). This plant is edible and known as a heavy metal accumulator, especially for cadmium (Cd). About 30 rootsstocks of Indian mustard grown hydroponically in laboratory dishes were kept in uranyl (UO22+) nitrate solutions (initially 0.5 mmol/l) at 25degC for 24, 48 and 72 hours (h). The average U concentrations in leaves increased until 48 h up to about 0.6 mg/g and then decreased slightly. Those in roots showed similar trends, but with much higher maximum U concentrations of about 30 mg/g. Backscattered electron images under SEM of the roots showed that U was accumulated on the cell edges. EPMA elemental mapping indicated that phosphorus (P) distribution had a very strong correlation with that of U. The distribution of sulfur (S) appeared to be somewhat different form these U and P distributions. These results suggest that U can be absorbed into plant roots as uranyl (UO22+) and might be fixed at the phospholipid rich cell membranes. This U accumulation mechanism appeared to be different from that for Cd which has a close association with S. (author)

  17. Uranium

    International Nuclear Information System (INIS)

    The author discusses the contribution made by various energy sources in the production of electricity. Estimates are made of the future nuclear contribution, the future demand for uranium and future sales of Australian uranium. Nuclear power growth in the United States, Japan and Western Europe is discussed. The present status of the six major Australian uranium deposits (Ranger, Jabiluka, Nabarlek, Koongarra, Yeelerrie and Beverley) is given. Australian legislation relevant to the uranium mining industry is also outlined

  18. Uranium

    International Nuclear Information System (INIS)

    The development, prospecting, research, processing and marketing of South Africa's uranium industry and the national policies surrounding this industry form the headlines of this work. The geology of South Africa's uranium occurences and their positions, the processes used in the extraction of South Africa's uranium and the utilisation of uranium for power production as represented by the Koeberg nuclear power station near Cape Town are included in this publication

  19. Analysis of mdr1-1Δ mutation of MDR1 gene in the “Cimarron Uruguayo” dog

    Directory of Open Access Journals (Sweden)

    Rosa Gagliardi B.

    2013-08-01

    Full Text Available Objective. The aim of this paper is to analyze the frequency of the mdr1-1D mutation of the MDR1 gene in a dog sample of the Uruguayan Cimarron breed with the objective of increasing the knowledge of this breed’s genome. Materials and methods. Thirty-six animals of this breed were analyzed. The MDR1 gene region, which includes the location where the mutation would be present, was amplified by PCR. Results. The mutation was not detected in any of the analyzed Uruguayan Cimarron. Conclusions. The lack of described ivermectin intoxication cases in veterinary clinic in this breed is explained by the lack of the mutation object of this study. The sequence studied in Cimarron dogs is kept compared to other breeds, except Collies and related breeds (Border Collie, Bearded Collie, Old English sheepdog.

  20. Uranium

    International Nuclear Information System (INIS)

    A discussion is given of uranium as an energy source in The Australian economy. Figures and predictions are presented on the world supply-demand position and also figures are given on the added value that can be achieved by the processing of uranium. Conclusions are drawn about Australia's future policy with regard to uranium (R.L.)

  1. Uranium

    International Nuclear Information System (INIS)

    The geological setting of uranium resources in the world can be divided in two basic categories of resources and are defined as reasonably assured resources, estimated additional resources and speculative resources. Tables are given to illustrate these definitions. The increasing world production of uranium despite the cutback in the nuclear industry and the uranium requirements of the future concluded these lecture notes

  2. Uranium in soil, forest litter and living plant material above three uranium mineralizations in Northern Sweden

    International Nuclear Information System (INIS)

    In order to investigate the feasibility of biogeochemical sampling media in uranium exploration, samples from the most common trees and low bushes together with forest litter were collected over the areas of three uranium mineralizations in Northern Sweden and analyzed for uranium. The results were compared with uranium content of the till and its radioactivity. The average uranium content was low for all sample types and considerably lower in the ash of the organic sample types compared to that of the till. No sample type showed any tendency of having higher uranium concentration above mineralizations compared to background areas. These results suggest that, under conditions prevailing in Sweden, the investigated sample types are not suitable for uranium exploration

  3. Neutronics Studies Of Uranium-Based Fully Ceramic Micro-Encapsulated Fuel For PWRs

    International Nuclear Information System (INIS)

    This study evaluates the core neutronics and fuel cycle characteristics that result from employing uranium-based fully ceramic micro-encapsulated (FCM) fuel in a pressurized water reactor (PWR). Specific PWR bundle designs with FCM fuel have been developed, which by virtue of their TRISO particle based elements, are expected to safely reach higher fuel burnups while also increasing the tolerance to fuel failures. The SCALE 6.1 code package, developed and maintained at ORNL, was the primary software employed to model these designs. Analysis was performed using the SCALE double-heterogeneous (DH) fuel modeling capabilities. For cases evaluated with the NESTLE full-core three-dimensional nodal simulator, because the feature to perform DH lattice physics branches with the SCALE/TRITON sequence is not yet available, the Reactivity-Equivalent Physical Transformation (RPT) method was used as workaround to support the full core analyses. As part of the fuel assembly design evaluations, fresh feed lattices were modeled to analyze the within-assembly pin power peaking. Also, a color-set array of assemblies was constructed to evaluate power peaking and power sharing between a once-burned and a fresh feed assembly. In addition, a parametric study was performed by varying the various TRISO particle design features; such as kernel diameter, coating layer thicknesses, and packing fractions. Also, other features such as the selection of matrix material (SiC, Zirconium) and fuel rod dimensions were perturbed. After evaluating different uranium-based fuels, the higher physical density of uranium mononitride (UN) proved to be favorable, as the parametric studies showed that the FCM particle fuel design will need roughly 12% additional fissile material in comparison to that of a standard UO2 rod in order to match the lifetime of an 18-month PWR cycle. Neutronically, the FCM fuel designs evaluated maintain acceptable design features in the areas of fuel lifetime, temperature

  4. Behavior of silicon in nitric media. Application to uranium silicides fuels reprocessing

    International Nuclear Information System (INIS)

    Uranium silicides are used in some research reactors. Reprocessing them is a solution for their cycle end. A list of reprocessing scenarios has been set the most realistic being a nitric dissolution close to the classic spent fuel reprocessing. This uranium silicide fuel contains a lot of silicon and few things are known about polymerization of silicic acid in concentrated nitric acid. The study of this polymerization allows to point out the main parameters: acidity, temperature, silicon concentration. The presence of aluminum seems to speed up heavily the polymerization. It has been impossible to find an analytical technique smart and fast enough to characterize the first steps of silicic acid polymerization. However the action of silicic species on emulsions stabilization formed by mixing them with an organic phase containing TBP has been studied, Silicon slows down the phase separation by means of oligomeric species forming complex with TBP. The existence of these intermediate species is short and heating can avoid any stabilization. When non irradiated uranium silicide fuel is attacked by a nitric solution, aluminum and uranium are quickly dissolved whereas silicon mainly stands in solid state. That builds a gangue of hydrated silica around the uranium silicide particulates without preventing uranium dissolution. A small part of silicon passes into the solution and polymerize towards the highly poly-condensed forms, just 2% of initial silicon is still in molecular form at the end of the dissolution. A thermal treatment of the fuel element, by forming inter-metallic phases U-Al-Si, allows the whole silicon to pass into the solution and next to precipitate. The behavior of silicon in spent fuels should be between these two situations. (author)

  5. Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Montierth, Leland M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-07-19

    The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element design for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition as part of a fuel meat thickness optimization effort for reactor performance other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.

  6. Modeling of advanced fossil fuel power plants

    Science.gov (United States)

    Zabihian, Farshid

    The first part of this thesis deals with greenhouse gas (GHG) emissions from fossil fuel-fired power stations. The GHG emission estimation from fossil fuel power generation industry signifies that emissions from this industry can be significantly reduced by fuel switching and adaption of advanced power generation technologies. In the second part of the thesis, steady-state models of some of the advanced fossil fuel power generation technologies are presented. The impacts of various parameters on the solid oxide fuel cell (SOFC) overpotentials and outputs are investigated. The detail analyses of operation of the hybrid SOFC-gas turbine (GT) cycle when fuelled with methane and syngas demonstrate that the efficiencies of the cycles with and without anode exhaust recirculation are close, but the specific power of the former is much higher. The parametric analysis of the performance of the hybrid SOFC-GT cycle indicates that increasing the system operating pressure and SOFC operating temperature and fuel utilization factor improves cycle efficiency, but the effects of the increasing SOFC current density and turbine inlet temperature are not favourable. The analysis of the operation of the system when fuelled with a wide range of fuel types demonstrates that the hybrid SOFC-GT cycle efficiency can be between 59% and 75%, depending on the inlet fuel type. Then, the system performance is investigated when methane as a reference fuel is replaced with various species that can be found in the fuel, i.e., H2, CO2, CO, and N 2. The results point out that influence of various species can be significant and different for each case. The experimental and numerical analyses of a biodiesel fuelled micro gas turbine indicate that fuel switching from petrodiesel to biodiesel can influence operational parameters of the system. The modeling results of gas turbine-based power plants signify that relatively simple models can predict plant performance with acceptable accuracy. The unique

  7. Evaluation of the reactivity feedback coefficients of the Da Lat Nuclear Research Reactor using highly enriched uranium fuels and low enriched uranium fuels

    International Nuclear Information System (INIS)

    This article presents the calculation results of fuel and moderator temperature coefficients of reactivity for the Da Lat nuclear research reactor (DNRR), using Highly Enriched Uranium (HEU) fuels (36%) and Low Enriched Uranium (LEU) fuels (19.75%). This study uses the WIMSD code to calculate the cross sections of all the reactor components at different temperatures and these group constants were used then in the CITATION code to calculate the effective multiplication factor at distinct moderator and fuel temperatures using 69 energy groups. Only one parameter was changed where all other parameters were kept constant. The results are the average of fuel temperature coefficient of reactivity for LEU: -2.15(x10-5 ∆k/k/oC), HEU: -1.91(x10-5∆k/k/oC); the average of the moderator temperature coefficients of reactivity for LEU: -2.44(x10-4 ∆k/k/oC),HEU: -2.67(x10-4 ∆k/k/oC). The calculated feedback coefficients were compared with the measured data from DNRR. Good agreements were obtained. Moreover,getting the trend of these coefficients versus the rise in temperature. (author)

  8. Experimental determination of heat transfer coefficients in uranium zirconium hydride fuel rod

    International Nuclear Information System (INIS)

    This work presents the experiments and theoretical analysis to determine the temperature parameter of the uranium zirconium hydride fuel elements, used in the TRIGA IPR-R1 Research Nuclear Reactor. The fuel thermal conductivity and the heat transfer coefficient from the cladding to the coolant were evaluated experimentally. It was also presented a correlation for the gap conductance between the fuel and the cladding. In the case of nuclear fuels the heat parameters become functions of the irradiation as a result of change in the chemical and physical composition. The value of the heat transfer coefficients should be determined experimentally. (author)

  9. Experimental determination of heat transfer coefficients in uranium zirconium hydride fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Z.; Rezende, Hugo C.; Costa, Antonio Carlos L. da [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil)]. E-mail: amir@cdtn.br; hcr@cdtn.br; aclc@cdtn.br

    2005-07-01

    This work presents the experiments and theoretical analysis to determine the temperature parameter of the uranium zirconium hydride fuel elements, used in the TRIGA IPR-R1 Research Nuclear Reactor. The fuel thermal conductivity and the heat transfer coefficient from the cladding to the coolant were evaluated experimentally. It was also presented a correlation for the gap conductance between the fuel and the cladding. In the case of nuclear fuels the heat parameters become functions of the irradiation as a result of change in the chemical and physical composition. The value of the heat transfer coefficients should be determined experimentally. (author)

  10. Direct FuelCell/Turbine Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Hossein Ghezel-Ayagh

    2008-09-30

    This report summarizes the progress made in development of Direct FuelCell/Turbine (DFC/T{reg_sign}) power plants for generation of clean power at very high efficiencies. The DFC/T system employs an indirectly heated Turbine Generator to supplement fuel cell generated power. The concept extends the high efficiency of the fuel cell by utilizing the fuel cell's byproduct heat in a Brayton cycle. Features of the DFC/T system include: electrical efficiencies of up to 75% on natural gas, minimal emissions, reduced carbon dioxide release to the environment, simplicity in design, direct reforming internal to the fuel cell, and potential cost competitiveness with existing combined cycle power plants. Proof-of-concept tests using a sub-MW-class DFC/T power plant at FuelCell Energy's (FCE) Danbury facility were conducted to validate the feasibility of the concept and to measure its potential for electric power production. A 400 kW-class power plant test facility was designed and retrofitted to conduct the tests. The initial series of tests involved integration of a full-size (250 kW) Direct FuelCell stack with a 30 kW Capstone microturbine. The operational aspects of the hybrid system in relation to the integration of the microturbine with the fuel cell, process flow and thermal balances, and control strategies for power cycling of the system, were investigated. A subsequent series of tests included operation of the sub-MW Direct FuelCell/Turbine power plant with a Capstone C60 microturbine. The C60 microturbine extended the range of operation of the hybrid power plant to higher current densities (higher power) than achieved in initial tests using the 30kW microturbine. The proof-of-concept test results confirmed the stability and controllability of operating a fullsize (250 kW) fuel cell stack in combination with a microturbine. Thermal management of the system was confirmed and power plant operation, using the microturbine as the only source of fresh air supply

  11. Uranium from Seawater Program Review; Fuel Resources Uranium from Seawater Program DOE Office of Nuclear Energy

    Energy Technology Data Exchange (ETDEWEB)

    None

    2013-07-01

    For nuclear energy to remain sustainable in the United States, economically viable sources of uranium beyond terrestrial ores must be developed. The goal of this program is to develop advanced adsorbents that can extract uranium from seawater at twice the capacity of the best adsorbent developed by researchers at the Japan Atomic Energy Agency (JAEA), 1.5 mg U/g adsorbent. A multidisciplinary team from Oak Ridge National Laboratory, Lawrence Berkeley National Laboratory, Pacific Northwest National Laboratory, and the University of Texas at Austin was assembled to address this challenging problem. Polymeric adsorbents, based on the radiation grafting of acrylonitrile and methacrylic acid onto high surface-area polyethylene fibers followed by conversion of the nitriles to amidoximes, have been developed. These poly(acrylamidoxime-co-methacrylic acid) fibers showed uranium adsorption capacities for the extraction of uranium from seawater that exceed 3 mg U/g adsorbent in testing at the Pacific Northwest National Laboratory Marine Sciences Laboratory. The essence of this novel technology lies in the unique high surface-area trunk material that considerably increases the grafting yield of functional groups without compromising its mechanical properties. This technology received an R&D100 Award in 2012. In addition, high surface area nanomaterial adsorbents are under development with the goal of increasing uranium adsorption capacity by taking advantage of the high surface areas and tunable porosity of carbon-based nanomaterials. Simultaneously, de novo structure-based computational design methods are being used to design more selective and stable ligands and the most promising candidates are being synthesized, tested and evaluated for incorporation onto a support matrix. Fundamental thermodynamic and kinetic studies are being carried out to improve the adsorption efficiency, the selectivity of uranium over other metals, and the stability of the adsorbents. Understanding

  12. Characteristics of Modular Fast Reactor SVBR-100 Using Thorium-Uranium (233) Fuel

    International Nuclear Information System (INIS)

    Conclusions: • The performed computations for three different types of fuel (oxide , nitride and metallic), have revealed that maximum of uranium-233 breeding ratio, which equals to 0.9, is achieved when nitride type of fuel is used. • Adding breeding zones or increasing of the core dimensions result in increasing uranium-233 breeding ratio (up to BR = 0,97 or BR = 0,96 respectively). • There is opportunity of using plutonium as initial fissile isotope to implement U-Th-Pu fuel cycle. Breeding ratio is assessed by 0,98 if nitride fuel composition (Th+Pu)N with effective density of 12.5 is used. • The obtained data have demonstrated that both for U-Th FC and U-Th-Pu FC there is an opportunity to achieve a value of U-233 BR to be over unity when using the breeding zones and slightly increased the core dimensions

  13. Feasibility study of using of low enriched uranium fuel for research reactor in Sofia

    International Nuclear Information System (INIS)

    The results of the study of arrangement of the reactor core of pool type Research Reactor 200 kW in Sofia are presented. The number of planned horizontal experimental channels of Research Reactor is equal to 8 and vertical ones to 9. The beginning of study performed for fresh IRT2M tubular fuel assemblies containing highly enriched uranium with 36% enrichment is presented. The core modelling calculations are performed by the MCNP4C code. The IRT-2M fuel assembly modelling is tested by experimental data. The configuration of the reactor core including three-tube and four-tube fuel assemblies as well as beryllium blocks is analysed. The aspects of feasibility of using of developed in Russia pin-type fuel assemblies IRT-MR with low enriched uranium (LEU<20%) in core design instead of IRT-2M ones are discussed. (author)

  14. Thermal conductivity of uranium-plutonium oxide fuel for fast reactors

    International Nuclear Information System (INIS)

    A new thermal conductivity correlation for fully dense uranium-plutonium oxide fuel for fast reactors was formulated for fuel pin thermal analysis under beginning of irradiation conditions. The data set used in correlating the equation was systematically selected to minimize experimental uncertainty. The electron conduction term for uranium dioxide formulated by Harding and Martin [J. Nucl. Mater. 166 (1989) 223] was adopted to compensate for so few high temperature measurements. The excellent predictability of the new correlation was validated by comparing the calculated with measured fuel center temperatures in an instrumented irradiation test in the experimental fast reactor JOYO for low oxygen-to-metal (O/M) ratio fuel up to 1850 K

  15. Procedure and results of decommissioning of R and D facility of uranium fuel

    International Nuclear Information System (INIS)

    From 1998 through 2005, the facilities for research and development (R and D) of uranium ore-dressing and uranium fuel etc. were decommissioned and soil contaminated by uranium was collected. All the pieces of apparatus in the nuclear facilities which might be contaminated with uranium were treated as radioactive wastes. At the time of the decommissioning activity, there was no specific value to judge as radioactive wastes. So MMC considered and adopted the pragmatic procedure to judge that soil was radioactive waste or not. During decommissioning facilities and collecting soil, the environmental monitoring was conducted. And it was confirmed that these activities had no influence on the surrounding areas. All decommissioning activities were finished with no difficulty. The wastes generated from the decommissioning activities were packed in the steel containers and have been stored safely in the storehouse built in the same area. In this report, the details of decommissioning activities are described. (author)

  16. Development of the uranium continuous casting technology for metal rods with simulated metallic spent fuels

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Y. S.; Kim, C. K.; Kim, K. H. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-05-01

    To dispose spent fuel efficiently, Korea Atomic Energy Research Institute has been developing a storage process, in which uranium metal abstracted by Li-reduction is being formed to long rods and then the rods are being arranged uniformly in canister. The purpose of this study is to develop a continuous casting technology for uranium metal rods with small diameter and long length. For this purpose, the vertical continuous casting equipment was developed and optimum condition for continuous casting process of uranium was investigated. With Si3N4 mold which was designed to have a proper thermal gradient, we has succeeded to get a optimum process condition and to cast a 13.7 mm diameter and 2500 mm length of uranium rod. 24 refs., 23 figs., 7 tabs. (Author)

  17. Volatile behaviour of enrichment uranium in the total nuclear fuel price

    International Nuclear Information System (INIS)

    In this article the historical high volatile behaviour of the total nuclear fuel price is evaluated quantitatively and it is concluded that it has been due mainly to the fluctuations of the price of the principal components of enriched uranium (concentrates and enrichment). In order to avoid the negative effects of this volatiles behaviour as far as possible, a basic strategy in the uranium procurement activities is recommended (union of buyers, diversification of supplier, stock management, optimisation of contract portfolio and suitable currency management that guarantees a reliable uranium supply at reasonable prices. These guidelines are those that ENUSA has been following on behalf of the Spanish Utilities in the Commission of Uranium Procurement (CAU in Spanish). (Author) 11 refs

  18. Reactivity temperature coefficient evaluation of uranium zirconium hydride fuel element in power reactor

    International Nuclear Information System (INIS)

    Highlights: ► We develop an in-core fuel management code package for uranium zirconium hydride power reactor. ► The influence of changes on U–ZrHx fuel element is calculated and analyzed theoretically. ► Increased uranium contents in U–ZrHx reduce prompt negative temperature coefficient markedly. ► Additional poison erbium makes prompt negative temperature coefficient much more negative. ► The characteristics of inherent safety of U–ZrHx core can be retained in power reactors. -- Abstract: An in-core fuel management code package for uranium zirconium hydride power reactor, which is developed on the basis of the assembly lattice code TPFAP and the core calculating code BMFGD for LWR, is firstly introduced in this paper. The inherent safety of the U–ZrHx element which is mainly caused by the high prompt negative temperature coefficient is then evaluated, because the weight percentage of uranium, fuel rod radius and fuel temperature of U–ZrHx element will be different in power reactor from those in research reactor, and these changes may make obvious effect on the prompt negative temperature coefficient. The influence of weight percentage of uranium, fuel rod radius, fuel temperature, content of hydrogen and additional poison on prompt negative temperature coefficient for uranium zirconium hydride element are calculated respectively in this paper, and then the results are analyzed theoretically. The study shows that the absolute value of prompt negative temperature coefficient reduces observably along with the increasing of Uranium weight percentage from 10 wt% in research reactor to maximum 45 wt% in power reactor. Smaller radius, higher operating temperature and longer core life make little effect on the prompt negative temperature coefficient in the condition of high weight percentage of U. Additional poison erbium in fuel makes prompt negative temperature coefficient much more negative. Anyway, high prompt negative temperature coefficient can

  19. Tritium management in PWR fuel reprocessing plants

    International Nuclear Information System (INIS)

    Activity, quantity and nature of tritium compounds obtained during head end process (cutting and dissolution) are determined to estimate environmental release hazards in fuel reprocessing plants. Measurements on representative PWR reactor fuels (burnup 33,000 MWdt-1, specific power 30 MW dt-1) show that about 60% of the tritium produced in the reactor diffuses in the cladding where it is fixed. Remaining tritium stays in the irradiated oxide and is found as tritiated water in the solution obtained during fuel dissolution. In the UP3 plant at La Hague (France) tritiated water is disposed into the sea without environmental problems. In the case of a reprocessing plant far from the sea, the PUREX process is slightly modified for concentration of tritium in a limited amount of water (TRILEX process). It is verified experimentally in αβγ lab on actual fuel and by simulation at the pilot seale that the supplementary step ''tritium washing'' of the solvent can be obtained in pulsed columns. 4 tables, 7 figs

  20. Neutronics studies of uranium-based fully ceramic micro-encapsulated fuel for PWRs

    International Nuclear Information System (INIS)

    This study evaluates the core neutronics and fuel cycle characteristics using uranium-based fully ceramic micro-encapsulated (FCM) fuel in a pressurized water reactor (PWR). Specific PWR assembly designs with FCM fuel have been developed, which by virtue of their TRISO particle-based elements are expected to achieve higher fuel burnups while also increasing the tolerance to fuel failures. The SCALE 6.1 code package, developed and maintained at ORNL, was the primary software used to model the assembly designs. Analysis was performed using the SCALE double-heterogeneous (DH) fuel modeling capabilities; however, the Reactivity-Equivalent Physical Transformation (RPT) method was used for lattice calculations due to the long run times associated with the SCALE DH capability. In order to understand the impact on reactivity and reactor operating cycle length, a parametric study was performed by varying TRISO particle design features, such as kernel diameter, coating layer thicknesses, and packing fraction. Also, other features such as the selection of matrix material (SiC, zirconium) and fuel rod dimensions were studied. After evaluating different uranium-based fuels, the higher compound density of uranium mononitride (UN) proved to be favorable, as the parametric studies showed that the FCM particle fuel design will need roughly 12% additional fissile material in comparison to that of a standard UO2 rod in order to match the lifetime of an 18-month PWR cycle. Neutronically, the FCM fuel designs evaluated maintain acceptable design features in the areas of fuel lifetime and temperature coefficients of reactivity, as well as pin cell and assembly peaking factors. (authors)

  1. Neutronics studies of uranium-based fully ceramic micro-encapsulated fuel for PWRs

    Energy Technology Data Exchange (ETDEWEB)

    George, N. M.; Maldonado, I. [Dept. of Nuclear Engineering, Univ. of Tennessee Knoxville, Knoxville, TN 37996-2300 (United States); Terrani, K.; Godfrey, A.; Gehin, J. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)

    2012-07-01

    This study evaluates the core neutronics and fuel cycle characteristics using uranium-based fully ceramic micro-encapsulated (FCM) fuel in a pressurized water reactor (PWR). Specific PWR assembly designs with FCM fuel have been developed, which by virtue of their TRISO particle-based elements are expected to achieve higher fuel burnups while also increasing the tolerance to fuel failures. The SCALE 6.1 code package, developed and maintained at ORNL, was the primary software used to model the assembly designs. Analysis was performed using the SCALE double-heterogeneous (DH) fuel modeling capabilities; however, the Reactivity-Equivalent Physical Transformation (RPT) method was used for lattice calculations due to the long run times associated with the SCALE DH capability. In order to understand the impact on reactivity and reactor operating cycle length, a parametric study was performed by varying TRISO particle design features, such as kernel diameter, coating layer thicknesses, and packing fraction. Also, other features such as the selection of matrix material (SiC, zirconium) and fuel rod dimensions were studied. After evaluating different uranium-based fuels, the higher compound density of uranium mononitride (UN) proved to be favorable, as the parametric studies showed that the FCM particle fuel design will need roughly 12% additional fissile material in comparison to that of a standard UO{sub 2} rod in order to match the lifetime of an 18-month PWR cycle. Neutronically, the FCM fuel designs evaluated maintain acceptable design features in the areas of fuel lifetime and temperature coefficients of reactivity, as well as pin cell and assembly peaking factors. (authors)

  2. HTGR fuel development: loading of uranium on carboxylic acid cation-exchange resins using solvent extraction of nitrate

    International Nuclear Information System (INIS)

    The reference fuel kernel for recycle of 233U to HTGR's (High-Temperature Gas-Cooled Reactors) is prepared by loading carboxylic acid cation-exchange resins with uranium and carbonizing at controlled conditions. The purified 233UO2(NO3)2 solution from a fuel reprocessing plant contains excess HNO3 (NO3-/U ratio of approximately 2.2). The reference flowsheet for a 233U recycle fuel facility at Oak Ridge uses solvent extraction of nitrate by a 0.3 M secondary amine in a hydrocarbon diluent to prepare acid-deficient uranyl nitrate. This nitrate extraction, along with resin loading and amine regeneration steps, was demonstrated in 14 runs. No significant operating difficulties were encountered. The process is controlled via in-line pH measurements for the acid-deficient uranyl nitrate solutions. Information was developed on pH values for uranyl nitrate solution vs NO3-/U mole ratios, resin loading kinetics, resin drying requirements, and other resin loading process parameters. Calculations made to estimate the capacities of equipment that is geometrically safe with respect to control of nuclear criticality indicate 100 kg/day or more of uranium for single nitrate extraction lines with one continuous resin loading contactor or four batch loading contactors. (auth)

  3. Material surveillance and verification program at a uranium enriching plant

    International Nuclear Information System (INIS)

    A license for a nuclear facility in the United States is approved only after a licensee demonstrates by procedure or practice that an adequate material control system exists. A license can specify acceptable material control practices. Therefore, processors in the United States receiving uranium hexafluoride (UF6) from a U. S. Government-owned enriching plant can accept shipper's values for nuclear material accounting purposes if: there is surveillance during withdrawal of the UF6, an independent sample is obtained, and certain measurement verification is subsequently performed by the receiver or the receiver's agent. Because of the high equipment and operating costs, essentially all UF6 processors have adopted a surveillance and verification program. A resident observer is employed to perform surveillance, obtain samples, and tamper-safe the shipping cylinders. Samples are analyzed by the receiver or by an independent laboratory. The observer determines by surveillance that withdrawals, or transfers of material, weighings, and sampling are accomplished in accordance with accepted procedures. Surveillance of the withdrawals includes observing the transfer of UF6 from the enriching plant cylinder to the shipping cylinder(s) and the withdrawal of samples. In addition, it inclu

  4. Role of uranium speciation in the uptake and translocation of uranium by plants

    International Nuclear Information System (INIS)

    Uranium (U) uptake and translocation by plants was characterized using a computer speciation model to develop a nutrient culture system that provided U as a single predominant species in solution. A hydroponic uptake study determined that at pH 5.0, the uranyl (UO22+) cation was more readily taken up and translocated by peas (Pisum sativum) than the hydroxyl and carbonate U complexes present in the solution at pH 6.0 and 8.0, respectively. A subsequent experiment tested the extent to which various monocot and dicot species take up and translocate the uranyl cation. Of the species screened, tepary bean (Phaseolus acutifolius) and red beet (Beta vulgaris) were the species showing the greatest accumulation of U. In addition to providing fundamental information regarding U uptake by plants, the results obtained also have implications for the phytoremediation of U-contaminated soils. The initial characterization of U uptake by peas suggested that in the field, a soil pH of <5.5 would be required in order to provide U in the most plant-available form. A pot study using U-contaminated soil was therefore conducted to assess the extent to which two soil amendments, HEDTA and citric acid, were capable of acidifying the soil, increasing U solubility, and enhancing U uptake by red beet. Of these two amendments, only citric acid proved effective, decreasing the soil pH to 5.0 and increasing U accumulation by a factor of 14. The results of this pot study provide a basis for the development of an effective phytoremediation strategy for U-contaminated soils. (author)

  5. Metallography of plutonium, uranium and thorium fuels: two decades of experience in Radiometallurgy Division

    International Nuclear Information System (INIS)

    Ever since the inception of Radiometallurgy Laboratory (RML) in its early seventies optical metallography has played a key role in development and fabrication of plutonium, uranium and thorium bearing nuclear fuels. In this report, an album of photomicrographs depicts the different types of metallic, ceramic and dispersion fuels and welded section that have been evaluated in RML during the last two decades. (author). 14 refs., 1 tab

  6. MUICYCL and MUIFAP: models tracking minor uranium isotopes in the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Two computer programs have been written to provide information on the buildup of minor uranium isotopes in the nuclear fuel cycle. The Minor Uranium Isotope Cycle Program, MUICYCL, tracks fuel through a multiyear campaign cycle of enrichment, reactor burnup, reprocessing, enrichment, etc. MUICYCL facilities include preproduction stockpiles, U235 escalation, and calculation of losses. The Minor Uranium Isotope Flowsheet Analyzer Program, MUIFAP, analyzes one minor isotope in one year of an enrichment operation. The formulation of the enrichment cascade, reactors, and reprocessing facility is presented. Input and output descriptions and sample cases are presented. The programs themselves are documented by short descriptions of each routine, flowcharts, definitions of common blocks and variables, and internal documentation. The programs are written in FORTRAN for use in batch mode

  7. The taking and verification of a physical inventory in a Low Enriched Uranium Fabrication Plant

    International Nuclear Information System (INIS)

    The paper contains a description of a Low Enriched Uranium Fabrication Plant making fuel elements for Light Water Reactors in Europe and is subject to Euratom and IAEA Safeguards. The process starts from UO-2 powder and ends with the finished elements and has an inventory of about 500 Te. The operators' actions to clean up the plant in order to establish the inventory and to prepare the material in a form so that the inspectors can verify it are described together with an estimate of the cost. A short description of the computerised quasi-real time system of accountancy and control which enables the operator to prepare a list of inventory items from which the sampling plans are made is included. The inspection activities are described in some detail including the basis of the sampling plans used and the selection of the samples. This is followed by the results of the measurements made using a neutron interrogation device for bulk materials and an active neutron coincidence collar for the finished fuel elements. The results and conclusions are surveyed including the calculation of the LEMUF

  8. Irradiation behavior of uranium oxide - Aluminum dispersion fuel

    International Nuclear Information System (INIS)

    An oxide version of the DART code has been generated in order to assess the irradiation behavior of UO2-Al dispersion fuel. The aluminum-fuel interaction models were developed based on U3O8-Al irradiation data. Deformation of the fuel element occurs due to fuel particle swelling driven by both solid and gaseous fission products and as a consequence of the interaction between the fuel particles and the aluminum matrix. The calculations show that, with the assumption that the correlations derived from U3O8 are valid for UO2, the LEU UO2-Al with a 42% fuel volume loading (4 g U/cm3 ) irradiated at fuel temperatures greater than 413 K should undergo breakaway swelling at core burnups greater than about 1.12 x 1027 fissions m-3 (∼63% 235U burnup). (author)

  9. The Cigar Lake uranium deposit: Analog information for Canada's nuclear fuel waste disposal concept

    International Nuclear Information System (INIS)

    The Cigar Lake uranium deposit, located in northern Saskatchewan, has many features that parallel those being considered within the Canadian concept for disposal of nuclear fuel waste. The study of these natural structures and processes provides valuable insight toward the eventual design and site selection of a nuclear fuel waste repository. The main feature of this analog is the absence of any indication on the surface of the rich uranium ore 450 m below. This shows that the combination of natural barriers has been effective in isolating the uranium ore from the surface environment. More specifically, the deposit provides analog information relevant to the stability of UO2 fuel waste, the performance of clay-based and general aspects of water-rock interaction. The main geotechnical studies on this deposit focus on the evolution of groundwater compositions in the deposit and on their redox chemistry with respect to the uranium, iron and sulphide systems. This report reviews and summarizes the analog information and data from the Cigar Lake analog studies for the processes and scenarios expected to occur in the disposal system for used nuclear fuel proposed in Canada. (author). 45 refs., 10 figs

  10. Romanian irradiation experiment on AHWR type fuel elements containing mixed oxide of thorium and uranium pellets

    International Nuclear Information System (INIS)

    One of the main objectives of the Institute for Nuclear Research (ICN) - Nuclear Fuel R and D Program is the development of new types of fuel based on: Slightly Enriched Uranium (SEU), Recycled Uranium (RU) and Thorium. Two experimental fuel elements (A23 and A24) were irradiated in TRIGA Research Reactor of ICN-Romania (C1 device) in a power ramp conditions. Element A23, contained mixed oxide of thorium and uranium pellets, achieved a maximum linear power of 51 KW/m and has reached a discharge burn-up around 189.2 MWh/KgHE; element A24, contained dioxide of uranium pellets, achieved a maximum linear power of 63 kW/m and has reached a discharge burn-up of around 207.8 MWh/KgHE. The experiments simulation has been performed using an improved version of ROFEM Code, version developed through the efforts of researchers from ICN - Nuclear Fuel Performance Department. The simulation results are in good agreement with experimental data. (author)

  11. Characteristics of Modular Fast Reactor SVBR-100 Using Thorium-Uranium (233) Fuel

    International Nuclear Information System (INIS)

    Natural reserves of thorium are three times as much as those of uranium. For that reason, thorium is a very promising raw material for manufacturing an artificial fissionable isotope of uranium-233 that is formed when neutrons are absorbed by thorium. Many countries are investigating characteristics of reactors using thorium-uranium (233) fuel. First, breeding ratio (BR) is of interest because only when BR = 1, the reactor can operate in a closed fuel cycle in a mode of fuel self-providing without makeup by other fissionable isotopes. The report presents the results of calculations of neutron-physical and thermal-hydraulic characteristics of SVBR-100 - lead-bismuth cooled small power modular fast reactor using thorium-uranium (233) fuel. Reactor SVBR-100 has specific properties of inherent self-protection and passive safety. The NPP modular power-units, which power equals to a value divisible by 100 MWe, can be constructed on the basis of reactor modules SVBR-100. (author)

  12. Prospect of Uranium Silicide fuel element with hypostoichiometric (Si ≤3.7%)

    International Nuclear Information System (INIS)

    An attempt to obtain high uranium-loading in silicide dispersion fuel element using the fabrication technology applicable nowadays can reach Uranium-loading slightly above 5 gU/cm3. It is difficult to achieve a higher uranium-loading than that because of fabricability constraints. To overcome those difficulties, the use of uranium silicide U3Si based is considered. The excess of U is obtained by synthesising U3Si2 in Si-hypostoichiometric stage, without applying heat treatment to the ingot as it can generate undesired U3Si. The U U will react with the matrix to form U alx compound, that its pressure is tolerable. This experiment is to consider possibilities of employing the U3Si2 as nuclear fuel element which have been performed by synthesising U3Si2-U with the composition of 3.7 % weigh and 3 % weigh U. The ingot was obtained and converted into powder form which then was fabricated into experimental plate nuclear fuel element. The interaction between free U and Al-matrix during heat-treatment is the rolling phase of the fuel element was observed. The study of the next phase will be conducted later

  13. Critical review of analytical techniques for safeguarding the thorium-uranium fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Hakkila, E.A.

    1978-10-01

    Conventional analytical methods applicable to the determination of thorium, uranium, and plutonium in feed, product, and waste streams from reprocessing thorium-based nuclear reactor fuels are reviewed. Separations methods of interest for these analyses are discussed. Recommendations concerning the applicability of various techniques to reprocessing samples are included. 15 tables, 218 references.

  14. Comparison of radiotoxicity of uranium, plutonium, and thorium spent nuclear fuel at long-term storage

    International Nuclear Information System (INIS)

    Time dependence of radio-toxicity of actinides from spent uranium, MOX-plutonium, and thorium fuel calculated for storage during 1000 years is discussed in the paper. Calculations are based on the nuclear fuel of the VVER-1000 type reactor. Recommendations for uranium and plutonium spent fuel could be done to perform chemical separation of plutonium, americium, curium before long-term controllable storage. Americium should be separated after 50-70 years of storage for sufficient conversion of Pu-241 in Am-241. Cm-244 decays almost completely after 100 years. Extracted americium (possibly, with long-lived curium isotopes) should be directed to transmutation and plutonium should be reused. The separation of actinides is also effective to reduce decay heat power. In thorium spent fuel, the overwhelming share of radio-toxicity is determined by U-232. It is obvious that the repeated use of thorium fuel will be accompanied by accumulation of radio-toxicity. For a one-fold use of thorium fuel with deep U-233 burnup, it is necessary to perform additional deep burn-out (transmutation) of uranium fraction containing both U-233 and U-232. The further reduction of radio-toxicity by several orders can be obtained by extraction and transmutation of plutonium fraction (Pu-238). The transmutation of Th-228 - daughter nuclide of U-232 - is not necessary because Th-228 decays practically completely in 10 years together with its short-lived daughter nuclides

  15. Building dismantlement and site remediation at the Apollo Fuel Plant: When is technology the answer?

    International Nuclear Information System (INIS)

    The Apollo fuel plant was located in Pennsylvania on a site known to have been used continuously for stell production from before the Civil War until after World War II. Then the site became a nuclear fuel chemical processing plants. Finally it was used to convert uranium hexafluoride to various oxide fuel forms. After the fuel manufacturing operations were teminated, the processing equipment was partially decontaminated, removed, packaged and shipped to a licensed low-level radioactive waste burial site. The work was completed in 1984. In 1990 a detailed site characterization was initiated to establishe the extent of contamination and to plan the building dismantlement and soil remediation efforts. This article discusses the site characterization and remedial action at the site in the following subsections: characterization; criticality control; mobile containment; soil washing; in-process measurements; and the final outcome of the project

  16. Standard specification for uranium metal enriched to more than 15 % and less Than 20 % 235U

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2000-01-01

    1.1 This specification covers nuclear grade uranium metal that has either been processed through an enrichment plant, or has been produced by the blending of highly enriched uranium with other uranium, to obtain uranium of any 235U concentration below 20 % (and greater than 15 %) and that is intended for research reactor fuel fabrication. The scope of this specification includes specifications for enriched uranium metal derived from commercial natural uranium, recovered uranium, or highly enriched uranium. Commercial natural uranium, recovered uranium and highly enriched uranium are defined in Section 3. The objectives of this specification are to define the impurity and uranium isotope limits for commercial grade enriched uranium metal. 1.2 This specification is intended to provide the nuclear industry with a standard for enriched uranium metal which is to be used in the production of research reactor fuel. In addition to this specification, the parties concerned may agree to other appropriate conditions. ...

  17. Thorium–based fuel cycles : saving uranium in a 200 MWth pebble bed high temperature reactor / S.K. Gintner

    OpenAIRE

    Gintner, Stephan Konrad

    2010-01-01

    The predominant nuclear fuel used globally at present is uranium which is a finite resource. Thorium has been identified as an alternative nuclear fuel source that can be utilized in almost all existing uranium–based reactors and can significantly help in conserving limited uranium reserves. Furthermore, the elimination of proliferation risks associated with thorium–based fuel cycles is a key reason for re–evaluating the possible utilization of thorium in high temperature reactors. In additio...

  18. Study of Irradiation Effect onto Uranium silicide Fuel

    International Nuclear Information System (INIS)

    The irradiation effect onto the U3Si-Al and U3Si2-Al dispersion type of fuel element has been studied. The fuel material performs swelling during irradiation due to boehmite (Al2O3(H2O)) formation in which might occurs inside the meat and on the cladding surface, the interaction between the fuel and aluminium matrix that produce U(Al,Si)3 phase, and the formation of fission gas bubble inside the fuel. At a constant fission density, the U3Si-Al fuel swelling is higher than that of U3Si2-Al fuel. The swellings of both fuels increase with the increasing of fission density. The difference of swelling behavior was caused by formation of large bubble gases generated from fission product of U3Si fuel and distributed non-uniformly over all of fuel zone. On the other hand, the U3Si2 fission produced small bubble gases, and those were uniformly distributed. The growth rate of fission gas bubble in the U3Si fuel has shown high diffusivity, transformation into amorph material and thus decrease its mechanical strength

  19. Uranium, resources, production and demand including other nuclear fuel cycle data

    International Nuclear Information System (INIS)

    The uranium reserves exploitable at a cost below 15 dollars/lb U3O8, are 210,000 tonnes. While present uranium production capacities amount to 26,000 tonnes uranium per year, plans have been announced which would increase this capacity to 44,000 tonnes by 1978. Given an appropriate economic climate, annual capacities of 60,000 tonnes and 87,000 tonnes could be attained by 1980 and 1985, respectively, based on presently known reserves. However, in order to maintain or increase such a capacity beyond 1985, substantial additional resources would have to be identified. Present annual demand for natural uranium amounts to 18,000 tonnes and is expected to establish itself at 50,000 tonnes by 1980 and double this figure by 1985. Influences to increase this demand in the medium term could come from shortages in other fuel cycle capacities, i.e. enrichment (higher tails assays) and reprocessing (no uranium and plutonium recycle). However, the analysis of the near term uranium supply and demand situation does not necessarily indicate a prolongation of the current tight uranium market. Concerning the longer term, the experts believe that the steep increase in uranium demand foreseen in the eighties, according to present reactor programmes, with doubling times of the order of 6 to 7 years, will pose formidable problems for the uranium industry. For example, in order to provide reserves sufficient to support the required production rates, annual additions to reserves must almost triple within the next 15 years. Efforts to expand world-wide exploration levels to meet this challenge would be facilitated if a co-ordinated approach were adopted by the nuclear industry as a whole

  20. Commissioning of the fuel reprocessing plant of Rokkasho-Mura; Demarrage de l'usine de retraitement nucleaire de Rokkasho-Mura

    Energy Technology Data Exchange (ETDEWEB)

    Anon

    2006-03-15

    The first Japanese plant dedicated to the separation of uranium and plutonium from spent fuel was commissioned on march 2006 after a 12 year long construction on the Rokkasho-Mura site. This plant, that has been built from the model of the UP3 plant at La-Hague, will enable Japan to process its spent fuel and to produce MOX fuel. A technology transfer agreement was signed in 1987 between Areva and the Japan nuclear fuel company (JNFL). This plant will be on test till mid 2007 when its industrial use is scheduled to begin. (A.C.)

  1. Uranium production and raw materials for the nuclear fuel cycle - Supply and demand, economics, the environment and energy security. Proceedings of an international symposium

    International Nuclear Information System (INIS)

    The IAEA periodically organizes technical meetings and international symposia on all areas of the uranium production cycle. This publication contains the papers and associated material presented at the 2005 international symposium on 'Uranium Production and Raw Materials for the Nuclear Fuel Cycle - Supply and Demand, Economics, the Environment and Energy Security'. The topic areas include uranium supply and demand; uranium geology and deposits; uranium exploration; uranium mining and milling; waste management; and environment and regulation. Each of the 38 papers was indexed individually

  2. Uranium-zirconium hydride TRIGA-LEU fuel

    International Nuclear Information System (INIS)

    The development and testing of TRIGA-LEU fuel with up to 45 wt-% U is described. Topics that are discussed include properties of hydride fuels, the prompt negative temperature coefficient, pulse heating tests, fission product retention, and the limiting design basis parameter and values. General specifications for Er-U-ZrH TRIGA-LEU fuel with 8.5 to 45 wt-% U and an outline of the inspections during manufacture of the fuel are also included. (author). 8 figs, 1 tab

  3. Uptake of Uranium and Other Elements of Concern by Plants Growing on Uranium Mill Tailings Disposal Cells

    Science.gov (United States)

    Joseph, C. N.; Waugh, W.; Glenn, E.

    2015-12-01

    The U.S. Department of Energy (DOE) is responsible for long-term stewardship of disposal cells for uranium mill tailings throughout the United States. Rock-armored disposal cell covers create favorable habitat for deep-rooted plants by reducing soil evaporation, increasing soil water storage, and trapping windblown dust, thereby providing water and nutrients for plant germination and establishment. DOE is studying the tradeoffs of potential detrimental and beneficial effects of plants growing on disposal cell covers to develop a rational and consistent vegetation management policy. Plant roots often extend vertically through disposal cell covers into underlying tailings, therefore, uptake of tailings contaminants and dissemination through animals foraging on stems and leaves is a possible exposure pathway. The literature shows that plant uptake of contaminants in uranium mill tailings occurs, but levels can vary widely depending on plant species, tailings and soil chemistry, and cover soil hydrology. Our empirical field study measured concentrations of uranium, radium, thorium, molybdenum, selenium, manganese, lead, and arsenic in above ground tissues harvested from plants growing on disposal cells near Native American communities in western states that represent a range of climates, cover designs, cover soil types, and vegetation types. For risk screening, contaminant levels in above ground tissues harvested from plants on disposal cells were compared to Maximum Tolerance Levels (MTLs) set for livestock by the National Research Council, and to tissue levels in the same plant species growing in reference areas near disposal cells. Although tailings were covered with uncontaminated soils, for 14 of 46 comparisons, levels of uranium and other contaminants were higher in plants growing on disposal cells compared to reference area plants, indicating possible mobilization of these elements from the tailing into plant tissues. However, with one exception, all plant

  4. Neutronic and thermohydraulic characteristics of a new breeding thorium–uranium mixed SCWR fuel assembly

    International Nuclear Information System (INIS)

    Highlights: • A new Th–U mixed fuel assembly for SCWR has been introduced and investigated. • Neutronic and thermohydraulic characteristics of the new assembly have been studied. • The new fuel assembly satisfies design rules of SCWR. • The introduced fuel assembly can fulfill the sustainable breeding Th–U cycle. • The new fuel assembly also has advantages with respect to lower generation of minor actinides and reactor safety. - Abstract: The exploitation of thorium fuel is a promising way to overcome the pressing problems of nuclear fuel supply, nuclear waste and nuclear proliferation. In this paper, a novel conceptual design of a breeding thorium–uranium (Th–U) mixed fuel assembly in SCWR is proposed, which is aimed to achieve the breeding ratio bigger than 1.0, so as to fulfill the sustainable breeding thorium–uranium cycle. Through the calculations of neutronics and neutronic/thermohydraulic (N–T) coupling, the results indicate that the introduced conceptual design of a breeding Th–U mixed fuel assembly in SCWR satisfies design rules of SCWR, with considerable advantages with respect to breeding performance, lower minor actinide generation and reactor safety

  5. Considerations in recycling used natural uranium fuel from CANDU reactors in Canada

    International Nuclear Information System (INIS)

    This paper identifies the key factors that would affect the recycling of used natural uranium (NU) fuel from CANDU reactors which are in operation in Canada and in several other countries. There has been little analysis of those considerations over the past 25 years and this paper provides a framework for such analysis. In particular, the large energy potential of the plutonium in used CANDU NU fuel provides a driver for consideration of used-fuel recycling. There would be a long lead-time (at least 30 years) and a large investment required for establishing the infrastructure for used-fuel recycling. While this paper does not promote the recycling of used CANDU NU fuel in Canadian CANDU reactors, it does suggest that it is timely to start the analysis and to consider the key factors or circumstances that warrant the recycling of used CANDU NU fuel. (author)

  6. Chemical states of fission products in irradiated uranium-plutonium mixed oxide fuel

    International Nuclear Information System (INIS)

    The chemical states of fission products (FPs) in irradiated uranium-plutonium mixed oxide (MOX) fuel for the light water reactor (LWR) were estimated by thermodynamic equilibrium calculations on system of fuel and FPs by using ChemSage program. A stoichiometric MOX containing 6.1 wt. percent PuO2 was taken as a loading fuel. The variation of chemical states of FPs was calculated as a function of oxygen potential. Some pieces of information obtained by the calculation were compared with the results of the post-irradiation examination (PIE) of UO2 fuel. It was confirmed that the multicomponent and multiphase thermodynamic equilibrium calculation between fuel and FPs system was an effective tool for understanding the behavior of FPs in fuel. (author)

  7. Fabrication of uranium silicide dispersion fuel by atomization for research reactor

    International Nuclear Information System (INIS)

    Atomizing technology has been developed to eliminate the difficulties in comminution of the tough U3Si and to take advantage of the spherical shape and the rapid solidification. The comparison between the conventional dispersion fuel with comminuted powder and the newly developed fuel with atomized powder has been made. As a result, the processes, powdering uranium silicide and heat treatment to U3Si, become simplified. The extruding pressure of blended powder with atomized powder was lower than that of blended powder with comminuted powder. The elongation of the atomization processed fuel meat was much higher than that of comminution processed fuel meats. It appears that the loading density of U3Si in fuel meat can be increased by using atomized U3Si powder. The thermal conductivity and the thermal compatibility of fuel meat have been investigated and found to be much improved due to the spherical shape of atomized powder. (author)

  8. Specific features of the WWER Uranium-Gadolinium fuel behavior at BOL

    International Nuclear Information System (INIS)

    The calculated-experimental analysis of the WWER fuel behavior with 5%wt of gadolinium oxide at the beginning of life (BOL) is presented. The results are based on the data on fuel centerline temperature measurements, gas media pressure inside the cladding and fuel elongation obtained during irradiation of the test fuel rods in HBWR (Halden). Computer analysis of experimental data is performed with TOPRA-2, version 2 code. It is shown that specific features of the uranium-gadolinium fuel behavior at the early of life is due to presence of burnable absorber influencing the average linear heat rating, radial power distribution and lower thermal conductivity. In particular, the analysis of “late” relocation effect on the maximum Gd fuel temperature is presented. (authors)

  9. Promotion of uranium enrichment business

    International Nuclear Information System (INIS)

    The Committee on Nuclear Power has studied on the basic nuclear power policy, establishing its five subcommittees, entrusted by the Ministry of Nternational Trade and Industry. The results of examination by the subcommittee on uranium enrichment business are given along with a report in this connection by the Committee. In order to establish the nuclear fuel cycle, the aspect of uranium enrichment is essential. The uranium enrichment by centrifugal process has proceeded steadily in Power Reactor and Nuclear Fuel Development Corporation. The following matters are described: the need for domestic uranium enrichment, the outlook for overseas enrichment services and the schedule for establishing domestic enrichment business, the current state of technology development, the position of the prototype enrichment plant, the course to be taken to establish enrichment business the main organization operating the prototype and commercial plants, the system of supplying centrifuges, the domestic conversion of natural uranium the subsidies for uranium enrichment business. (J.P.N.)

  10. The behavior of uranium in the soil/plant system with special consideration of the uranium input by mineral phosphorus fertilizer

    International Nuclear Information System (INIS)

    The fate of uranium in the environment and, consequently, its hazard potential for human beings is still discussed controversially in the scientific literature. Mineral phosphorous fertilizer can contain uranium as impurity, so that their application can cause an additional input of uranium into agricultural environments. It is still unclear whether and to what extent fertilizer-derived uranium can enter the human food chain by the consumption of contaminated waters or vegetable crop products. The mobility and availability of uranium in the agricultural ecosystem is mainly determined by its behavior in the pedosphere. Due to interactions with organic and inorganic components, the pedosphere is an effective storage and filter system for pollutants and thus plays an important role for the fate of uranium in the environment. In order to improve the assessment of the hazard potential, the present study investigates the behavior of uranium in the soil/plant-system with a focus on the uranium input by mineral phosphorous fertilizer. The specific objectives were (A) to investigate the general distribution of uranium in soils, (B) to determine the effect of CaCO3 on the sorption behavior of uranium and to quantify the effects of (C - D) varying substrate properties and (E) the application of phosphorus fertilizers on the uranium uptake by ryegrass. The results of these experiments imply that the use of mineral phosphorous fertilizers does not pose an acute risk within the meaning of consumer protection. The studied soils predominantly had a high to very high sorption capability for uranium. At the same time, a small soil-to-plant-transfer of uranium was determined, where the majority of uranium accumulated in/to the plant roots. The availability of uranium in soils and its uptake by plants can thus be classified as generally low. Furthermore, some soil parameters were identified which seem to favor a higher uranium-availability. This study found that very high and very low

  11. Conversion and standardization of university reactor fuels using low-enrichment uranium: Plans and schedules

    Energy Technology Data Exchange (ETDEWEB)

    Young, H.H.; Brown, K.R.; Matos, J.E.

    1986-01-01

    The highly-enriched uranium (HEU) fuel used in twenty United States university reactors can be viewed as contributing to the risk of theft or diversion of weapons-useable material. To minimize this risk, the US Nuclear Regulatory Commission issued its final rule on ''Limiting the Use of Highly Enriched Uranium in Domestically Licensed Research and Test Reactors,'' in February 1986. This paper describes the plans and schedules developed by the US Department of Energy to coordinate an orderly transition from HEU to LEU fuel in most of these reactors. An important element in the planning process has been the desire to standardize the LEU fuels used in US university reactors and to enhance the performance and utilization of a number of these reactors. The program is estimated to cost about $10 million and to last about five years.

  12. The contamination of the low enriched uranium fuel with oxygen during the manufacturing process

    International Nuclear Information System (INIS)

    The manufacturing of TRIGA fuel rods with low enriched uranium (≤ 20% 235U) follows in principle the same route as for high-enriched uranium (93% 235U). The contamination with chemical elements during the fabrication process deteriorates the fuel properties and its quality required by the run in the reactor. From the standpoint of the phase relations, the most important impurity is the oxygen. The oxygen concentration influences the kinetics of the zirconium hydriding process. If during the hydriding process contamination with oxygen occurs a decrease of the hydriding rate will take place. This paper presents the aspects of the TRIGA fuel contamination with oxygen during manufacturing process and ways to reduce it. The permanent control of the oxygen concentration in the working zone avoids the accidental contamination. (authors)

  13. Development of uranium-silicide and U-Mo alloy fuels by centrifugal atomization

    International Nuclear Information System (INIS)

    U3Si and U3Si2 powders for application dispersant for research reactor dispersion fuel elements are produced by atomizing alloy melt instead of comminuting. Many benefits are introduced by applying the atomization technique: reduction of the process, increase of the thermal conductivity, decrease of the thermal reaction with aluminium due to the spherical shape of particle, and increase of the uranium loading. Also, in order to improve the uranium loading of fuel, high density U-Mo powder is prepared by centrifugal atomization. U-Mo powder has spherical shape, narrow size distributions, high density and fine grain structure with isotropic γ-U phase. U-Mo fuel meats, especially U-10wt.%Mo, show good thermal compatibility with Al matrix and maintain microstructure stability. (author). 21 refs, 12 figs, 2 tabs

  14. Characteristics and performance of the plant for refabrication of vibrocompacted fuel elements for the BOR-60 reactor

    International Nuclear Information System (INIS)

    The paper deals with the characterization of an automated pilot plant for the refabrication of fuel elements for the Soviet experimental BOR-60 fast breeder reactor. In addition, it describes the results, experience, and problems encountered in respect of the technological process of fuel element manufacture and quality assurance. The electrodynamic vibration procedure for densification of the uranium-plutonium mixed oxide is given special attention. (orig.)

  15. Analysis of Uranium and Thorium in Radioactive Wastes from Nuclear Fuel Cycle Process

    International Nuclear Information System (INIS)

    The assessment of analysis method for uranium and thorium in radioactive wastes generated from nuclear fuel cycle process have been carried out. The uranium and thorium analysis methods in the assessment are consist of Titrimetry, UV-VIS Spectrophotometry, Fluorimetry, HPLC, Polarography, Emission Spectrograph, XRF, AAS, Alpha Spectrometry and Mass Spectrometry methods. From the assessment can be concluded that the analysis methods of uranium and thorium content in radioactive waste for low concentration level using UV-VIS Spectrometry is better than Titrimetry method. While for very low concentration level in part per billion (ppb) can be used by Neutron Activation Analysis (NAA), Alpha Spectrometry and Mass Spectrometry. Laser Fluorimetry is the best method of uranium analysis for very low concentration level. Alpha Spectrometry and ICP-MS (Inductively Coupled Plasma Mass Spectrometry) methods for isotopic analysis are favourable in the precision and accuracy aspects. Comparison of the ICP-MS and Alpha Spectrometry methods shows that the both of methods have capability to determining of uranium and thorium isotopes content in the waste samples with results comparable very well, but the time of its analysis using ICP-MS method is faster than the Alpha Spectrometry, and also the cost of analysis for ICP-MS method is cheaper. NAA method can also be used to analyze the uranium and thorium isotopes, but this method needs the reactor facility and also the time of its analysis is very long. (author)

  16. NEUTRONICS STUDIES OF URANIUM-BASED FULLY CERAMIC MICRO-ENCAPSULATED FUEL FOR PWRs

    Energy Technology Data Exchange (ETDEWEB)

    George, Nathan M [ORNL; Maldonado, G Ivan [ORNL; Terrani, Kurt A [ORNL; Gehin, Jess C [ORNL; Godfrey, Andrew T [ORNL

    2012-01-01

    This study evaluates the core neutronics and fuel cycle characteristics that result from employing uranium-based fully ceramic micro-encapsulated (FCM) fuel in a pressurized water reactor (PWR). Specific PWR bundle designs with FCM fuel have been developed, which by virtue of their TRISO particle based elements, are expected to safely reach higher fuel burnups while also increasing the tolerance to fuel failures. The SCALE 6.1 code package, developed and maintained at ORNL, was the primary software employed to model these designs. Analysis was performed using the SCALE double-heterogeneous (DH) fuel modeling capabilities. For cases evaluated with the NESTLE full-core three-dimensional nodal simulator, because the feature to perform DH lattice physics branches with the SCALE/TRITON sequence is not yet available, the Reactivity-Equivalent Physical Transformation (RPT) method was used as workaround to support the full core analyses. As part of the fuel assembly design evaluations, fresh feed lattices were modeled to analyze the within-assembly pin power peaking. Also, a color-set array of assemblies was constructed to evaluate power peaking and power sharing between a once-burned and a fresh feed assembly. In addition, a parametric study was performed by varying the various TRISO particle design features; such as kernel diameter, coating layer thicknesses, and packing fractions. Also, other features such as the selection of matrix material (SiC, Zirconium) and fuel rod dimensions were perturbed. After evaluating different uranium-based fuels, the higher physical density of uranium mononitride (UN) proved to be favorable, as the parametric studies showed that the FCM particle fuel design will need roughly 12% additional fissile material in comparison to that of a standard UO2 rod in order to match the lifetime of an 18-month PWR cycle. Neutronically, the FCM fuel designs evaluated maintain acceptable design features in the areas of fuel lifetime, temperature

  17. The Shipment of Russian-Origin Highly Enriched Uranium Research Reactor Spent Nuclear Fuel from Belarus

    Energy Technology Data Exchange (ETDEWEB)

    Sikorin, S.N.; Polazau, S.A.; Hryharovich, T.K. [Joint Institute for Power and Nuclear Research ' Sosny' , Academik Krasin Street, Minsk (Belarus); Bolshinsky, I. [Idaho National Laboratory, N. Fremont Avenue Idaho Falls, Idaho (United States); Thomas, J.E. [Savannah River National Laboratory, Aiken, South Carolina (United States)

    2011-07-01

    In October 2010, the Global Threat Reduction Initiative and the Joint Institute for Power and Nuclear Research - 'Sosny' of the National Academy of Sciences of the Republic of Belarus completed a shipment that returned 43 kilograms of Russian-origin highly enriched uranium (HEU) spent nuclear fuel to the Russian Federation. The spent fuel was legacy material, discharged from the two decommissioned reactors, the Pamir-630D mobile reactor and the IRT-M research reactor. This shipment marked the complete removal of all HEU spent nuclear fuel from Belarus. This paper discusses the planning, preparations, and coordination required to complete this important international shipment successfully. (author)

  18. The Shipment of Russian-Origin Highly Enriched Uranium Research Reactor Spent Nuclear Fuel from Belarus

    International Nuclear Information System (INIS)

    In October 2010, the Global Threat Reduction Initiative and the Joint Institute for Power and Nuclear Research - 'Sosny' of the National Academy of Sciences of the Republic of Belarus completed a shipment that returned 43 kilograms of Russian-origin highly enriched uranium (HEU) spent nuclear fuel to the Russian Federation. The spent fuel was legacy material, discharged from the two decommissioned reactors, the Pamir-630D mobile reactor and the IRT-M research reactor. This shipment marked the complete removal of all HEU spent nuclear fuel from Belarus. This paper discusses the planning, preparations, and coordination required to complete this important international shipment successfully. (author)

  19. Soil and plant selenium at a reclaimed uranium mine.

    Science.gov (United States)

    Sharmasarkar, Shankar; Vance, George F

    2002-01-01

    Selenium (Se) associated with reclaimed uranium (U) mine lands may result in increased food chain transfer and water contamination. To assess post-reclamation bioavailability of Se at a U mine site in southeastern Wyoming, we studied soil Se distribution, dissolution, speciation, and sorption characteristics and plant Se accumulation. Phosphate-extractable soil Se exceeded the critical limit of 0.5 mg/kg in all the samples, whereas total soil Se ranged from a low (0.6 mg/kg) to an extremely high (26 mg/kg) value. Selenite was the dominant species in phosphate and ammonium bicarbonate-diethylenetriamine pentaacetic acid (AB-DTPA) extracts, whereas selenate was the major Se species in hot water extracts. Extractable soil Se concentrations were in the order of KH2PO4 > AB-DTPA > hot water > saturated paste. The soils were undersaturated with respect to various Se solid phases, albeit with high levels of extractable Se surpassing the critical limit. Calcium and Mg minerals were the potential primary solids controlling Se dissolution, with dissolved organic carbon in the equilibrium solutions resulting in enhanced Se availability. Adsorption was a significant (r2 = 0.76-0.99 at P phytoremediation, or the palatable forage species may be used as animal feed supplements in Se-deficient areas. PMID:12371169

  20. Leaching of cesium and uranium from spent PWR fuel in the gel-state clays

    International Nuclear Information System (INIS)

    The amounts of cesium and uranium released from crushed spent PWR fuel in the gel-state clays with a few ml of supernatant at hot cell temperature under Ar-atmosphere have been measured. The fractions of cesium dissolved from the fuel for 873 days were 0.29 and 0.25% in Boom clay/Boom-clay water and Ca-bentonite/synthetic granitic groundwater, respectively. These cesium fractions were very close to the gap inventory of cesium, which was determined to be around 0.30% in the previous experiment. The fraction of uranium released up to 193 days in the Boom clay media was 0.011% and this fraction has been retained until 873 days. Such this phenomenon was also obtained in the Ca-bentonite media even though the released fraction was higher than that in Boom clay. The increase of less than 0.001% in the dissolved uranium fraction between 193 and 873 days suggests that the long-term leach rate of uranium from spent fuel would be much less than 24 μg·m-2·day-1. (author)

  1. 77 FR 14010 - Rocky Ridge Wind Project, LLC, Blackwell Wind, LLC, CPV Cimarron Renewable Energy Company, LLC...

    Science.gov (United States)

    2012-03-08

    ... Cimarron Renewable Energy Company, LLC, Minco Wind Interconnection Services, LLC, Shiloh III Lessee, LLC, California Ridge Wind Energy LLC, Perrin Ranch Wind, LLC, Erie Wind, LLC: Notice of Effectiveness of Exempt... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY...

  2. 75 FR 16098 - Southern Turner Cimarron I, LLC; Supplemental Notice That Initial Market-Based Rate Filing...

    Science.gov (United States)

    2010-03-31

    ... From the Federal Register Online via the Government Publishing Office ] DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Southern Turner Cimarron I, LLC; Supplemental Notice That Initial Market-Based Rate Filing Includes Request for Blanket Section 204 Authorization March 24, 2010. This is a supplemental notice in the...

  3. Material surveillance and verification programme at a uranium enriching plant

    International Nuclear Information System (INIS)

    A licence for a nuclear facility in the United States of America is approved only after a licencee demonstrates by procedure or practice that an adequate material control system exists. A licence can specify acceptable material control practices. Therefore, processors in the United States receiving uranium hexafluoride (UF6) from a United States Government-owned enriching plant can accept shipper's values for nuclear material accounting purposes if: (1) there is surveillance during withdrawal of the UF6; (2) an independent sample is obtained; and (3) certain measurement verification is subsequently performed by the receiver or the receiver's agent. Because of the high equipment and operating costs, essentially all UF6 processors have adopted a surveillance and verification programme. A resident observer is employed to perform surveillance, obtain samples, and make the shipping cylinders tamper-safe. Samples are analysed by the receiver or by an independent laboratory. The observer determines by surveillance that withdrawals, or transfers of material, weighings, and sampling are accomplished in accordance with accepted procedures. Surveillance of the withdrawals includes observing the transfer of UF6 from the enriching plant cylinder to the shipping cylinder(s) and the withdrawal of samples. In addition, it includes observing the weighing of all cylinders associated with a sample lot of UF6. Following the surveillance of withdrawals, weighings, and sampling, the cylinders are made tamper-safe by the application of tamper-indicating devices. Statistics for the verification programme have shown shipper and receiver measurements to be within the limits acceptable for adequate material control. (author)

  4. Development of very-high-density low-enriched uranium fuels

    International Nuclear Information System (INIS)

    The RERTR (=Reduced Enrichment for Research and Test Reactors) program has begun an aggressive effort to develop dispersion fuels for research and test reactors with uranium densities of 8 to 9 g U/cm3, based on the use of γ-stabilized uranium alloys. Fabrication development teams and facilities are being put into place, and preparations for the first irradiation test are in progress. The first screening irradiations are expected to begin in late April 1997 and the first results should be available by the end of 1997. Discussions with potential international partners in fabrication development and irradiation testing have begun. (author)

  5. Effect of high-density fuel loading on criticality of low enriched uranium fueled material test research reactors

    International Nuclear Information System (INIS)

    The effect of high-density fuel loading on the criticality of low enriched uranium fueled material test reactors was studied using the standard reactor physics simulation codes WIMS-D/4 and CITATION. Three strategies were considered to increase the fuel loading per plate: (1) by substituting the high-density fuel in place of low-density fuel keeping meat thickness and water channel width constant, (2) by substituting the high-density fuel in place of low-density fuel keeping fuel meat thickness fixed and optimizing the water channel width between the fuel plates and (3) by increasing the fuel meat thickness of fixed density fuel and optimizing the water channel width between the fuel plates. The fuel requirements for critical and first high power cores were determined in each case for higher fuel loadings per plate. It has been found that in the first case, core volume reduces with increasing fuel loadings per plate but requirement of fuel also increases. In the second and third case, core volume as well as fuel requirement decreases with increasing fuel loadings per plate. However in the second case, core volume reduces more rapidly than in case 3 with increasing fuel loadings per plate. Employing standard computer code PARET, steady state thermal hydraulic analysis of all these cores was performed. The thermal hydraulic analysis reveals that cores with higher densities and fixed water channel width are better from thermal hydraulic point of view and have fuel and clad temperatures within the acceptable limits. But the core with higher densities and optimum water channel width is a better choice in terms of core compaction, less 235U loading and higher neutron fluxes. Finally, the core was compacted in three steps to exploit the benefits of both types of cores. The strategy resulted in 36% reduction in the core volume, 50% increase in thermal neutron flux for irradiation and isotope production and a slight reduction in 235U loading. All this was achieved with

  6. RU-43 a new uranium fuel bundle design for using in CANDU type reactors

    International Nuclear Information System (INIS)

    A unique feature of the CANDU reactor design is its ability to use alternative fuel cycles other than natural uranium (NU), without requiring major modifications to the basic reactor design. These alternative fuel cycles, which are known as advanced fuel cycles, utilize a variety of fissile materials, including Slightly Enriched Uranium (SEU) from enrichment facilities, and Recovered Uranium (RU) obtained from the reprocessing of the spent fuel of light-water reactors (LWR). A fissile content in the RU of 0.9 to 1.0 % makes it impossible for reuse in an LWR without re-enrichment, but CANDU reactors have a sufficient high neutron economy to use RU as fuel. RU from spent LWR fuel can be considered as a lower cost source of enrichment at the optimal enrichment level for CANDU fuel pellets. In Europe the feedstock of RU is approaching thousands tones and would provide sufficient fuel for hundreds CANDU-6 reactors years of operation. The use of RU fuel offers significant benefits to CANDU reactor operators. RU fuels improve fuel cycle economics by increasing the fuel burnup, which enables large cost reductions in fuel consumption and in spent fuel disposal. RU fuel offers enhanced operating margins that can be applied to increase reactor power. These benefits can be realized using existing fuel production technologies and practices, and with almost negligible changes to fuel receipt and handling procedures at the reactor. The application of RU fuel could be an important element in NPP Cernavoda from Romania. For this reason the Institute for Nuclear Research (INR), Pitesti has started a research programme aiming to develop a new fuel bundle RU-43 for extended burnup operation. The current version of the design is the result of a long process of analyses and improvements, in which successive preliminary design versions have been evaluated. The most relevant calculations performed on this fuel element design version are presented. Also, the stages of an experimental

  7. Initiation of depleted uranium oxide and spent fuel testing for the spent fuel sabotage aerosol ratio program

    International Nuclear Information System (INIS)

    We provide a detailed overview of an ongoing, multinational test program that is developing aerosol data for some spent fuel sabotage scenarios on spent fuel transport and storage casks. Experiments are being performed to quantify the aerosolized materials plus volatilized fission products generated from actual spent fuel and surrogate material test rods, due to impact by a high energy density device, HEDD. The program participants in the U.S. plus Germany, France, and the U.K., part of the international Working Group for Sabotage Concerns of Transport and Storage Casks, WGSTSC have strongly supported and coordinated this research program. Sandia National Laboratories, SNL, has the lead role for conducting this research program; test program support is provided by both the U.S. Department of Energy and Nuclear Regulatory Commission. WGSTSC partners need this research to better understand potential radiological impacts from sabotage of nuclear material shipments and storage casks, and to support subsequent risk assessments, modeling, and preventative measures. We provide a summary of the overall, multi-phase test design and a description of all explosive containment and aerosol collection test components used. We focus on the recently initiated tests on ''surrogate'' spent fuel, unirradiated depleted uranium oxide, and forthcoming actual spent fuel tests. The depleted uranium oxide test rodlets were prepared by the Institut de Radioprotection et de Surete Nucleaire, in France. These surrogate test rodlets closely match the diameter of the test rodlets of actual spent fuel from the H.B. Robinson reactor (high burnup PWR fuel) and the Surry reactor (lower, medium burnup PWR fuel), generated from U.S. reactors. The characterization of the spent fuels and fabrication into short, pressurized rodlets has been performed by Argonne National Laboratory, for testing at SNL. The ratio of the aerosol and respirable particles released from HEDD-impacted spent fuel to the aerosol

  8. Uranium production and raw materials for the nuclear fuel cycle - Supply and demand, economics, the environment and energy security. Proceedings of an international symposium

    International Nuclear Information System (INIS)

    The companion CD contains other papers presented at the 2005 international symposium on 'Uranium Production and Raw Materials for the Nuclear Fuel Cycle - Supply and Demand, Economics, the Environment and Energy Security'. The topic areas include uranium supply and demand; uranium geology and deposits; uranium production; waste management; and environment and regulation. Each of the 12 papers was indexed individually

  9. Sulfur balance in biomass-fueled plants

    International Nuclear Information System (INIS)

    The aim of this project has been to establish a standard deduction for sulphur retained in the ash. This is accomplished by establishing sulphur balances for biomass plants in order to document the in- and outgoing flows. The ingoing flow is the sulphur in the input fuel while the outgoing flows are different ash fractions and sulphur dioxide measured in the stack. Four balances have been established for straw fired units, three balances for wood chip fired units, and two balances for wood pellet fired units. Two previous projects provide further data on both straw and wood fired units. The main conclusions and recommendations are: For wood pellets the sulphur tax should be removed as the sulphur content in the pellets is extremely low and the emitted fraction very small. For pellets manufactured with a binder containing sulphur, the taxation should continue but with a standard deduction of 60 to 70%. Also, the rate should be reduced as the sulphur content in pellets produced with a binder containing sulphur is lower than the estimated 0,2% of the fuel. Statistics indicate that 0,1% reflects the true sulphur content in these pellets; For wood chips the tax should be removed as the sulphur content based on the fuel is considerably lower than the limit in the law (0,034% versus 0,05%). Furthermore, the emission from these plants are only between 20 and 32%. It is recommended that the plants keep the ph-value in the scrubber water above 7 as it is believed that this improves the absorption of SO2 greatly; For straw the tax should remain, but a standard deduction of 35-40% should be made. Technologies for improving the sulphur retentions should be developed. This could be scrubbers as they are very efficient towards removing especially sulphur in the form of SO2, which is by far the largest source of sulphur emission from straw fired plants. (au) 11 refs

  10. Uranium enrichment

    International Nuclear Information System (INIS)

    Canada is the world's largest producer and exporter of uranium, most of which is enriched elsewhere for use as fuel in LWRs. The feasibility of a Canadian uranium-enrichment enterprise is therefore a perennial question. Recent developments in uranium-enrichment technology, and their likely impacts on separative work supply and demand, suggest an opportunity window for Canadian entry into this international market. The Canadian opportunity results from three particular impacts of the new technologies: 1) the bulk of the world's uranium-enrichment capacity is in gaseous diffusion plants which, because of their large requirements for electricity (more than 2000 kW·h per SWU), are vulnerable to competition from the new processes; 2) the decline in enrichment costs increases the economic incentive for the use of slightly-enriched uranium (SEU) fuel in CANDU reactors, thus creating a potential Canadian market; and 3) the new processes allow economic operation on a much smaller scale, which drastically reduces the investment required for market entry and is comparable with the potential Canadian SEU requirement. The opportunity is not open-ended. By the end of the century the enrichment supply industry will have adapted to the new processes and long-term customer/supplier relationships will have been established. In order to seize the opportunity, Canada must become a credible supplier during this century

  11. Irradiation program of slightly enriched fuel elements at the Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    An irradiation program of fuel elements with slightly enriched uranium is implemented, tending to the homogenization of core at Atucha I nuclear power plant. The main benefits of the enrichment program are: a) to extend the average discharge burnup of fuel elements, reducing the number of elements used to generate the same amount of energy. This implies a smaller annual consumption of elements and consequently the reduction of transport and replacement operations and of the storage pool systems as well as that of radioactive wastes; b) the saving of uranium and structural materials (Zircaloy and others). In the initial stage of program an homogeneous core enrichment of 0.85% by weight of U-235 is anticipated. The average discharge burnup of fuel elements, as estimated by previous studies, is approximately 11.6 MW d/kg U. The annual consumption of fuel elements is reduced from 396 of natural uranium to 205, with a load factor of 0.85. It is intended to reach the next equilibrium steps with an enrichment of 1.00 and 1.20% in U-235. (Author)

  12. Containment and storage of uranium hexafluoride at US Department of Energy uranium enrichment plants

    International Nuclear Information System (INIS)

    Isotopically depleted UF6 (uranium hexafluoride) accumulates at a rate five to ten times greater than the enriched product and is stored in steel vessels at the enrichment plant sites. There are approximately 55,000 large cylinders now in storage at Paducah, Kentucky; Portsmouth, Ohio; and Oak Ridge, Tennessee. Most of them contain a nominal 14 tons of depleted UF6. Some of these cylinders have been in the unprotected outdoor storage environment for periods approaching 40 years. Storage experience, supplemented by limited corrosion data, suggests a service life of about 70 years under optimum conditions for the 48-in. diameter, 5/16-in.-wall pressure vessels (100 psi working pressure), using a conservative industry-established 1/4-in.-wall thickness as the service limit. In the past few years, however, factors other than atmospheric corrosion have become apparent that adversely affect the serviceability of small numbers of the storage containers and that indicate the need for a managed program to ensure maintenance ofcontainment integrity for all the cylinders in storage. The program includes periodic visual inspections of cylinders and storage yards with documentation for comparison with other inspections, a group of corrosion test programs to permit cylinder life forecasts, and identification of (and scheduling for remedial action) situations in which defects, due to handling damage or accelerated corrosion, can seriously shorten the storage life or compromise the containment integrity of individual cylinders. The program also includes rupture testing to assess the effects of certain classes of damage on overall cylinder strength, aswell as ongoing reviews of specifications, procedures, practices, and inspection results to effect improvements in handling safety, containment integrity, and storage life

  13. Nuclear power and uranium fuel: requirements in the Western United States

    International Nuclear Information System (INIS)

    The results are presented of a survey of nuclear power and associated uranium fuel cycle requirements in the Western Region recently conducted by the Western Interstate Nuclear Board (WINB). The survey covers the period 1975--1990 for electric energy and nuclear power except for uranium and other fuel cycle requirements which is limited to the period 1975--1985. This shorter time span coverage for fuel cycle requirements was chosen due to many uncertainties in future federal policy in such areas as uranium enrichment, plutonium utilization in light water reactors, spent fuel reprocessing, and high-level radioactive waste management. Geographically, the survey covered the twelve western states served by WINB. Also contained is information on projected national and western regional trends in electric energy demand and supply developed by the U. S. Energy Research and Development Administration (ERDA), the Western Systems Coordinating Council (WSCC), and other organizations. This information is presented as background material for assessing nuclear power in the West and nation in its proper perspective as one of several major energy sources available for general application during the period 1975--1990

  14. How can Korea secure uranium enrichment and spent fuel reprocessing rights?

    International Nuclear Information System (INIS)

    South Korea is heavily dependent on energy resources from other countries and nuclear energy accounts for 31% of Korea's electric power generation as a major energy. However, Korea has many limitations in uranium enrichment and spent fuel reprocessing under the current Korea-U.S. nuclear agreement, although they are economically and politically important to Korea due to a significant problems in nuclear fuel storages. Therefore, in this paper, we first examine those example countries – Japan, Vietnam, and Iran – that have made nuclear agreements with the U.S. or have changed their agreements to allow the enrichment of uranium and the reprocessing of spent fuel. Then, we analyze those countries' nuclear energy policies and review their strategic repositioning in the relationship with the U.S. We find that a strong political stance for peaceful usage of nuclear energy including the legislation of nuclear laws as was the case of Japan. In addition, it is important for Korea to acquire advanced technological capability such as sodium-cooled fast reactor (SFR) because SFR technologies require plutonium to be used as fuel rather than uranium-235. In addition, Korea needs to leverage its position in nuclear agreement between China and the U.S. as was the case of Vietnam

  15. Post irradiation examinations of 87F-2A capsule containing uranium-plutonium mixed carbide fuels

    International Nuclear Information System (INIS)

    One fuel pin filled with hyperstoichiometric uranium-plutonium mixed carbide pellets was encapsulated in 87F-2A and irradiated in JMTR up to 4.4 %FIMA at an average linear power of 60 kW/m. The capsule cooled for ∼4 months was transported to Reactor Fuel Examination Facility and subjected to non-destructive and destructive post irradiation examinations. It was found from the radial cross section of fuel pin that the helium gap between the pellets and the cladding tube was completely closed. Compared with the results obtained so far, very low open porosity and fission gas release rate as well as mild restructuring was observed owing to the adoption of thermally stable pellets. The diametric increase of fuel pin reached ∼0.06mm at the position of maximum reading, although it might not affect the fuel performance itself. The inner surface of cladding tube did not show signs of carburization. (author)

  16. Recycling of plutonium and uranium in water reactor fuel. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    The Technical Committee Meeting on Recycling of Plutonium and Uranium in Water Reactor Fuel was recommended by the International Working Group on Fuel Performance and Technology (IWGFPT). Its aim was to obtain an overall picture of MOX fabrication capacity and technology, actual performance of this kind of fuel, and ways explored to dispose of the weapons grade plutonium. The subject of this meeting had been reviewed by the International Atomic Energy Agency every 5 to 6 years and for the first time the problem of weapons grade plutonium disposal was included. The papers presented provide a summary of experience on MOX fuel and ongoing research in this field in the participating countries. The meeting was hosted by British Nuclear Fuels plc, at Newby Bridge, United Kingdom, from 3 to 7 July 1995. Fifty-six participants from twelve countries or international organizations took part. Refs, figs, tabs

  17. Improvement of the homogeneity of atomized particles dispersed in high uranium density research reactor fuels

    International Nuclear Information System (INIS)

    A study on improving the homogeneous dispersion of atomized spherical particles in fuel meats has been performed in connection with the development of high uranium density fuel. In comparing various mixing methods, the better homogeneity of the mixture could be obtained as in order of Spex mill, V-shape tumbler mixer, and off-axis rotating drum mixer. The Spex mill mixer required some laborious work because of its small capacity per batch. Trough optimizing the rotating speed parameter for the V-shape tumbler mixer, almost the same homogeneity as with the Spex mill could be obtained. The homogeneity of the extruded fuel meats appeared to improve through extrusion. All extruded fuel meats with U3 Si powder of 50-volume % had fairly smooth surfaces. The homogeneity of fuel meats by V-shaped tumbler mixer revealed to be fairly good on micrographs. (author)

  18. Fission gas release of uranium-plutonium mixed nitride and carbide fuels

    International Nuclear Information System (INIS)

    Uranium-plutonium mixed nitride and carbide for advanced fast reactor fuels were irradiated at JRR-2, and the fission gas release from these fuels were determined. It is confirmed from the irradiation tests that the application of the cold fuel concept to these fuels on the basis of their advantageous thermal properties may realize low fission gas release. Furthermore, the introduction of the thermal stable pellets with dense matrix and relatively large pores can lower the fission gas release to a few percent up to the burnup of 5.5% FIMA. In spite of the retention of fission gas release in the thermal stable pellets, no significant enhancement of the fuel-clad mechanical interaction was observed in the examined range of burnup. It is also suggested that the open porosity would strongly influence the fission gas release from nitride and carbide. (author). 18 refs, 6 figs, 6 tabs

  19. Are world uranium resources sufficient to fuel global growth in nuclear generating capacity?

    International Nuclear Information System (INIS)

    Increased uranium prices since 2003 have produced more activity in the sector than the previous 20 years. Nuclear reactor construction is proceeding in some countries, ambitious expansion plans have been announced in others and several, particularly in the developing world, are considering introducing nuclear power as a means of meeting rising electricity demand without increasing greenhouse gas emissions. Others have recently decided to either withdraw from the use of nuclear power or not proceed with development plans following the accident at the Fukushima Dai-ichi nuclear power plant in Japan in March 2011. Since the mid-1960, the OECD Nuclear Energy Agency and the International Atomic Energy Agency have jointly prepared a comprehensive update of global uranium resources, production and demand (commonly known as the 'Red Book'. The Red Book is based on government responses to a questionnaire that requests information on uranium exploration and mine development activity, resources and plans for nuclear development to 2035. This presentation provides an overview of the global situation based on the recently published 2011 edition. It features a compilation of global uranium resources, projected mine development and production capability in all the countries currently producing uranium or with plans to do so in the near future. This is compared to updated, post-Fukushima demand projections, reflecting nuclear phase-out plans announced in some countries and ambitious expansion plans of others. The 2011 Red Book shows that currently defined uranium resources are sufficient to meet high case projections of nuclear power development to 2035. (authors)

  20. Fuel breaks affect nonnative species abundance in Californian plant communities

    Science.gov (United States)

    Merriam, K.E.; Keeley, J.E.; Beyers, J.L.

    2006-01-01

    We evaluated the abundance of nonnative plants on fuel breaks and in adjacent untreated areas to determine if fuel treatments promote the invasion of nonnative plant species. Understanding the relationship between fuel treatments and nonnative plants is becoming increasingly important as federal and state agencies are currently implementing large fuel treatment programs throughout the United States to reduce the threat of wildland fire. Our study included 24 fuel breaks located across the State of California. We found that nonnative plant abundance was over 200% higher on fuel breaks than in adjacent wildland areas. Relative nonnative cover was greater on fuel breaks constructed by bulldozers (28%) than on fuel breaks constructed by other methods (7%). Canopy cover, litter cover, and duff depth also were significantly lower on fuel breaks constructed by bulldozers, and these fuel breaks had significantly more exposed bare ground than other types of fuel breaks. There was a significant decline in relative nonnative cover with increasing distance from the fuel break, particularly in areas that had experienced more numerous fires during the past 50 years, and in areas that had been grazed. These data suggest that fuel breaks could provide establishment sites for nonnative plants, and that nonnatives may invade surrounding areas, especially after disturbances such as fire or grazing. Fuel break construction and maintenance methods that leave some overstory canopy and minimize exposure of bare ground may be less likely to promote nonnative plants. ?? 2006 by the Ecological Society of America.

  1. Digital data sets that describe aquifer characteristics of the alluvial and terrace deposits along the Cimarron River from Freedom to Guthrie in northwestern Oklahoma

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — This data set consists of digital water-level elevation contours for the alluvial and terrace deposits along the Cimarron River in northwestern Oklahoma during...

  2. Digital data sets that describe aquifer characteristics of the alluvial and terrace deposits along the Cimarron River from Freedom to Guthrie in northwestern Oklahoma

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — This data set consists of digital aquifer boundaries for the alluvial and terrace deposits along the Cimarron River from Freedom to Guthrie in northwestern...

  3. Digital data sets that describe aquifer characteristics of the alluvial and terrace deposits along the Cimarron River from Freedom to Guthrie in northwestern Oklahoma

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — This data set consists of digital polygons of constant hydraulic conductivity values for the alluvial and terrace deposits along the Cimarron River from Freedom to...

  4. Digital data sets that describe aquifer characteristics of the alluvial and terrace deposits along the Cimarron River from Freedom to Guthrie in northwestern Oklahoma

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — This data set consists of digital polygons of a constant recharge rate for the alluvial and terrace deposits along the Cimarron River from Freedom to Guthrie in...

  5. Melting characteristics of the stainless steel generated from the uranium conversion plant

    International Nuclear Information System (INIS)

    The partition ratio of cerium (Ce) and uranium (U) in the ingot, slag and dust phases has been investigated for the effect of the slag type, slag concentration and basicity in an electric arc melting process. An electric arc furnace (EAF) was used to melt the stainless steel wastes, simulated by uranium oxide and the real wastes from the uranium conversion plant in Korea Atomic Energy Research Institute (KAERI). The composition of the slag former used to capture the contaminants such as uranium, cerium, and cesium during the melt decontamination process generally consisted of silica (SiO2), calcium oxide (CaO) and aluminum oxide (Al2O3). Also, Calcium fluoride (CaF2 ), nickel oxide (NiO), and ferric oxide (Fe2O3) were added to provide an increase in the slag fluidity and oxidative potential. Cerium was used as a surrogate for the uranium because the thermochemical and physical properties of cerium are very similar to those of uranium. Cerium was removed from the ingot phase to slag phase by up to 99% in this study. The absorption ratio of cerium was increased with an increase of the amount of the slag former. And the maximum removal of cerium occurred when the basicity index of the slag former was 0.82. The natural uranium (UO2) was partitioned from the ingot phase to the slag phase by up to 95%. The absorption of the natural uranium was considerably dependent on the basicity index of the slag former and the composition of the slag former. The optimum condition for the removal of the uranium was about 1.5 for the basicity index and 15 wt% of the slag former. According to the increase of the amount of slag former, the absorption of uranium oxide in the slag phase was linearly increased due to an increase of its capacity to capture uranium oxide within the slag phase. Through experiments with various slag formers, we verified that the slag formers containing calcium fluoride (CaF2) and a high amount of silica were more effective for a melt decontamination of

  6. Dynamic simulation of a direct carbonate fuel cell power plant

    Energy Technology Data Exchange (ETDEWEB)

    Ernest, J.B. [Fluor Daniel, Inc., Irvine, CA (United States); Ghezel-Ayagh, H.; Kush, A.K. [Fuel Cell Engineering, Danbury, CT (United States)

    1996-12-31

    Fuel Cell Engineering Corporation (FCE) is commercializing a 2.85 MW Direct carbonate Fuel Cell (DFC) power plant. The commercialization sequence has already progressed through construction and operation of the first commercial-scale DFC power plant on a U.S. electric utility, the 2 MW Santa Clara Demonstration Project (SCDP), and the completion of the early phases of a Commercial Plant design. A 400 kW fuel cell stack Test Facility is being built at Energy Research Corporation (ERC), FCE`s parent company, which will be capable of testing commercial-sized fuel cell stacks in an integrated plant configuration. Fluor Daniel, Inc. provided engineering, procurement, and construction services for SCDP and has jointly developed the Commercial Plant design with FCE, focusing on the balance-of-plant (BOP) equipment outside of the fuel cell modules. This paper provides a brief orientation to the dynamic simulation of a fuel cell power plant and the benefits offered.

  7. Development and Evaluation of Mixed Uranium-Refractory Carbide/Refractory Carbide Cer-Cer Fuels Project

    Data.gov (United States)

    National Aeronautics and Space Administration — A new carbon-based fuel is introduced with outstanding potential to eliminate the loss of uranium, minimize the loss of carbon, and retain fission products for many...

  8. Radiotoxicity and decay heat power of spent uranium-plutonium and thorium fuel at long-term storage

    International Nuclear Information System (INIS)

    Changes of radiotoxicity and decay heat power of actinides from spent uranium- plutonium and thorium nuclear fuel of WWER-1000 type reactors at storage during 300 years are investigated in report. (author)

  9. Concept of erbium doped uranium oxide fuel cycle in light water reactors

    International Nuclear Information System (INIS)

    This paper is aimed at the development of a fuel cycle concept for host countries with a lack of nuclear infrastructure. To minimize plutonium proliferation concern the adoption of long-life core with no fuel radiochemical treatment on site is suggested. Current investigation relies upon light water reactor technology and plutonium-free fresh fuel. Erbium doped to uranium oxide (enrichment 19.8%) fuel is selected as the reference. Such a high enrichment is selected in attempt to approach the longest irradiation time in one batch mode. In addition to that, uranium enriched up to 20% does not consider as a nuclear material for direct use in weapon manufacture. A sequence of two irradiation cycles for the same fuel rods in two different light water reactors is the key feature of the advocated approach. It is found that the synergism of PWR and pressure tube graphite reactor offers fuel burnup up to 140 GWd/tHM without compromising safety characteristics. Being as large as 8% in the final isotopic vector, fraction of 238Pu serves as an inherent protective measure against plutonium proliferation. (author)

  10. Design of process cell equipment layout and its associated piping in typical nuclear fuel reprocessing plant

    International Nuclear Information System (INIS)

    Nuclear fuel reprocessing plant processes spent nuclear fuel discharged from the nuclear reactor to separate chemically the uranium and plutonium. Spent nuclear fuel emits radiation due to the presence of fission products, actinides and activation products. The major operation steps in reprocessing plant are dismantling of spent fuel subassemblies, chopping of fuel pins and dissolution in concentrated nitric acid. Subsequently, this solution containing uranium and plutonium, fission products and actinides is subjected to solvent extraction with tributyl phosphate in diluent as solvent for separating uranium and plutonium from fission products and other actinides. In the design of a fuel reprocessing plant, apart from problems associated with conventional chemical process industries such as corrosion, materials handling, industrial and fire safety and economy, specific considerations such as health hazards from radioactivity (radiological safety) and damage to material by radiation are considered. This necessitates the processing of spent fuel inside the shielded process cells (concrete and lead cells) with remote operation and maintenance philosophy to prevent the contamination as well as radiation exposure to the operators and prevention of criticality in process tanks and equipments. Reprocessing plant consists of number of shielded process cells depending on the processing capacity and type of spent fuel handled. Concrete cells and lead cells houses various type of storage tanks, equipments, liquid transfer devices, etc with interconnecting small bore pipe lines for liquid transfer and supply of services, which runs in multiple layers, forming a high density piping inside the cells. In addition to this, cells have remote handling systems and gadgets for remote operation and maintenance wherever required. This paper highlights the design of process cells, its equipment layout and piping in typical reprocessing plant; the suitable material of construction

  11. An evaluation of the dissolution process of natural uranium ore as an analogue of nuclear fuel

    International Nuclear Information System (INIS)

    The assumption of congruent dissolution of uraninite as a mechanism for the dissolution behaviour of spent fuel was critically examined with regard to the fate of toxic radionuclides. The fission and daughter products of uranium are typically present in spent unreprocessed fuel rods in trace abundances. The principles of trace element geochemistry were applied in assessing the behaviour of these radionuclides during fluid/solid interactions. It is shown that the behaviour of radionuclides in trace abundances that reside in the crystal structure can be better predicted from the ionic properties of these nuclides rather than from assuming that they are controlled by the dissolution of uraninite. Geochemical evidence from natural uranium ore deposits (Athabasca Basin, Northern Territories of Australia, Oklo) suggests that in most cases the toxic radionuclides are released from uraninite in amounts that are independent of the solution behaviour of uranium oxide. Only those elements that have ionic and thus chemical properties similar to U4+, such as plutonium, americium, cadmium, neptunium and thorium can be satisfactorily modelled by the solution properties of uranium dioxide and then only if the environment is reducing. (84 refs., 7 tabs.)

  12. Environmental report of Purex Plant and Uranium Oxide Plant - Hanford reservation

    International Nuclear Information System (INIS)

    A description of the site, program, and facilities is given. The data and calculations indicate that there will be no significant adverse environmental impact from the resumption of full-scale operations of the Purex and Uranium Oxide Plants. All significant pathways of radionuclides in Purex Plant effluents are evaluated. This includes submersion in the airborne effluent plumes, consumption of drinking water and foodstuffs irrigated with Columbia River water, ingestion of radioactive iodine through the cow-to-milk pathway, consumption of fish, and other less significant pathways. A summary of research and surveillance programs designed to assess the possible changes in the terresstrial and aquatic environments on or near the Hanford Reservation is presented. The nonradiological discharges to the environment of prinicpal interest are chemicals, sewage, and solid waste. These discharges will not lead to any significant adverse effects on the environment

  13. Development of ISA procedure for uranium fuel fabrication and enrichment facilities

    International Nuclear Information System (INIS)

    The integrated safety analysis (ISA) procedure has been developed to apply risk-informed regulation to uranium fuel fabrication and enrichment facilities. The major development efforts are as follows: (a) preparing the risk level matrix as an index for items-relied-on-for-safety (IROFS) identification, (b) defining requirements of IROFS, and (c) determining methods of IROFS importance based on the results of risk- and scenario-based analyses. For the risk level matrix, the consequence and likelihood categories have been defined by taking into account the Japanese regulatory laws, rules, and safety standards. The trial analyses using the developed procedure have been performed for several representative processes of the reference uranium fuel fabrication and enrichment facilities. This paper presents the results of the ISA for the sintering process of the reference fabrication facility. The results of the trial analyses have demonstrated the applicability of the procedure to the risk-informed regulation of these facilities. (author)

  14. Research and Test Reactor Conversion to Low Enriched Uranium Fuel: Technical and Programmatic Progress

    International Nuclear Information System (INIS)

    The U.S Department of Energy (DOE) initiated a program - the Reduced Enrichment for Research and Test Reactors (RERTR) - in 1978 to develop the technology necessary to reduce the use of High Enriched Uranium (HEU) fuel in research reactors by converting them to low enriched uranium (LEU) fuel. In 2004, the reactor conversion program became the driving pilar of the Global Threat Reduction Initiative (GTRI), a program established by the U.S. DOE's National Nuclear Security Administration. The overall GTRI objectives are the conversion, removal or protection of vunerable civilian radiological and nuclear material. As part of the GTRI, the Conversion Program has accelerated the schedules and plans for conversion of additional research reactors operating with HEU. This paper provides an update on the progress made since 2007 and describes current technical challenges that the program faces. (author)

  15. Radiological health aspects of commercial uranium conversion, enrichment, and fuel fabrication

    International Nuclear Information System (INIS)

    Detailed information concerning occupational exposures, health physics practices, and regulatory procedures at commercial conversion, enrichment and fuel fabrication facilities is given. Sites visits were the primary source of information, which is divided into four sections. The first section discusses health physics practices that are common to the conversion, enrichment, and fuel fabrication phases of the commercial uranium industry. The next three sections review process descriptions, radiological health practices, and regulatory procedures for the three phases. Nonradiological exposures are considered only as they influence the interpretation of the health effects of radiological exposures. The review of regulatory procedures indicates the types of exposure evaluation records being kept on uranium workers and the responsibility for maintaining the records

  16. Concentration of uranium and plutonium in unsaturated spent fuel tests

    International Nuclear Information System (INIS)

    Commercial spent fuel is being tested under oxidizing conditions at 90 C in drip tests with simulated groundwater to evaluate its long-term performance in a potential repository at Yucca Mountain [1-4]. The tests allow us to monitor the dissolution behavior of the spent fuel matrix and the release rates of individual radionuclides. This paper reports the U and Pu concentrations in the leachates of drip tests during 3.7 years of reaction. Changes in these concentrations are correlated with changes in the measured pH and the appearance of alteration products on the fuel surface. Although there is little thermodynamic information at 90 C for either uranyl or plutonium compounds, some data are available at 25 C [5-8]. The literature data for the U and Pu solubilities of U and Pu compounds were compared to the U and Pu concentrations in the leachates. We also compare Wilson's [9] U and Pu concentrations in semi-static tests at 85 C on spent fuel with our results

  17. Proliferation resistance and energy security advantages of a thorium-uranium dioxide once-through fuel cycle for light water reactors

    International Nuclear Information System (INIS)

    This study analyzes whether spent light reactor (LWR) thorium-uranium dioxide fuel poses a significantly lower risk for nuclear weapon proliferation than spent uranium-dioxide fuel, based on the isotopic composition of the contained uranium and plutonium. Mixed Th/U fuel with an initial enrichment of 19.5% U235 can achieve an average burnup of 70,000 MWd/tHM in a PWR using 30% UO2 and 70% ThO2. To get the equivalent burnup, LEU fuel requires an initial enrichment of 8.0% U235. Two computer codes, MCNP and ORIGEN2, are used to perform the depletion calculation. The spent mixed thorium-uranium dioxide fuel discharged from a pressurized-water reactor has a plutonium isotopic composition and higher decay heat production per kilogram of plutonium more proliferation resistant than spent low enriched uranium dioxide fuel, while significantly reducing the quantity of plutonium produced. The U233 + U235 mixture in spent thorium-uranium fuel is low enriched and contaminated with gamma-emitting U232. With respect to energy security, the introduction of a thorium-uranium fuel cycle could reduce concern over uranium fuel supply of a resource-poor nation since thorium reserve is much larger, compared to fuel cycles using 4.5% LEU, while its uranium saving is almost equivalent to plutonium recycling. Overall, spent thorium-uranium fuel appears significantly more proliferation resistant in terms of the weapons-usability of the contained fissile material than spent low enriched uranium fuel, although use of 19.5% enriched uranium in fresh fuel would facilitate production of weapons-grade uranium at a higher rate in countries with clandestine enrichment facilities. (S.Y.)

  18. UNIFRAME interim design report. [Fuel element size reduction plant

    Energy Technology Data Exchange (ETDEWEB)

    Strand, J.B.; Baer, J.W.; Cook, E.J.

    1977-12-01

    A fuel element size reduction system has been designed for the ''cold'' pilot-scale plant for an HTGR Fuel Reference Recycle Facility. This report describes in detail the present design.

  19. Criteria for the safe storage of enriched uranium at the Y-12 Plant

    Energy Technology Data Exchange (ETDEWEB)

    Cox, S.O.

    1995-07-01

    Uranium storage practices at US Department of Energy (DOE) facilities have evolved over a period spanning five decades of programmatic work in support of the nuclear deterrent mission. During this period, the Y-12 Plant in Oak Ridge, Tennessee has served as the principal enriched uranium facility for fabrication, chemical processing, metallurgical processing and storage. Recent curtailment of new nuclear weapons production and stockpile reduction has created significant amounts of enriched uranium available as a strategic resource which must be properly and safely stored. This standard specifies criteria associated with the safe storage of enriched uranium at the Y-12 Plant. Because programmatic needs, compliance regulations and desirable materials of construction change with time, it is recommended that these standards be reviewed and amended periodically to ensure that they continue to serve their intended purpose.

  20. Recovery of fissile materials from plutonium residues, miscellaneous spent nuclear fuel, and uranium fissile wastes

    International Nuclear Information System (INIS)

    A new process is proposed that converts complex feeds containing fissile materials into a chemical form that allows the use of existing technologies (such as PUREX and ion exchange) to recover the fissile materials and convert the resultant wastes to glass. Potential feed materials include (1) plutonium scrap and residue, (2) miscellaneous spent nuclear fuel, and (3) uranium fissile wastes. The initial feed materials may contain mixtures of metals, ceramics, amorphous solids, halides, and organics. 14 refs., 4 figs

  1. Purification of uranium alloys by differential solubility of oxides and production of purified fuel precursors

    International Nuclear Information System (INIS)

    A method is described for purifying metallic alloys of uranium for use as nuclear reactor fuels in which the metal alloy is first converted to an oxide and then dissolved in nitric acid. Initial removal of metal oxide impurities not soluble in nitric acid is accomplished by filtration or other physical means. Further purification can be accomplished by carbonate leaching of uranyl ions from the partially purified solution or using traditional methods such as solvent extraction. 3 figs

  2. Quantitative nondestructive assay of uranium bearing fuel rods by high resolution gamma spectrometry

    International Nuclear Information System (INIS)

    A gamma spectrometric method is presented for quantitative assay of uranium bearing fuel rods for safeguards purposes. For determination of enrichment the 60 ... 100 keV region and intrinsic calibration method was used. The determination of the total amount of 238U was based on the measurement of 1001 keV gamma-ray and a careful self-attenuation correction. The method works without use of standards. (author)

  3. Determination of uranium in thorium matrix- a novel approach to quality control in nuclear fuel cycle

    International Nuclear Information System (INIS)

    In view of potential role of thorium in the future nuclear programme, research and development work in various aspects of thorium fuel cycle has been undertaken. Besides, the recent interest in thorium utilisation in Pressurised Heavy Water Reactors (PHWRs) necessitates monitoring of its purity with respect to a number of metallics which are of significance to its performance. Prominent among these metals and equally difficult to determine at trace level is uranium

  4. Rules and regulations for uranium fuel elements smuggled into Hungary, regarding public health services

    International Nuclear Information System (INIS)

    Case studies with known illegal radioactive material import to Hungary are presented. The National Research Institute of Radiobiology and Radiohygiene was commissioned to assess the health consequences of unreported radioactivity transportation. The majority of the unreported radioactive materials were identified as uranium fuel elements. Radiation protection and health measures for such incident have been elaborated by NRIRR to cope with the occurrences in the future. (N.T.) 2 figs.; 4 tabs

  5. Benefits/impacts of utilizing depleted uranium silicate glass as backfill for spent fuel waste packages

    International Nuclear Information System (INIS)

    An assessment has been made of the benefits and impacts which can be derived by filling a spent nuclear fuel multi-purpose canister with depleted uranium silicate (DUS) glass at a reactor site. Although the primary purpose of the DUS glass fill would be to enhance repository performance assessment and control criticality of geologic times, a number of benefits to the waste management system can be derived from adding the DUS glass prior to shipment from the reactor site

  6. Neutronic analysis of PROMETHEUS reactor fueled with various compounds of thorium and uranium

    International Nuclear Information System (INIS)

    In this study, neutronic performance of the DT driven blanket in the PROMETHEUS-H (heavy ion) fueled with different fuels, namely, ThO2, ThC, UO2, UC, U3Si2 and UN is investigated. Helium is used as coolant, and SiC is used as cladding material to prevent fission products from contaminating coolant and direct contact fuel with coolant in the blanket. Calculations of neutronic data per DT fusion neutron are performed by using SCALE 4.3 Code. M (energy multiplication factor) changes from 1.480 to 2.097 depending on the fuel types in the blanket under resonance-effect. M reaches the highest value in the blanket fueled with UN. Therefore, the investigated reactor can produce substantial electricity in situ. UN has the highest value of 239Pu breeding capability among the uranium fuels whereas UO2 has the lowest one. 239Pu production ratio changes from 0.119 to 0.169 according to the uranium fuel types, and 233U production values are 0.125 and 0.140 in the blanket fueled with ThO2 and ThC under resonance-effect, respectively. Heat production per MW (D,T) fusion neutron load varies from 1.30 to 7.89 W/cm3 in the first row of fissile fuel breeding zone depending on the fuel types. Heat production attains the maximum value in the blanket fueled with UN. Values of TBR (tritium breeding ratio) being one of the most important parameters in a fusion reactor are greater than 1.05 for all type of fuels so that tritium self-sufficiency is maintained for DT fusion driver. Values of peak-to-average fission power density ratio, Γ, are in the range of 1.390 and ∼1.476 depending on the fuel types in the blanket. Values of neutron leakage out of the blanket for all fuels are quite low due to SiC reflector. The maximum neutron leakage is only ∼0.025. Consequently, for all cases, the investigated reactor has high neutronic performance and can produce substantial electricity in situ, fissile fuel and tritium required for (D,T) fusion reaction

  7. The obtainment of highly concentrated uranium pellets for plate type (MTR) fuel by dispersion of uranium aluminides in aluminium

    International Nuclear Information System (INIS)

    The use of the intermetallic UAl3 for manufacturing plate type MTR fuel with 20% U235 enriched uranium and a density of about 20 kg/m3 is analyzed. The technique used is the dispersion of UAl3 particles in aluminium powder. The obtainment of the UAl3 intermetallic was performed by fusion in an induction furnace in an atmosphere of argon at a pressure of 0.7 BAR (400 mm) using an alumina melting pot. To make the aluminide powder and attain the wished granulometry a cutting and a rotating crusher were used. Aluminide powders of different granulometries and different pressures of compactation were analyzed. In each case the densities were measured. The compacts were colaminated with the 'Picture Frame' technique at temperatures of 490 and 0 deg C with excellent results from the manufacturing view point. (M.E.L.)

  8. Evaluation of the uranium market and its consequences in the strategy of a nuclear fuel supplier that is also a uranium producer

    International Nuclear Information System (INIS)

    On January 2005, the uranium spot market price reached the value of $21.00/lbU3O8. One month before, at the end of December, the average price was $20.70/lbU3O8 and in November the spot price registered $20.50. When we review this abstract, on July 2005, the price has reached $30.00/lbU3O8. In 1984, the uranium spot price dropped below the twenties and remained so reaching meanwhile even one-digit values, even considering that the uranium offer in this period was always below the demand. The main reason for that distortion in the market was and still is, the interference of the developing countries governments after the end of the cold war The Industrias Nucleares do Brasil - INB is in an odd situation in the market of fuel suppliers due to being also a uranium producer and in short future will also be an enrichment services supplier. This peculiar position brings additional advantages due to the flexibility to play with the uranium costs versus tail assay to optimize its nuclear fuel costs. That odd position, equivalent only in the market to AREVA, allows INB to exchange uranium by SWU and vice versa according to its uranium cost (not market sell price) and in the future to the SWU's costs obtaining a better margin that can not be reached by other fuel suppliers. In the first part of this paper it is evaluated, based on the recent market information, the consequences in the 2004 uranium spot price, expected to be more emphasized during 2005. This paper also evaluate the market mechanisms for expecting the price to cross the $40/lbU3O8 in short time The market supply mechanisms used up to now to fulfil the market deficit may be interrupted in case the developing countries governments stop the availability of the non civil uranium reserves from its stockpile. Different hypotheses for supplying the primary uranium deficit in this last case are analyzed in this work and evaluated its consequences. The solution of reducing the actual tails assay used aiming at

  9. Uranium and plutonium determinations for evaluation of high burnup fuel performance

    International Nuclear Information System (INIS)

    Purpose of this work is to experimentally test computational methods being developed for reactor fuel operation. Described are the analytical techniques used in the determination of uranium and plutonium compositions on PWR fuel that has spanned five power cycles, culminating in 55,000 to 57,000 MWd/T burnup. Analyses have been performed on ten samples excised from selected sections of the fuel rods. Hot cell operations required the separation of fuel from cladding and the comminution of the fuel. These tasks were successfully accomplished using a SpectroMil, a ball pestle impact grinding and blending instrument manufactured by Chemplex Industries, Inc., Eastchester, New York. The fuel was dissolved using strong mineral acids and bomb dissolution techniques. Separation of the fuel from fission products was done by solvent (hexone) extraction. Fuel isotopic compositions and assays were determined by the mass spectrometric isotope dilution (MSID) method using NBS standards SRM-993 and SRM-996. Alpha spectrometry was used to determine the 238Pu composition. Relative correlations of composition with burnup were obtained by gamma-ray spectrometry of selected fission products in the dissolved fuel

  10. Thorium and the Third Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Dukert, Joseph M.

    1970-01-01

    This booklet discusses energy sources for nuclear power plants. Uranium-235 by itself will not be able to handle the energy needs. The two man-made supplements that can be used for nuclear power plants energy sources are plutonium and uranium-233. Uranium-233 is an isotope that appears as a result of radioactive decay after neutrons have been absorbed in thorium-232. This uranium-233 is called the third fuel.

  11. Uranium transfer in the food chain from soil to plants, animals and man

    International Nuclear Information System (INIS)

    Our investigations aimed at following up the scientific basis of uranium transfer from the soils of different geological origins and from the immediate vicinity of uranium waste dumps in the vegetation, in waters (drinking water, mineral water and medicinal water), vegetable and animal foodstuffs and beverages; the regional human uranium intake, excretion, apparent absorption and balance in Germany and Mexico. Another aim of the investigations was to draw conclusions from the rules of transfer of this element from the rocks and soils to plants, animals and man. (authors)

  12. Selected bibliography for the extraction of uranium from seawater: chemical process and plant design feasibility study

    Energy Technology Data Exchange (ETDEWEB)

    Binney, S.E.; Polkinghorne, S.T.; Jante, R.R.; Rodman, M.R.; Chen, A.C.T.; Gordon, L.I.

    1979-02-01

    A selected annotated bibliography of 521 references was prepared as a part of a feasibility study of the extraction of uranium from seawater. For the most part, these references are related to the chemical processes whereby the uranium is removed from the seawater. A companion docment contains a similar bibliography of 471 references related to oceanographic and uranium extraction plant siting considerations, although some of the references are in common. The bibliography was prepared by computer retrieval from Chemical Abstracts, Nuclear Science Abstracts, Energy Data Base, NTIS, and Oceanic Abstracts. References are listed by author, country of author, and selected keywords.

  13. Selected bibliography for the extraction of uranium from seawater: chemical process and plant design feasibility study

    International Nuclear Information System (INIS)

    A selected annotated bibliography of 521 references was prepared as a part of a feasibility study of the extraction of uranium from seawater. For the most part, these references are related to the chemical processes whereby the uranium is removed from the seawater. A companion docment contains a similar bibliography of 471 references related to oceanographic and uranium extraction plant siting considerations, although some of the references are in common. The bibliography was prepared by computer retrieval from Chemical Abstracts, Nuclear Science Abstracts, Energy Data Base, NTIS, and Oceanic Abstracts. References are listed by author, country of author, and selected keywords

  14. Position paper Oak Ridge Y-12 Plant storage of uranium in plastics

    International Nuclear Information System (INIS)

    As a result of the end of the Cold War, the United States nuclear weapon stockpile is being reduced from approximately 20,000 warheads to fewer than 10,000 by the end of the century. The Oak Ridge Y-12 Plant is the Department of Energy (DOE) site charged with the responsibility of providing safe, secure storage for the uranium recovered from these weapons. In addition to weapons material, Y-12 has traditionally processed and stored uranium from nonweapon programs and presumably will continue to do so. The purpose of this document is to evaluate the suitability of plastics for use in the containment of uranium

  15. Uranium silicide pellet fabrication by powder metallurgy for accident tolerant fuel evaluation and irradiation

    Science.gov (United States)

    Harp, Jason M.; Lessing, Paul A.; Hoggan, Rita E.

    2015-11-01

    In collaboration with industry, Idaho National Laboratory is investigating uranium silicide for use in future light water reactor fuels as a more accident resistant alternative to uranium oxide base fuels. Specifically this project was focused on producing uranium silicide (U3Si2) pellets by conventional powder metallurgy with a density greater than 94% of the theoretical density. This work has produced a process to consistently produce pellets with the desired density through careful optimization of the process. Milling of the U3Si2 has been optimized and high phase purity U3Si2 has been successfully produced. Results are presented from sintering studies and microstructural examinations that illustrate the need for a finely ground reproducible particle size distribution in the source powder. The optimized process was used to produce pellets for the Accident Tolerant Fuel-1 irradiation experiment. The average density of these pellets was 11.54 ± 0.06 g/cm3. Additional characterization of the pellets by scanning electron microscopy and X-ray diffraction has also been performed. Pellets produced in this work have been encapsulated for irradiation, and irradiation in the Advanced Test Reactor is expected soon.

  16. Uranium silicide pellet fabrication by powder metallurgy for accident tolerant fuel evaluation and irradiation

    International Nuclear Information System (INIS)

    In collaboration with industry, Idaho National Laboratory is investigating uranium silicide for use in future light water reactor fuels as a more accident resistant alternative to uranium oxide base fuels. Specifically this project was focused on producing uranium silicide (U3Si2) pellets by conventional powder metallurgy with a density greater than 94% of the theoretical density. This work has produced a process to consistently produce pellets with the desired density through careful optimization of the process. Milling of the U3Si2 has been optimized and high phase purity U3Si2 has been successfully produced. Results are presented from sintering studies and microstructural examinations that illustrate the need for a finely ground reproducible particle size distribution in the source powder. The optimized process was used to produce pellets for the Accident Tolerant Fuel-1 irradiation experiment. The average density of these pellets was 11.54 ±0.06 g/cm3. Additional characterization of the pellets by scaning electron microscopy and X-ray diffraction has also been performed. As a result, pellets produced in this work have been encapsulated for irradiation, and irradiation in the Advanced Test Reactor is expected soon

  17. The mechanisms of the metal uranium fuel radiation creep and its temperature un-lineses

    International Nuclear Information System (INIS)

    Early it is shown, that in metal uranium at moderate temperatures and high neutron flux densities the radiation point defect concentrations considerably may exceed thermally equilibrium, also it is submitted the theoretical model of the established radiation creep within the framework of the mechanism of dislocation gliding and climbing, based on the conception of a dislocation as nonideal sink for point radiation defects. Are developed the 'dynamics' method of mathematical simulation of gliding and climbing of flexible dislocation interacting to obstacles of a various type (spherical centre of extension, dislocation prismatic loop and their spatially random distributions), and a computer program complex for mathematical simulation of radiation creep and radiation creep rate estimations by the Monte Carlo method. Computer simulation of metal uranium fuel radiation creep has revealed it's temperature nonlinearity. These temperature nonlinearities have served the reason for search of nonlinear effects and modes in open nonlinear stochastic system which the system metal uranium fuel under an irradiation. The generalized diagrams of radiation-thermal creep for not fission (constructional) metals and for fission (fuel) metals are submitted. The regions of various radiation creep theoretical models are qualitatively determined. It is shown, that the nonlinearities of curve thermal dependence of the radiation creep established rate are caused by the features of used radiation creep theoretical model based on the concept of a dislocation as nonideal sink for point radiation defects and the features of point radiation defects generation in divisional metals

  18. Uranium resource utilization improvements in the once-through PWR fuel cycle

    International Nuclear Information System (INIS)

    In support of the Nonproliferation Alternative Systems Assessment Program (NASAP), Combustion Engineering, Inc. performed a comprehensive analytical study of potential uranium utilization improvement options that can be backfit into existing PWRs operating on the once-through uranium fuel cycle. A large number of potential improvement options were examined as part of a preliminary survey of candidate options. The most attractive of these, from the standpoint of uranium utilization improvement, economic viability, and ease of implementation, were then selected for detailed analysis and were included in a single composite improvement case. This composite case represents an estimate of the total savings in U3O8 consumption that can be achieved in current-design PWRs by implementing improvements which can be developed and demonstrated in the near term. The improvement options which were evaluated in detail and included in the composite case were a new five-batch, extended-burnup fuel management scheme, low-leakage fuel management, modified lattice designs, axial blankets, reinsertion of initial core batches, and end-of-cycle stretchout

  19. Neutronic calculations of PARR-1 cores using leu-silicide fuel. [leu (low enriched uranium)

    Energy Technology Data Exchange (ETDEWEB)

    Arshad, M.; Bakhtyar, S.; Hayat, T.; Salahuddin, A.

    1991-08-01

    Detailed neutronic calculations have been carried out for different PARR-1 cores utilizing Low Enriched Uranium (LEU) silicide fuel and operating at an upgraded power of 9 MW. The calculations include the search for critical loadings in open and stall ends of the pool, neutronic analysis of the first full power operation and the equilibrium cores. The burnup study of the equilibrium core and calculations for discharged fuel inventory have also been carried out. Further, the reactivity coefficients of the first full power operation core are evaluated for use in the accident analysis.

  20. Assessing depleted uranium (DU) contamination of soil, plants and earthworms at UK weapons testing sites

    OpenAIRE

    Oliver, I.W.; Graham, M C; Mackenzie, A. B.; Ellam, R.M.; Farmer, J.G.

    2007-01-01

    Depleted uranium (DU) weapons testing programmes have been conducted at two locations within the UK. An investigation was therefore carried out to assess the extent of any environmental contamination arising from these test programmes using both alpha spectrometry and mass spectrometry techniques. Uranium isotopic signatures indicative of DU contamination were observed in soil, plant and earthworm samples collected in the immediate vicinity of test firing points and targets, but contamination...

  1. Production of uranium oxide concentrates by the Nuclear Fuels Corporation of South Africa

    International Nuclear Information System (INIS)

    South Africa has a relatively long history of large-scale production of uranium from many of its gold mines. The final processing is undertaken in a single plant, operated by Nufcor for the producing mines, which has been in continuous commercial use for over 25 years. The plant is adapted to the handling of incoming material of varying chemical and physical qualities, and the equipment can be readily adjusted for optimum performance at widely different levels of production. This paper describes the plant equipment and its operation, emphasising the special conditions which led to its design and the features which have contributed to its success. All aspects of the operation are described, from collecting aqueous uranium slurry at the mines through drying and calcining to the sampling, analyses, packing and final transport of oxide concentrates for export. (author)

  2. Organization of Basic Materials Auditing in a Fuel-Element Fabrication Plant

    International Nuclear Information System (INIS)

    The authors describe the organization of auditing at the Annecy plant of the Société industrielle des combustibles nucléaires which produces, principally from natural uranium provided by the CEA, fuel elements for reactors of the graphite-gas type (in particular, G1, G2, G3, EDF1, EDF2, EDF3). The plant may be schematically divided into five large sections: the melting shop and its annexes; machine shop; cladding shop; turnings recovery shop; warehouse. The auditing records used at Annecy may be classified in three categories: The ''follow-through'' slips, the stock slips and the despatch notes. The balance sheet established each month for uranium consists of two parts: one part showing stocks existing at the beginning and end of the month and internal and external movements during the month, and a recapitulation balance, which is so to say an account of the use of uranium, showing in parallel the amounts of uranium used and the production. The debit balance of this account corresponds to real or apparent losses. (author)

  3. Methodology for comparing the health effects of electricity generation from uranium and coal fuels

    International Nuclear Information System (INIS)

    A methodology was developed for comparing the health risks of electricity generation from uranium and coal fuels. The health effects attributable to the construction, operation, and decommissioning of each facility in the two fuel cycle were considered. The methodology is based on defining (1) requirement variables for the materials, energy, etc., (2) effluent variables associated with the requirement variables as well as with the fuel cycle facility operation, and (3) health impact variables for effluents and accidents. The materials, energy, etc., required for construction, operation, and decommissioning of each fuel cycle facility are defined as primary variables. The materials, energy, etc., needed to produce the primary variable are defined as secondary requirement variables. Each requirement variable (primary, secondary, etc.) has associated effluent variables and health impact variables. A diverging chain or tree is formed for each primary variable. Fortunately, most elements reoccur frequently to reduce the level of analysis complexity. 6 references, 11 figures, 6 tables

  4. Chemical interaction in uranium-plutonium mixed oxide fuel pins for LMFBR

    International Nuclear Information System (INIS)

    A review is made on the current understanding and problems of chemical interaction between uranium-plutonium mixed oxide and stainless steel cladding for LMFBR fuel pins. The oxygen potential of the fuel was considered as one of the key factors that influences the interaction and the methods of its measurement, its change with irradiation, effect of oxygen redistribution and measured values of irradiated fuel are described. The mechanisms of conventional intergranular and matrix attacks and more recent cladding component chemical transport (CCCT), which was proposed by GE and has been often observed in highly irradiated fuel pins, are explained. Finally, description is given on a statistical analysis of the attack depth and method of inhibiting the cladding. (author)

  5. Safety of Uranium and Plutonium Mixed Oxide Fuel Fabrication Facilities. Specific Safety Guide

    International Nuclear Information System (INIS)

    This Safety Guide supplements the Safety Requirements publication Safety of Fuel Cycle Facilities and addresses all the stages in the life cycle of MOX fuel fabrication facilities, with emphasis placed on design and operation. It describes the actions, conditions and procedures for meeting safety requirements and deals specifically with the handling, processing and storage of plutonium oxide, depleted, natural or reprocessed uranium oxide or mixed oxide manufactured from the above to be used as a feed material to form MOX fuel rods and assemblies for export and subsequent use in water reactors and fast breeder reactors. The publication is intended to be of use to designers, operating organizations and regulators to ensure the safety of MOX fuel fabrication facilities. Contents: 1. Introduction; 2. General safety recommendations; 3. Site evaluation; 4. Design; 5. Construction; 6. Commissioning; 7. Operation; 8. Decommissioning; Annexes.

  6. Disposal of defense spent fuel and HLW from the Idaho Chemical Processing Plant

    International Nuclear Information System (INIS)

    Acid high-level radioactive waste (HLW) resulting from fuel reprocessing at the Idaho Chemical Processing Plant (ICPP) for the US Department of Energy (DOE) has been solidified to a calcine since 1963 and stored in stainless steel bins enclosed by concrete vaults. Several different types of unprocessed irradiated DOE-owned fuels are also in storage ate the ICPP. In April, 1992, DOE announced that spent fuel would no longer be reprocessed to recover enriched uranium and called for a shutdown of the reprocessing facilities at the ICPP. A new Spent Fuel and HLW Technology Development program was subsequently initiated to develop technologies for immobilizing ICPP spent fuels and HLW for disposal, in accordance with the Nuclear Waste Policy Act. The Program elements include Systems Analysis, Graphite Fuel Disposal, Other Spent Fuel Disposal, Sodium-Bearing Liquid Waste Processing, Calcine Immobilization, and Metal Recycle/Waste Minimization. This paper presents an overview of the ICPP radioactive wastes and current spent fuels, with an emphasis on the description of HLW and spent fuels requiring repository disposal

  7. Behavior of silicon in nitric media. Application to uranium silicides fuels reprocessing; Comportement du silicium en milieu nitrique. Application au retraitement des combustibles siliciures d'uranium

    Energy Technology Data Exchange (ETDEWEB)

    Cheroux, L

    2001-07-01

    Uranium silicides are used in some research reactors. Reprocessing them is a solution for their cycle end. A list of reprocessing scenarios has been set the most realistic being a nitric dissolution close to the classic spent fuel reprocessing. This uranium silicide fuel contains a lot of silicon and few things are known about polymerization of silicic acid in concentrated nitric acid. The study of this polymerization allows to point out the main parameters: acidity, temperature, silicon concentration. The presence of aluminum seems to speed up heavily the polymerization. It has been impossible to find an analytical technique smart and fast enough to characterize the first steps of silicic acid polymerization. However the action of silicic species on emulsions stabilization formed by mixing them with an organic phase containing TBP has been studied, Silicon slows down the phase separation by means of oligomeric species forming complex with TBP. The existence of these intermediate species is short and heating can avoid any stabilization. When non irradiated uranium silicide fuel is attacked by a nitric solution, aluminum and uranium are quickly dissolved whereas silicon mainly stands in solid state. That builds a gangue of hydrated silica around the uranium silicide particulates without preventing uranium dissolution. A small part of silicon passes into the solution and polymerize towards the highly poly-condensed forms, just 2% of initial silicon is still in molecular form at the end of the dissolution. A thermal treatment of the fuel element, by forming inter-metallic phases U-Al-Si, allows the whole silicon to pass into the solution and next to precipitate. The behavior of silicon in spent fuels should be between these two situations. (author)

  8. Blueprint for domestic uranium enrichment

    International Nuclear Information System (INIS)

    The AEC advisory committee on domestic production of uranium enrichment has studied for more than a year how to achieve the domestic enrichment of uranium by the construction and operation of a commercial enriching plant using centrifugal separation method, and the report was submitted to the Atomic Energy Commission on August 18, 1980. Japan has depended wholly on overseas services for her uranium enrichment needs, but the development of domestic enrichment has been carried on in parallel. The AEC decided to construct a uranium enrichment pilot plant using centrifuges, and it has been forwarded as a national project. The plant is operated by the Power Reactor and Nuclear Fuel Development Corp. since 1979. The capacity of the plant will be raised to approximately 75 ton SWU a year. The centrifuges already operated have provided the first delivery of fuel of about 1 ton for the ATR ''Fugen''. The demand-supply balance of uranium enrichment service, the significance of the domestic enrichment of uranium, the evaluation of uranium enrichment technology, the target for domestic enrichment plan, the measures to promote domestic uranium enrichment, and the promotion of the construction of a demonstration plant are reported. (Kako, I.)

  9. Uranium

    International Nuclear Information System (INIS)

    Canada produced one-third of the Western World's uranium production in 1989, twice as much from Saskatchewan as from Ontario, where mine closures have led to the loss of over 2,000 jobs. Canadian production in 1990 was about 8.8 Gg U. In 1990, Canada's primary producers were Denison Mines, Rio Algom, Cluff Mining, and Cameco. In Saskatchewan, there are three operations: Key Lake, Rabbit Lake/Collins Bay, and Cluff Lake. Canada stands fourth in uranium resources, but because of favourable geology remains the focus of much exploration activity, which cost about C$60 in 1989. Large stockpiles overhang the market, so new sources of uranium will not be needed before the mid 1990's, but long-term prospects seem good

  10. Uranium

    International Nuclear Information System (INIS)

    The uranium production industry is well into its third recession during the nuclear era (since 1945). Exploration is drastically curtailed, and many staffs are being reduced. Historical market price production trends are discussed. A total of 3.07 million acres of land was acquired for exploration; drastic decrease. Surface drilling footage was reduced sharply; an estimated 250 drill rigs were used by the uranium industry during 1980. Land acquisition costs increased 8%. The domestic reserve changes are detailed by cause: exploration, re-evaluation, or production. Two significant discoveries of deposits were made in Mohave County, Arizona. Uranium production during 1980 was 21,850 short tons U3O8; an increase of 17% from 1979. Domestic and foreign exploration highlights were given. Major producing areas for the US are San Juan basin, Wyoming basins, Texas coastal plain, Paradox basin, northeastern Washington, Henry Mountains, Utah, central Colorado, and the McDermitt caldera in Nevada and Oregon. 3 figures, 8 tables

  11. The technique for determination of surface contamination by uranium on U3Si2-Al plate-type fuel elements

    International Nuclear Information System (INIS)

    The NDT method for determining the surface contamination by uranium on U3Si2-Al plate-type fuel elements, the process of standard specimen preparation and the graduation curve are described. The measurement results of U3Si2-Al plate-type fuel elements show that the alpha counting method to measure the surface contamination by uranium on fuel plate is more reliable. The UB-1 type surface contamination meter, which was recently developed, has many advantages such as high sensitivity to determine the uranium pollution, short time in measuring, convenience for operation, and the minimum detectable amount of uranium is 5 x 10-10 g/cm2. The measuring device is controlled by a microcomputer. Besides data acquisition and processing, it has functions of statistics, output data on terminal or to printer and alarm. The procedures of measurement are fully automatic. All of these will meet the measuring needs in batch process

  12. Low-Enriched Uranium Fuel Conversion Activities for the High Flux Isotope Reactor, Annual Report for FY 2011

    Energy Technology Data Exchange (ETDEWEB)

    Renfro, David G [ORNL; Cook, David Howard [ORNL; Freels, James D [ORNL; Griffin, Frederick P [ORNL; Ilas, Germina [ORNL; Sease, John D [ORNL; Chandler, David [ORNL

    2012-03-01

    This report describes progress made during FY11 in ORNL activities to support converting the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum (UMo) alloy. With both radial and axial contouring of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current levels achieved with HEU fuel. Studies are continuing to demonstrate that the fuel thermal safety margins can be preserved following conversion. Studies are also continuing to update other aspects of the reactor steady state operation and accident response for the effects of fuel conversion. Technical input has been provided to Oregon State University in support of their hydraulic testing program. The HFIR conversion schedule was revised and provided to the GTRI program. In addition to HFIR conversion activities, technical support was provided directly to the Fuel Fabrication Capability program manager.

  13. Preliminary studies of the genetic structure of “Cimarron uruguayo” dog using microsatellite markers

    Directory of Open Access Journals (Sweden)

    Rosa Gagliardi B.

    2010-12-01

    Full Text Available Objetive. To analyze the population structure, using microsatellite markers in a sample of “Cimarron Uruguayo” dogs. Materials and methods. Thirty dogs were analyzed in different areas of Uruguay with a set of nine molecular microsatellite markers using PCR. The population structure was analyzed using the free distribution software “Structure’’. Results. According to our data, the preliminary results show that it is not possible to establish a subdivision among the animals in the sample. Conclusions. The study supports the hypothesis that the currently existing canines derive from a founding nucleus that took refuge in the Northeastern region of the country. The distribution of the breed among the different areas of Uruguay continues nowadays, so there is no isolation among the different groups of animals, and the exchange is constant

  14. The current uranium exploration activities of the Power Reactor and Nuclear Fuel Development Corporation (PNC), Japan

    International Nuclear Information System (INIS)

    As of November 1996, Japan's total installed commercial nuclear power generation capacity was 42 GW(e), accounting for 34% of total electric energy generation. By 2010, Japan intends to have an installed electricity generation capacity of 70.5 GW(e). This will increase the country's demand for nat Ural uranium from 7,700 t U in 1994 (13% of the world consumption) to 13,800 t U in 2010 (17%-19% of the world projected consumption). However, Japan's known uranium resources at Ningyo-Toge and Tono deposits, are estimated at roughly only 6,600 t U. The Long-term Programme for Research, Development and Utilization of Nuclear Energy (adopted in 1994) calls for diversification through long-term purchasing contracts, independent exploration and involvement in mining vent Ures, with the objective of ensuring independence and stability in Japan's development and utilization of nuclear energy. The Power Reactor and Nuclear Fuel Development Corporation (PNC) has been commissioned to carry out the task of independent exploration. PNC is carrying out exploration projects in Canada, Australia, USA and China targeting unconformity related type deposits with an eye to privatizing them. Currently about 40,000 t U of uranium resources are held by PNC. PNC has been carrying out the following related activities: (1) Reference surveys on uranium resources to delineate the promising areas; (2) Development of uranium exploration technology; (3) Information surveys on the nuclear industries to project long-term supply and demand; (4) International Cooperation programme on uranium exploration with Asian countries. (author)

  15. Resin-based preparation of HTGR fuels: operation of an engineering-scale uranium loading system

    International Nuclear Information System (INIS)

    The fuel particles for recycle of 233U to High-Temperature Gas-Cooled Reactors are prepared from uranium-loaded carboxylic acid ion exchange resins which are subsequently carbonized, converted, and refabricated. The development and operation of individual items of equipment and of an integrated system are described for the resin-loading part of the process. This engineering-scale system was full scale with respect to a hot demonstration facility, but was operated with natural uranium. The feed uranium, which consisted of uranyl nitrate solution containing excess nitric acid, was loaded by exchange with resin in the hydrogen form. In order to obtain high loadings, the uranyl nitrate must be acid deficient; therefore, nitric acid was extracted by a liquid organic amine which was regenerated to discharge a NaNO3 or NH4NO3 solution waste. Water was removed from the uranyl nitrate solution by an evaporator that yielded condensate containing less than 0.5 ppM of uranium. The uranium-loaded resin was washed with condensate and dried to a controlled water content via microwave heating. The loading process was controlled via in-line measurements of the pH and density of the uranyl nitrate. The demonstrated capacity was 1 kg of uranium per hour for either batch loading contractors or a continuous column as the resin loading contractor. Fifty-four batch loading runs were made without a single failure of the process outlined in the chemical flowsheet or any evidence of inability to control the conditions dictated by the flowsheet

  16. Preliminary Evaluation of Alternate Designs for HFIR Low-Enriched Uranium Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Renfro, David [ORNL; Chandler, David [ORNL; Cook, David [ORNL; Ilas, Germina [ORNL; Jain, Prashant [ORNL; Valentine, Jennifer [ORNL

    2014-10-30

    Engineering design studies of the feasibility of conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL) as part of an effort sponsored by the U.S. Department of Energy’s Global Threat Reduction Initiative (GTRI)/Reduced Enrichment for Research and Test Reactors (RERTR) program. The fuel type selected by the program for the conversion of the five high-power research reactors in the U.S. that still use HEU fuel is a new U-Mo monolithic fuel. Studies by ORNL have previously indicated that HFIR can be successfully converted using the new fuel provided (1) the reactor power can be increased from 85 MW to 100 MW and (2) the fuel can be fabricated to a specific reference design. Fabrication techniques for the new fuel are under development by the program but are still immature, especially for the “complex” aspects of the HFIR fuel design. In FY 2012, the program underwent a major shift in focus to emphasize developing and qualifying processes for the fabrication of reliable and affordable LEU fuel. In support of this new focus and in an effort to ensure that the HFIR fuel design is as suitable for reliable fabrication as possible, ORNL undertook the present study to propose and evaluate several alternative design features. These features include (1) eliminating the fuel zone axial contouring in the previous reference design by substituting a permanent neutron absorber in the lower unfueled region of all of the fuel plates, (2) relocating the burnable neutron absorber from the fuel plates of the inner fuel element to the side plates of the inner fuel element (the fuel plates of the outer fuel element do not contain a burnable absorber), (3) relocating the fuel zone inside the fuel plate to be centered on the centerline of the depth of the plate, and (4) reshaping the radial contour of the relocated fuel zone to be symmetric about this centerline. The

  17. Preliminary Evaluation of Alternate Designs for HFIR Low-Enriched Uranium Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Renfro, David G [ORNL; Chandler, David [ORNL; Cook, David Howard [ORNL; Ilas, Germina [ORNL; Jain, Prashant K [ORNL; Valentine, Jennifer R [ORNL

    2014-11-01

    Engineering design studies of the feasibility of conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL) as part of an effort sponsored by the U.S. Department of Energy s Global Threat Reduction Initiative (GTRI)/Reduced Enrichment for Research and Test Reactors (RERTR) program. The fuel type selected by the program for the conversion of the five high-power research reactors in the U.S. that still use HEU fuel is a new U-Mo monolithic fuel. Studies by ORNL have previously indicated that HFIR can be successfully converted using the new fuel provided (1) the reactor power can be increased from 85 MW to 100 MW and (2) the fuel can be fabricated to a specific reference design. Fabrication techniques for the new fuel are under development by the program but are still immature, especially for the complex aspects of the HFIR fuel design. In FY 2012, the program underwent a major shift in focus to emphasize developing and qualifying processes for the fabrication of reliable and affordable LEU fuel. In support of this new focus and in an effort to ensure that the HFIR fuel design is as suitable for reliable fabrication as possible, ORNL undertook the present study to propose and evaluate several alternative design features. These features include (1) eliminating the fuel zone axial contouring in the previous reference design by substituting a permanent neutron absorber in the lower unfueled region of all of the fuel plates, (2) relocating the burnable neutron absorber from the fuel plates of the inner fuel element to the side plates of the inner fuel element (the fuel plates of the outer fuel element do not contain a burnable absorber), (3) relocating the fuel zone inside the fuel plate to be centered on the centerline of the depth of the plate, and (4) reshaping the radial contour of the relocated fuel zone to be symmetric about this centerline. The present

  18. Forecast of environment influence of the Ukrainian nuclear fuel plant

    International Nuclear Information System (INIS)

    Problem of site selection for the Ukrainian nuclear fuel plant is considered. Ecological influence of the site and possible contamination levels are calculated for normal and emergency situations in plant operation

  19. International symposium on uranium production and raw materials for the nuclear fuel cycle - Supply and demand, economics, the environment and energy security. Extended synopses

    International Nuclear Information System (INIS)

    The IAEA periodically organizes technical meetings and international symposia on all areas of the uranium production cycle. This publication contains 160 extended synopses related to the 2005 international symposium on 'Uranium Production and Raw Materials for the Nuclear Fuel Cycle - Supply and Demand, Economics, the Environment and Energy Security'. They cover all areas of natural uranium resources and production cycle including uranium supply and demand; uranium geology and deposit; uranium exploration; uranium mining and milling; waste management; and environment and regulation. Each synopsis was indexed individually

  20. Improvement in thermal conductivity of uranium dioxide fuel

    International Nuclear Information System (INIS)

    In order to improve the thermal conductivity of UO2 fuel, an innovative composite fuel concept of continuous metal phase with a small amount of metal in UO2 pellet, was developed. Metal candidates of W, Mo and Cr were selected, and fabrication process was conceptually designed from thermodynamic calculations. We have experimentally found that a metal phase envelops perfectly UO2 grains, forming continuous channel throughout the pellet, and improving the thermal conductivity of pellet. The fabrication process are constituted by two thermal annealing step. The microstructure characteristics results indicated that the metal phases were oxidized and melted in the first heat treatment step. The melted oxide penetrated along the grain boundary of UO2 and is interconnected with each other. Under the second reduction step, the liquid phases were transformed to solid metals, which were precipitated along the grain boundary. The thermal diffusivity of UO2-W composite is increased by about 40∼80% than that of pure UO2 sample.