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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. 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)

  5. Preliminary concepts: coordinated safeguards for materials management in a thorium--uranium fuel reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Hakkila, E.A.; Barnes, J.W.; Dayem, H.A.; Dietz, R.J.; Shipley, J.P.

    1978-10-01

    This report addresses preliminary concepts for coordinated safeguards materials management in a typical generic thorium--uranium-fueled light-water reactor (LWR) fuels reprocessing plant. The reference facility is designed to recover thorium and uranium from first-generation (denatured /sup 235/U) startup fuels, first-recycle and equilibrium (denatured /sup 233/U) thorium--uranium LWR fuels, and to recover the plutonium generated in the /sup 238/U denaturant as well. 12 figures, 3 tables.

  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. 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

  12. 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.

  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. 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)

  16. 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

  17. 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

  18. 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)

  19. 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

  20. US-Russian collaboration for enhancing nuclear materials protection, control, and accounting at the Elektrostal uranium fuel-fabrication plant

    Energy Technology Data Exchange (ETDEWEB)

    Smith, H. [Los Alamos National Lab., NM (United States); Allentuck, J. [Brookhaven National Lab., Upton, NY (United States); Barham, M. [Oak Ridge National Lab., TN (United States); Bishop, M. [Sandia National Labs., Albuquerque, NM (United States); Wentz, D. [Lawrence Livermore National Lab., CA (United States); Steele, B.; Bricker, K. [Pacific Northwest National Lab., Richland, WA (United States); Cherry, R. [USDOE, Washington, DC (United States); Snegosky, T. [Dept. of Defense, Washington, DC (United States). Defense Nuclear Agency

    1996-09-01

    In September 1993, an implementing agreement was signed that authorized collaborative projects to enhance Russian national materials control and accounting, physical protection, and regulatory activities, with US assistance funded by the Nunn-Lugar Act. At the first US-Russian technical working group meeting in Moscow in February 1994, it was decided to identify a model facility where materials protection, control, and accounting (MPC and A) and regulatory projects could be carried out using proven technologies and approaches. The low-enriched uranium (LEU or RBMK and VVER) fuel-fabrication process at Elektrostal was selected, and collaborative work began in June 1994. Based on many factors, including initial successes at Elektrostal, the Russians expanded the cooperation by proposing five additional sites for MPC and A development: the Elektrostal medium-enriched uranium (MEU or BN) fuel-fabrication process and additional facilities at Podolsk, Dmitrovgrad, Obninsk, and Mayak. Since that time, multilaboratory teams have been formed to develop and implement MPC and A upgrades at the additional sites, and much new work is underway. This paper summarizes the current status of MPC and A enhancement projects in the LEU fuel-fabrication process and discusses the status of work that addresses similar enhancements in the MEU (BN) fuel processes at Elektrostal, under the recently expanded US-Russian MPC and A cooperation.

  1. 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

  2. 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)

  3. 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

  4. 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)

  5. 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.

  6. 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)

  7. 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

  8. 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)

  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. Container concept for the PBMR fuel plant

    Energy Technology Data Exchange (ETDEWEB)

    Braehler, G.; Kelly, M.; Sauer, L.; Welbers, P. [NUKEM Technologies GmbH (Germany)

    2009-07-01

    This presentation describes the types and the characteristics of the containers uniquely designed for and used in the PBMR Fuel Plant in Pelindaba, South Africa. The PBMR Fuel Plant is designed to manufacture so called Fuel Spheres for the High Temperature Reactor which shall be built in Koeberg near Cape Town. The PBMR fuel is based on a proven, high quality German fuel design consisting of 10% enriched Uranium triple-coated isotropic (LEU-TRISO) particles contained in a moulded graphite sphere. A coated particle comprises a kernel of Uranium dioxide surrounded by different layers. A fuel sphere consists of 9 g of Uranium (some 15 000 particles) and has a diameter of 60 mm; the total mass of a fuel sphere is 210 g. During normal operation the PBMR core contains a load of 456,000 fuel spheres. The PBMR Fuel plant is designed to produce 270,000 fuel spheres a year with the option to doubling the throughput. Two basic types of containers are foreseen for the plant, i.e. the Safe Geometry Containers for transportation and storage of Uranium bearing material and Containers in non restricted geometry for solid raw and auxiliary materials required by the different processes. Unlike other concepts in fuel element plants, the Safe Geometry Containers are allowed to be stored, from a criticality point of view, everywhere in the plant. Special identification measures, physical and IT based, prevent wrong material being processed. The sub-critical configuration is still maintained after the specified fire scenario for the PBMR Fuel Plant. (orig.)

  11. 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.)

  12. 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.

  13. 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)

  14. 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.)

  15. 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)

  16. 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

  17. 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.

  18. 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)

  19. 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

  20. 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.

  1. 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

  2. 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.

  3. 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.

  4. 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

  5. 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.

  6. 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)

  7. Yalina booster subcritical assembly performance with low enriched uranium fuel

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, Alberto; Gohar, Yousry, E-mail: alby@anl.gov [Argonne National Laboratory, Lemont, IL (United States)

    2011-07-01

    The YALINA Booster facility is a subcritical assembly located in Minsk, Belarus. The facility has special features that result in fast and thermal neutron spectra in different zones. The fast zone of the assembly uses a lead matrix and uranium fuels with different enrichments: 90% and 36%, 36%, or 21%. The thermal zone of the assembly contains 10% enriched uranium fuel in a polyethylene matrix. This study discusses the performance of the three YALINA Booster configurations with the different fuel enrichments. In order to maintain the same subcriticality level in the three configurations, the number of fuel rods in the thermal zone is increased as the uranium fuel enrichment in the fast zone is decreased. The maximum number of fuel rods that can be loaded in the thermal zone is about 1185. Consequently, the neutron multiplication of the configuration with 21% enriched uranium fuel in the fast zone is enhanced by changing the position of the boron carbide and the natural uranium absorber rods, located between the fast and the thermal zones, to form an annular rather than a square arrangement. (author)

  8. 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

  9. 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

  10. 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.)

  11. 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

  12. 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.

  13. Uranium thorium dioxide fuel-cycle and economic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Gale, J.D.; Spetz, S.W. [Framatome ANP, Inc., Lynchburg, Va. (United States)

    2001-07-01

    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 UO{sub 2} and UO{sub 2}/ThO{sub 2} 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 UO{sub 2}/ThO{sub 2} cycles due primarily to the higher cost in separative work units for enriching the uranium to 19.5 wt% {sup 235}U. (author)

  14. 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...... efficiency compared with other fuels. A methanator will be suggested to be included into the plants design in order to produce methane from the fuel before entering the anode side of the SOFC stacks. Increasing methane content will decrease the needed compressor effect and thereby increase the plant power....

  15. 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

  16. 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

  17. Improving the neutronic characteristics of a boiling water reactor by using uranium zirconium hydride fuel instead of uranium dioxide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Galahom, Ahmed Abdelghafar [Higher Technological Institute, Ramadan (Egypt)

    2016-06-15

    The present work discusses two different models of boiling water reactor (BWR) bundle to compare the neutronic characteristics of uranium dioxide (UO{sub 2}) and uranium zirconium hydride (UZrH{sub 1.6}) fuel. Each bundle consists of four assemblies. The BWR assembly fueled with UO{sub 2} contains 8 × 8 fuel rods while that fueled with UZrH{sub 1.6} contains 9 × 9 fuel rods. The Monte Carlo N-Particle Transport code, based on the Mont Carlo method, is used to design three dimensional models for BWR fuel bundles at typical operating temperatures and pressure conditions. These models are used to determine the multiplication factor, pin-by-pin power distribution, axial power distribution, thermal neutron flux distribution, and axial thermal neutron flux. The moderator and coolant (water) are permitted to boil within the BWR core forming steam bubbles, so it is important to calculate the reactivity effect of voiding at different values. It is found that the hydride fuel bundle design can be simplified by eliminating water rods and replacing the control blade with control rods. UZrH{sub 1.6} fuel improves the performance of the BWR in different ways such as increasing the energy extracted per fuel assembly, reducing the uranium ore, and reducing the plutonium accumulated in the BWR through burnup.

  18. 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

  19. 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.

  20. 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...

  1. 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

  2. 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.

  3. 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.

  4. 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)

  5. 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)

  6. 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

  7. 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

  8. 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

  9. 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.

  10. 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)

  11. 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

  12. 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

  13. 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...

  14. 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...

  15. 77 FR 65729 - Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services LLC...

    Science.gov (United States)

    2012-10-30

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

  16. 78 FR 23312 - Uranium Enrichment Fuel Cycle Inspection Reports Regarding Louisiana Energy Services, National...

    Science.gov (United States)

    2013-04-18

    ... COMMISSION Uranium Enrichment Fuel Cycle Inspection Reports Regarding Louisiana Energy Services, National... 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..., Uranium Enrichment Branch, Division of Fuel Cycle Safety, and Safeguards Office of Nuclear Material...

  17. 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.

  18. 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.

  19. 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.

  20. 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

  1. 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

  2. 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.

  3. 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)

  4. 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)

  5. 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

  6. 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

  7. 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)

  8. VASCULAR FLORA OF ALABASTER CAVERNS STATE PARK, CIMARRON GYPSUM HILLS: WOODWARD COUNTY, OKLAHOMA

    Directory of Open Access Journals (Sweden)

    Kristi D. Rice

    2012-12-01

    Full Text Available Alabaster Caverns State Park is located in the Cimarron Gypsum Hills of northwestern Oklahoma, a semi-arid region of the state. The majority of the park is dominated by mixed-grass prairie and gypsum outcrops, with some riparian habitat and wooded north-facing slopes. A vascular plant inventory conducted from 2004 through 2007 yielded 274 species in 199 genera and 66 families. The largest families were the Poaceae (52 species, Asteraceae (47, and Fabaceae (23. There were 100 annuals, 6 biennials, and 163 perennials, as well as 5 species that have more than one life history form. Forty-two species (15.3% were not native to North America. Three taxa currently being tracked by the Oklahoma Natural Heritage Inventory (2012 were present: Echinocereus reichenbachii (S3G5, Haploesthes greggii (S1G4?, and Marsilea vestita (S1G5. Compared to floristic inventories of sites in the Cimarron Gypsum Hills that are less impacted by public visitation, but more intensively grazed, Alabaster Caverns State Park has a higher number of species as well as a higher proportion of introduced species.

  9. 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)

  10. 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

  11. 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

  12. 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.

  13. 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)

  14. 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.

  15. 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

  16. Minor isotope measurements for safeguarding a uranium enrichment plant

    Energy Technology Data Exchange (ETDEWEB)

    Blumkin, S.; Levin, S.A.; Von Halle, E.

    1979-03-26

    A study of the behavior of the minor uranium isotopes in enrichment cascades indicates that the measurement of their concetrations in cascade feed and withdrawal streams can be used to confirm or deny material accountability results in safeguarding an enrichment plant. In further support of safeguard measures, an indirect non-intrusive method to measure cascade uranium inventory has been devised and tested satisfactorily in a gaseous diffusion cascade.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-02-01

    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.

  18. 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)

  19. 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

  20. 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.

  1. 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)

  2. 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.

  3. 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

  4. 75 FR 38809 - Southern Turner Cimarron I, LLC; Notice of Filing

    Science.gov (United States)

    2010-07-06

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal 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...

  5. 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.

  6. 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.

  7. 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)

  8. 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 ...

  9. 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

  10. 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

  11. 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)

  12. 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

  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. 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

  15. 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

  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. 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

  18. Conceptual Core Analysis of Long Life PWR Utilizing Thorium-Uranium Fuel Cycle

    Science.gov (United States)

    Rouf; Su'ud, Zaki

    2016-08-01

    Conceptual core analysis of long life PWR utilizing thorium-uranium based fuel has conducted. The purpose of this study is to evaluate neutronic behavior of reactor core using combined thorium and enriched uranium fuel. Based on this fuel composition, reactor core have higher conversion ratio rather than conventional fuel which could give longer operation length. This simulation performed using SRAC Code System based on library SRACLIB-JDL32. The calculation carried out for (Th-U)O2 and (Th-U)C fuel with uranium composition 30 - 40% and gadolinium (Gd2O3) as burnable poison 0,0125%. The fuel composition adjusted to obtain burn up length 10 - 15 years under thermal power 600 - 1000 MWt. The key properties such as uranium enrichment, fuel volume fraction, percentage of uranium are evaluated. Core calculation on this study adopted R-Z geometry divided by 3 region, each region have different uranium enrichment. The result show multiplication factor every burn up step for 15 years operation length, power distribution behavior, power peaking factor, and conversion ratio. The optimum core design achieved when thermal power 600 MWt, percentage of uranium 35%, U-235 enrichment 11 - 13%, with 14 years operation length, axial and radial power peaking factor about 1.5 and 1.2 respectively.

  19. 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

  20. 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.

  1. 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

  2. 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.

  3. 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)

  4. 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

  5. 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

  6. 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

  7. Preliminary Design Study of Medium Sized Gas Cooled Fast Reactor with Natural Uranium as Fuel Cycle Input

    Science.gov (United States)

    Meriyanti, Su'ud, Zaki; Rijal, K.; Zuhair, Ferhat, A.; Sekimoto, H.

    2010-06-01

    In this study a fesibility design study of medium sized (1000 MWt) gas cooled fast reactors which can utilize natural uranium as fuel cycle input has been conducted. Gas Cooled Fast Reactor (GFR) is among six types of Generation IV Nuclear Power Plants. GFR with its hard neuron spectrum is superior for closed fuel cycle, and its ability to be operated in high temperature (850° C) makes various options of utilizations become possible. To obtain the capability of consuming natural uranium as fuel cycle input, modified CANDLE burn-up scheme[1-6] is adopted this GFR system by dividing the core into 10 parts of equal volume axially. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. As an optimization results, a design of 1000 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input is discussed. The average discharge burn-up is about 280 GWd/ton HM. Enough margin for criticallity was obtained for this reactor.

  8. 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.

  9. 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.

  10. 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

  11. 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)

  12. 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

  13. 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

  14. 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.)

  15. 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

  16. 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

  17. 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...

  18. 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...

  19. 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)

  20. 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

  1. 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

  2. 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

  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. 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, puranium concentration in plant and the substrate (r=0.88, puranium concentration was significantly correlated with Mn, Fe, and Ni concentration of plants (puranium accumulation and concentration ratio can be useful for prospecting phytoremediation of stream carrying treated or untreated uranium mill tailings effluent.

  5. 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)

  6. 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

  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. 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)

  9. 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.

  10. 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

  11. 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.

  12. 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)

  13. 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)

  14. Pilot plant operation of the Uranium Chip Oxidation Facility at the Y-12 Plant

    Energy Technology Data Exchange (ETDEWEB)

    Childs, Y.C.

    1987-01-16

    Due to changing environmental regulations, the current practice of depleted uranium chip (machine turning) disposal via shallow land burial has become environmentally objectionable. The chips are pyrophoric and oxidize rapidly when exposed to air; therefore, long-term storage of the uranium chips presents a major fire hazard. The Oak Ridge Y-12 Plant Development Division was contacted to devise a disposal method that would eliminate chip burial and minimize storage space requirements. The proposed method of accomplishing this task was oxidizing the uranium chips to uranium oxide (U/sub 3/O/sub 8/) under controlled conditions. Pilot plant operation of the Uranium Chip Oxidation Facility (UCOF) was initiated on May 20, 1985, by the Y-12 Development Division. The purpose of this initial development testing was to evaluate the equipment, determine operating parameters, and provide on-the-job training for Waste Treatment Operations (WTO) personnel. Startup of the UCOF began with the check-out of the equipment using only the No. 1 oxidizer. Following the verification stage, the oxidizer was loaded with an initial charge of cold uranium oxide (U/sub 3/O/sub 8/) in preparation for test burning. Results of the test are given.

  15. 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

  16. 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.)

  17. 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)

  18. 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)

  19. 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)

  20. 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)

  1. 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

  2. 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

  3. 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.

  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. 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

  6. 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

  7. 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...

  8. 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...

  9. 77 FR 23476 - Cimarron Windpower II, LLC; Supplemental Notice That Initial Market-Based Rate Filing Includes...

    Science.gov (United States)

    2012-04-19

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Cimarron Windpower II, LLC; Supplemental Notice That Initial Market-Based... above-referenced proceeding of Cimarron Windpower II, LLC's application for market-based rate...

  10. 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... supplemental notice in the above-referenced proceeding of Southern Turner Cimarron I, LLC's application...

  11. 2010 Status of Uranium Conversion Plant Decommissioning Project

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, D. S.; Lee, K. I.; Choi, Y. D.; Chung, U. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    The Uranium Conversion Plant (UCP) was used to manufacture 100 tons of UO{sub 2}. 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

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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.

  17. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Fisher, D.R. (Pacific Northwest Lab., Richland, WA (USA)); Swint, M.J.; Kathren, R.L. (Hanford Environmental Health Foundation, Richland, WA (USA))

    1990-05-01

    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 (UF{sub 6}) 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.

  19. 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

  20. 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)

  1. 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)

  2. 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.)

  3. 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)

  4. 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)

  5. 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.

  6. 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

  7. 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…

  8. Experiments on light water lattices with enriched uranium fuel; Analyse des donnees experimentales sur les reseaux a eau legere et uranium enrichi

    Energy Technology Data Exchange (ETDEWEB)

    Audinet, M. [Societe des Forges et Ateliers du Creusot, 75 - Paris (France); Lamare, J. de [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Panossian, J. [Societe Alsacienne de Constructions Mecaniques (France)

    1958-07-01

    Experiments a light water lattices with slightly enriched uranium fuel, have been performed at Brookhaven and Bettis Plant Laboratories. The results are studied and compared with simple theories on reactor calculations. By taking into account shadow effects and non Maxwellian neutron spectrum, which are important in this kind of reactors, we have been able to explain the observed results fairly well. We can thus give a constituent set of formulas with which to calculate lattices similar to there we studied. (author) [French] Les resultats d'experiences effectuees aux Laboratoires de Brookbaven et de Bettis Plant, sur des reseaux heterogenes a eau legere et uranium metallique legerement enrichi, sont analyses et confrontes avec les theories simples du calcul de pile. En tenant compte des effets d'interaction et d'echauffement du spectre de neutrons qui sont importants dans ce type de reacteurs, on parvient a rendre compte convenablement des resultats observes. On a ainsi mis au point un formulaire permettant le calcul des reseaux quivpeuvent etre consideres comme assez semblables aux reseaux etudies. (auteur)

  9. 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.

  10. 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)

  11. 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.

  12. 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.

  13. 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

  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. 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

  16. 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

  17. 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

  18. 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

  19. 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. ...

  20. 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.)

  1. 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)

  2. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Blum, S.R.; McLaren, R.A.

    1979-10-01

    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, U/sup 235/ 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.

  4. 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)

  5. 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.

  6. 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)

  7. 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

  8. Low-temperature irradiation behavior of uranium-molybdenum alloy dispersion fuel

    Science.gov (United States)

    Meyer, M. K.; Hofman, G. L.; Hayes, S. L.; Clark, C. R.; Wiencek, T. C.; Snelgrove, J. L.; Strain, R. V.; Kim, K.-H.

    2002-08-01

    Irradiation tests have been conducted to evaluate the performance of a series of high-density uranium-molybdenum (U-Mo) alloy, aluminum matrix dispersion fuels. Fuel plates incorporating alloys with molybdenum content in the range of 4-10 wt% were tested. Two irradiation test vehicles were used to irradiate low-enrichment fuels to approximately 40 and 70 at.% 235U burnup in the advanced test reactor at fuel temperatures of approximately 65 °C. The fuel particles used to fabricate dispersion specimens for most of the test were produced by generating filings from a cast rod. In general, fuels with molybdenum contents of 6 wt% or more showed stable in-reactor fission gas behavior, exhibiting a distribution of small, stable gas bubbles. Fuel particle swelling was moderate and decreased with increasing alloy content. Fuel particles with a molybdenum content of 4 wt% performed poorly, exhibiting extensive fuel-matrix interaction and the growth of relatively large fission gas bubbles. Fuel particles with 4 or 6 wt% molybdenum reacted more rapidly with the aluminum matrix than those with higher-alloy content. Fuel particles produced by an atomization process were also included in the test to determine the effect of fuel particle morphology and microstructure on fuel performance for the U-10Mo composition. Both of the U-10Mo fuel particle types exhibited good irradiation performance, but showed visible differences in fission gas bubble nucleation and growth behavior.

  9. 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)

  10. 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

  11. 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...

  12. 10 CFR Appendix J to Part 110 - Illustrative List of Uranium Conversion Plant Equipment and Plutonium Conversion Plant Equipment...

    Science.gov (United States)

    2010-01-01

    ... and Plutonium Conversion Plant Equipment Under NRC Export Licensing Authority J Appendix J to Part 110... Plutonium Conversion Plant Equipment Under NRC Export Licensing Authority Note—Uranium conversion plants and... feed for electromagnetic enrichment. Note: Plutonium conversion plants and systems may perform one...

  13. 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

  14. 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

  15. 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

  16. 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

  17. 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)

  18. Feasibility study of boiling water reactor core based on thorium-uranium fuel concept

    Energy Technology Data Exchange (ETDEWEB)

    Nunez-Carrera, Alejandro [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779, Col Narvarte, 03020 Mexico D.F. (Mexico); Francois Lacouture, Juan Luis; Martin del Campo, Cecilia [Universidad Nacional Autonoma de Mexico, Facultad de Ingenieria, Paseo Cuauhnahuac 8532, Jiutepec, Mor. (Mexico); Espinosa-Paredes, Gilberto [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana Iztapalapa, Apartado Postal 55-534, Mexico D.F. 09340 (Mexico)], E-mail: gepe@xanum.uam.mx

    2008-01-15

    The design of a boiling water reactor (BWR) equilibrium core using the thorium-uranium (blanket-seed) concept in the same integrated fuel assembly is presented in this paper. The lattice design uses the thorium conversion capability to {sup 233}U in a BWR spectrum. A core design was developed to achieve an equilibrium cycle of one effective full power year in a standard BWR with a reload of 104 fuel assemblies designed with an average {sup 235}U enrichment of 7.5 w/o in the seed sub-lattice. The main core operating parameters were obtained. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The economic analysis shows that the fuel cycle cost of the proposed core design can be competitive with a standard uranium core design. Finally, a comparison of the toxicity of the spent fuel showed that the toxicity is lower in the thorium cycle than in other fuel cycles (UO{sub 2} and MOX uranium and plutonium) in the case of the once through cycle for light water reactors (LWR)

  19. 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.

  20. 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)

  1. 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)

  2. 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

  3. 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)

  4. 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)

  5. 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

  6. Raman spectroscopic investigation of thorium dioxide-uranium dioxide (ThO₂-UO₂) fuel materials.

    Science.gov (United States)

    Rao, Rekha; Bhagat, R K; Salke, Nilesh P; Kumar, Arun

    2014-01-01

    Raman spectroscopic investigations were carried out on proposed nuclear fuel thorium dioxide-uranium dioxide (ThO2-UO2) solid solutions and simulated fuels based on ThO2-UO2. Raman spectra of ThO2-UO2 solid solutions exhibited two-mode behavior in the entire composition range. Variations in mode frequencies and relative intensities of Raman modes enabled estimation of composition, defects, and oxygen stoichiometry in these compounds that are essential for their application. The present study shows that Raman spectroscopy is a simple, promising analytical tool for nondestructive characterization of this important class of nuclear fuel materials.

  7. 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)

  8. 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

  9. 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

  10. 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.

  11. 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

  12. 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

  13. Plant microbial fuel cell applied in wetlands

    NARCIS (Netherlands)

    Wetser, Koen; Liu, Jia; Buisman, Cees; Strik, David

    2015-01-01

    The plant microbial fuel cell (PMFC) has to be applied in wetlands to be able to generate electricity on a large scale. The objective of this PMFC application research is to clarify the differences in electricity generation between a Spartina anglica salt marsh and Phragmites australis peat soil

  14. 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

  15. 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)

  16. 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...

  17. 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

  18. 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)

  19. Processing of mixed uranium/refractory metal carbide fuels for high temperature space nuclear reactors

    Science.gov (United States)

    Knight, Travis; Anghaie, Samim

    2000-01-01

    Single phase, solid-solution mixed uranium/refractory metal carbides have been proposed as an advanced nuclear fuel for high performance, next generation space power and propulsion systems. These mixed carbides such as the pseudo-ternary, (U, Zr, Nb)C, hold significant promise because of their high melting points (typically greater than 3200 K), thermochemical stability in a hot hydrogen environment, and high thermal conductivity. However, insufficient test data exist under nuclear thermal propulsion conditions of temperature and hot hydrogen environment to fully evaluate their performance. Various compositions of (U, Zr, Nb)C were processed with 5% and 10% metal mole fraction of uranium. Stoichiometric samples were processed from the constituent carbide powders while hypostoichiometric samples with carbon-to-metal (C/M) ratios of 0.95 were processed from uranium hydride, graphite, and constituent refractory carbide powders. Processing techniques of cold pressing, sintering, and hot pressing were investigated to optimize the processing parameters necessary to produce dense (low porosity), homogeneous, single phase, solid-solution mixed carbide nuclear fuels for testing. This investigation was undertaken to evaluate and characterize the performance of these mixed uranium/refractory metal carbides for space power and propulsion applications. .

  20. 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.)

  1. Containment and storage of uranium hexafluoride at US Department of Energy uranium enrichment plants

    Energy Technology Data Exchange (ETDEWEB)

    Barlow, C.R.; Alderson, J.H.; Blue, S.C.; Boelens, R.A.; Conkel, M.E.; Dorning, R.E.; Ecklund, C.D.; Halicks, W.G.; Henson, H.M.; Newman, V.S.; Philpot, H.E.; Taylor, M.S.; Vournazos, J.P. [Oak Ridge K-25 Site, TN (United States). UEO Enrichment Technical Operations Div.; Russell, J.R. [USDOE Oak Ridge Field Office, TN (United States); Pryor, W.A. [PAI Corp., Oak Ridge, TN (United States); Ziehlke, K.T. [MJB Technical Associates (United States)

    1992-07-01

    Isotopically depleted UF{sub 6} (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 UF{sub 6}. 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.

  2. 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

  3. 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

  4. 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.)

  5. 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

  6. Radiological health aspects of commercial uranium conversion, enrichment, and fuel fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Stoetzel, G.A.; Hoenes, G.R.; Cummings, F.M.; McCormack, W.D.

    1982-11-01

    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.

  7. Standard test method for atom percent fission in uranium and plutonium fuel (Neodymium-148 Method)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    1996-01-01

    1.1 This test method covers the determination of stable fission product 148Nd in irradiated uranium (U) fuel (with initial plutonium (Pu) content from 0 to 50 %) as a measure of fuel burnup (1-3). 1.2 It is possible to obtain additional information about the uranium and plutonium concentrations and isotopic abundances on the same sample taken for burnup analysis. If this additional information is desired, it can be obtained by precisely measuring the spike and sample volumes and following the instructions in Test Method E267. 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

  8. 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)

  9. 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

  10. 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.)

  11. 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

  12. 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.

  13. 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

  14. 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

  15. 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)

  16. Evaluating the effectiveness of dilution of the recovered uranium with depleted uranium and low-enriched uranium to obtain fuel for VVER reactors

    Science.gov (United States)

    Smirnov, A. Yu; Sulaberidze, G. A.; Dudnikov, A. A.; Nevinitsa, V. A.

    2016-09-01

    The possibility of the recovered uranium enrichment in a cascade of gas centrifuges with three feed flows (depleted uranium, low-enriched uranium, recovered uranium) with simultaneous dilution of U-232,234,236 isotopes was shown. A series of numerical experiments were performed for different content of U-235 in low-enriched uranium. It has been demonstrated that the selected combination of diluents can simultaneously reduce the cost of separative work and the consumption of natural uranium, not only with respect to the previously used multi-flow cascade schemes, but also in comparison to the standard cascade for uranium enrichment.

  17. 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)

  18. 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

  19. 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

  20. Microbeam x-ray absorption spectroscopy study of chromium in large-grain uranium dioxide fuel

    Science.gov (United States)

    Mieszczynski, C.; Kuri, G.; Bertsch, J.; Martin, M.; Borca, C. N.; Delafoy, Ch; Simoni, E.

    2014-09-01

    Synchrotron-based microprobe x-ray absorption spectroscopy (XAS) has been used to study the local atomic structure of chromium in chromia-doped uranium dioxide (UO2) grains. The specimens investigated were a commercial grade chromia-doped UO2 fresh fuel pellet, and materials from a spent fuel pellet of the same batch, irradiated with an average burnup of ~40 MW d kg-1. Uranium L3-edge and chromium K-edge XAS have been measured, and the structural environments of central uranium and chromium atoms have been elucidated. The Fourier transform of uranium L3-edge extended x-ray absorption fine structure shows two well-defined peaks of U-O and U-U bonds at average distances of 2.36 and 3.83 Å. Their coordination numbers are determined as 8 and 11, respectively. The chromium Fourier transform extended x-ray absorption fine structure of the pristine UO2 matrix shows similar structural features with the corresponding spectrum of the irradiated spent fuel, indicative of analogous chromium environments in the two samples studied. From the chromium XAS experimental data, detectable next neighbor atoms are oxygen and uranium of the cation-substituted UO2 lattice, and two distinct subshells of chromium and oxygen neighbors, possibly because of undissolved chromia particles present in the doped fuels. Curve-fitting analyses using theoretical amplitude and phase-shift functions of the closest Cr-O shell and calculations with ab initio computer code FEFF and atomic clusters generated from the chromium-dissolved UO2 structure have been carried out. There is a prominent reduction in the length of the adjacent Cr-O bond of about 0.3 Å in chromia-doped UO2 compared with the ideal U-O bond length in standard UO2 that would be expected because of the change in effective Coulomb interactions resulting from replacing U4+ with Cr3+ and their ionic size differences. The contraction of shortest Cr-U bond is ~0.1 Å relative to the U-U bond length in bulk UO2. The difference in the

  1. 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.

  2. 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.

  3. 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)

  4. 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

  5. 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.

  6. 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...

  7. 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

  8. 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

  9. 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

  10. Transient and stability analysis of a BWR core with thorium-uranium fuel

    Energy Technology Data Exchange (ETDEWEB)

    Nunez-Carrera, Alejandro [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779 Col. Narvarte, 03020 Mexico, DF (Mexico); Espinosa-Paredes, Gilberto [Division de Ciencias Basicas e Ingenieria, Universidad Autonoma Metropolitana, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico, DF (Mexico)], E-mail: gepe@xanum.uam.mx; Francois, Juan-Luis [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532, 62550 Jiutepec Mor. (Mexico)

    2008-08-15

    The kinetic response of a boiling water reactor (BWR) equilibrium core using thorium as a nuclear material, in an integrated blanket-seed assembly, is presented in this work. Additionally an in-house code was developed to evaluate this core under steady state and transient conditions including a stability analysis. The code has two modules: (a) the time domain module for transient analysis and (b) the frequency domain module for stability analysis. The thermal-hydraulic process is modeled by a set of five equations, considering no homogeneous flow with drift-flux approximation and non-equilibrium thermodynamic. The neutronic process is calculated with a point kinetics model. Typical BWR reactivity effects are considered: void fraction, fuel temperature, moderator temperature and control rod density. Collapsed parameters were included in the code to represent the core using an average fuel channel. For the stability analysis, in the frequency domain, the transfer function is determined by applying Laplace-transforming to the calculated pressure drop perturbations in each of the considered regions where a constant total pressure drop was considered. The transfer function was used to study the system response in the frequency domain when an inlet flow perturbation is applied. The results show that the neutronic behavior of the core with thorium uranium fuel is similar to a UO{sub 2} core, even during transient conditions. The stability and transient analysis show that the thorium-uranium fuel can be operated safely in current BWRs.

  11. High-uranium-loaded U3O8-Al fuel element development program [contributed by N.M. Martin, ORNL

    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 % involve plate fabrication phenomena of voids and dogboning, fuel behavior under long irradiation, and potential for the thermite reaction between U3O8 and aluminum. (author)

  12. Fusion solution to dispose of spent nuclear fuel, transuranic elements, and highly enriched uranium

    Energy Technology Data Exchange (ETDEWEB)

    Gohar, Yousry E-mail: gohar@anl.gov

    2001-11-01

    The disposal of the nuclear spent fuel, the transuranic elements, and the highly enriched uranium represents a major problem under investigation by the international scientific community to identify the most promising solutions. The investigation of this paper focused on achieving the top rated solution for the problem, the elimination goal, which requires complete elimination for the transuranic elements or the highly enriched uranium, and the long-lived fission products. To achieve this goal, fusion blankets with liquid carrier, molten salts or liquid metal eutectics, for the transuranic elements and the uranium isotopes are utilized. The generated energy from the fusion blankets is used to provide revenue for the system. The long-lived fission products are fabricated into fission product targets for transmutation utilizing the neutron leakage from the fusion blankets. This paper investigated the fusion blanket designs for small fusion devices and the system requirements for such application. The results show that 334 MW of fusion power from D-T plasma for 30 years with an availability factor of 0.75 can dispose of the 70,000 tons of the U.S. inventory of spent nuclear fuel generated up to the year 2015. In addition, this fusion solution eliminates the need for a geological repository site, which is a major advantage. Meanwhile, such utilization of the fusion power will provide an excellent opportunity to develop fusion energy for the future.

  13. Role of thermal analysis in uranium oxide fuel fabrication process

    International Nuclear Information System (INIS)

    The present paper discusses the application of thermal analysis, particularly, differential thermal analysis (Dta) at various stages of fuel fabrication process. The useful role of Dta in knowing the decomposition pattern and calcination temperature of Adu along with de-nitration temperature is explained. The decomposition pattern depends upon the type of drying process adopted for wet ADU cake (ADU C). Also, the paper highlights the utility of DTA in determining the APS and SSA of UO2+x and U3O8 powders as an alternate technique. Further, the temperature difference (ΔTmax) between the two exothermic peaks obtained in UO2+x powder oxidation is related to sintered density of UO2 pellets. (author)

  14. Data base for a CANDU-PHW operating on a once-through, natural uranium fuel cycle

    International Nuclear Information System (INIS)

    This report, prepared for INFCE, describes a standard 600 MW(e) CANDU-PHW reactor operating on a once-through natural uranium fuel cycle. Subsequently, data are given for an extrapolated 1000 MW(e) design (the nominal capacity adopted for the INFCE study) operating on the same fuel cycle. (author)

  15. Powder Metallurgy of Uranium Alloy Fuels for TRU-Burning Reactors Final Technical Report

    International Nuclear Information System (INIS)

    beginning of the materials processing setup. Also included within this section is a thesis proposal by Jeff Hausaman. Appendix C contains the public papers and presentations introduced at the 2010 American Nuclear Society Winter Meeting. Appendix A - MSNE theses of David Garnetti and Grant Helmreich and proposal by Jeff Hausaman A.1 December 2009 Thesis by David Garnetti entitled 'Uranium Powder Production Via Hydride Formation and Alpha Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications' A.2 September 2009 Presentation by David Garnetti (same title as document in Appendix B.1) A.3 December 2010 Thesis by Grant Helmreich entitled 'Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications' A.4 October 2010 Presentation by Grant Helmreich (same title as document in Appendix B.3) A.5 Thesis Proposal by Jeffrey Hausaman entitled 'Hot Extrusion of Alpha Phase Uranium-Zirconium Alloys for TRU Burning Fast Reactors' Appendix B - External presentations introduced at the 2010 ANS Winter Meeting B.1 J.S. Hausaman, D.J. Garnetti, and S.M. McDeavitt, 'Powder Metallurgy of Alpha Phase Uranium Alloys for TRU Burning Fast Reactors,' Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.2 PowerPoint Presentation Slides from C.1 B.3 G.W. Helmreich, W.J. Sames, D.J. Garnetti, and S.M. McDeavitt, 'Uranium Powder Production Using a Hydride-Dehydride Process,' Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.4. PowerPoint Presentation Slides from C.3 B.5 Poster Presentation from C.3 Appendix C - Fuel cycle research and development undergraduate materials and poster presentation C.1 Poster entitled 'Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys' presented at the Fuel Cycle Technologies Program Annual Meeting C.2 April 2011 Honors Undergraduate Thesis by William Sames, Research Fellow

  16. Powder Metallurgy of Uranium Alloy Fuels for TRU-Burning Reactors Final Technical Report

    Energy Technology Data Exchange (ETDEWEB)

    McDeavitt, Sean M

    2011-04-29

    outlining the beginning of the materials processing setup. Also included within this section is a thesis proposal by Jeff Hausaman. Appendix C contains the public papers and presentations introduced at the 2010 American Nuclear Society Winter Meeting. Appendix A—MSNE theses of David Garnetti and Grant Helmreich and proposal by Jeff Hausaman A.1 December 2009 Thesis by David Garnetti entitled “Uranium Powder Production Via Hydride Formation and Alpha Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications” A.2 September 2009 Presentation by David Garnetti (same title as document in Appendix B.1) A.3 December 2010 Thesis by Grant Helmreich entitled “Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications” A.4 October 2010 Presentation by Grant Helmreich (same title as document in Appendix B.3) A.5 Thesis Proposal by Jeffrey Hausaman entitled “Hot Extrusion of Alpha Phase Uranium-Zirconium Alloys for TRU Burning Fast Reactors” Appendix B—External presentations introduced at the 2010 ANS Winter Meeting B.1 J.S. Hausaman, D.J. Garnetti, and S.M. McDeavitt, “Powder Metallurgy of Alpha Phase Uranium Alloys for TRU Burning Fast Reactors,” Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.2 PowerPoint Presentation Slides from C.1 B.3 G.W. Helmreich, W.J. Sames, D.J. Garnetti, and S.M. McDeavitt, “Uranium Powder Production Using a Hydride-Dehydride Process,” Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.4. PowerPoint Presentation Slides from C.3 B.5 Poster Presentation from C.3 Appendix C—Fuel cycle research and development undergraduate materials and poster presentation C.1 Poster entitled “Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys” presented at the Fuel Cycle Technologies Program Annual Meeting C.2 April 2011 Honors Undergraduate Thesis

  17. Powder Metallurgy of Uranium Alloy Fuels for TRU-Burning Reactors Final Technical Report

    Energy Technology Data Exchange (ETDEWEB)

    McDeavitt, Sean M

    2011-04-29

    outlining the beginning of the materials processing setup. Also included within this section is a thesis proposal by Jeff Hausaman. Appendix C contains the public papers and presentations introduced at the 2010 American Nuclear Society Winter Meeting. Appendix A—MSNE theses of David Garnetti and Grant Helmreich and proposal by Jeff Hausaman A.1 December 2009 Thesis by David Garnetti entitled “Uranium Powder Production Via Hydride Formation and Alpha Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications” A.2 September 2009 Presentation by David Garnetti (same title as document in Appendix B.1) A.3 December 2010 Thesis by Grant Helmreich entitled “Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications” A.4 October 2010 Presentation by Grant Helmreich (same title as document in Appendix B.3) A.5 Thesis Proposal by Jeffrey Hausaman entitled “Hot Extrusion of Alpha Phase Uranium-Zirconium Alloys for TRU Burning Fast Reactors” Appendix B—External presentations introduced at the 2010 ANS Winter Meeting B.1 J.S. Hausaman, D.J. Garnetti, and S.M. McDeavitt, “Powder Metallurgy of Alpha Phase Uranium Alloys for TRU Burning Fast Reactors,” Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.2 PowerPoint Presentation Slides from C.1 B.3 G.W. Helmreich, W.J. Sames, D.J. Garnetti, and S.M. McDeavitt, “Uranium Powder Production Using a Hydride-Dehydride Process,” Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.4. PowerPoint Presentation Slides from C.3 B.5 Poster Presentation from C.3 Appendix C—Fuel cycle research and development undergraduate materials and poster presentation C.1 Poster entitled “Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys” presented at the Fuel Cycle Technologies Program Annual Meeting C.2 April 2011 Honors Undergraduate Thesis

  18. Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Cowell, B.S.; Fisher, S.E.

    1999-02-01

    The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option.

  19. Enhanced Low-Enriched Uranium Fuel Element for the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pope, M. A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); DeHart, M. D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Morrell, S. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Jamison, R. K. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nef, E. C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nigg, D. W. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    Under the current US Department of Energy (DOE) policy and planning scenario, the Advanced Test Reactor (ATR) and its associated critical facility (ATRC) will be reconfigured to operate on low-enriched uranium (LEU) fuel. This effort has produced a conceptual design for an Enhanced LEU Fuel (ELF) element. This fuel features monolithic U-10Mo fuel foils and aluminum cladding separated by a thin zirconium barrier. As with previous iterations of the ELF design, radial power peaking is managed using different U-10Mo foil thicknesses in different plates of the element. The lead fuel element design, ELF Mk1A, features only three fuel meat thicknesses, a reduction from the previous iterations meant to simplify manufacturing. Evaluation of the ELF Mk1A fuel design against reactor performance requirements is ongoing, as are investigations of the impact of manufacturing uncertainty on safety margins. The element design has been evaluated in what are expected to be the most demanding design basis accident scenarios and has met all initial thermal-hydraulic criteria.

  20. Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel

    International Nuclear Information System (INIS)

    The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option

  1. Fabrication of uranium-plutonium mixed nitride fuel pins (89F-3A) for second irradiation test at JMTR

    International Nuclear Information System (INIS)

    A couple of fuel pins of uranium-plutonium mixed nitride, which has a potential as an advanced FBR fuel, was fabricated for the 2nd irradiation test at JMTR in order to obtain informations on the fuel behavior and to establish the fuel performance. A ferritic stainless steel-cladded fuel pin was adopted to depress the mechanical interaction of nitride fuels with cladding materials (FCMI). A comparative examination is planned by the combined use of an austenitic stainless steel-cladded pin. The irradiation of these pins started on January 1991 for the goal burnup of about 50 GWd/t. (author)

  2. Gas-flow detector for uranium contamination on finned-can surface of a reactor fuel

    International Nuclear Information System (INIS)

    This gas-flow detector is a gridded proportional-counter specially designed for detecting uranium contamination on the finned-can surface of a reactor fuel. A conventional proportional-counter constructed only by a cathode and collector can hardly detect alpha particles emitted from uranium which is contaminated on rugged surfaces such as those of the fins of a reactor-fuel can because of the lack of uniformity of the electric field near the surface. This is the reason why we have constructed the gridded proportional-counter. This counter comprises the fuel, a grid, collectors and a cathode which are cylindrical in construction and arranged coaxially. The fuel is placed in the centre of the grid and negative voltage is applied. The space between the fuel and the grid serves as an ion-collecting space. The grid is made of fine parallel tungsten wires which are constructed cylindrically around the fuel and connected to ground potential. The collectors are 16 fine tungsten wires constructed similarly to the grid, but each wire is electrically insulated from the others. Through 50-kΩ resistors all collectors are connected together and to positive high voltage via a feeding resistor. The space between the grid, the collectors and the cathode serves as a gas-multiplication space just like a conventional proportional-counter. Each 50-kΩ resistor separates the stray capacity of the connected collector from the others. The detector output is coupled with a low input-impedance-current amplifier. The low inputimpedance also lessens the bad influence of the stray capacity of the input circuitry. These result in a good S/N ratio and allow the sensitive detection of alpha particles. Before measurement, the counter is first evacuated by a rotary pump and then PR gas (A: 90%, Methane: 10%) is admitted. By the use of this new equipment we have successfully detected alpha particles emitted from 1 x 10-5 g natural-uranium contamination of the finned-can surface of a Calder Hall

  3. Fabrication of high-uranium-loaded U/sub 3/O/sub 8/-Al developmental fuel plates

    Energy Technology Data Exchange (ETDEWEB)

    Copeland, G.L.; Martin, M.M.

    1980-12-01

    A common plate-type fuel for research and test reactors is U/sub 3/O/sub 8/ dispersed in aluminum and clad with an aluminum alloy. There is an impetus to reduce the /sup 235/U 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.

  4. 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

  5. Processing used nuclear fuel with nanoscale control of uranium and ultrafiltration

    Science.gov (United States)

    Wylie, Ernest M.; Peruski, Kathryn M.; Prizio, Sarah E.; Bridges, Andrea N. A.; Rudisill, Tracy S.; Hobbs, David T.; Phillip, William A.; Burns, Peter C.

    2016-05-01

    Current separation and purification technologies utilized in the nuclear fuel cycle rely primarily on liquid-liquid extraction and ion-exchange processes. Here, we report a laboratory-scale aqueous process that demonstrates nanoscale control for the recovery of uranium from simulated used nuclear fuel (SIMFUEL). The selective, hydrogen peroxide induced oxidative dissolution of SIMFUEL material results in the rapid assembly of persistent uranyl peroxide nanocluster species that can be separated and recovered at moderate to high yield from other process-soluble constituents using sequestration-assisted ultrafiltration. Implementation of size-selective physical processes like filtration could results in an overall simplification of nuclear fuel cycle technology, improving the environmental consequences of nuclear energy and reducing costs of processing.

  6. RUSSIAN-ORIGIN HIGHLY ENRICHED URANIUM SPENT NUCLEAR FUEL SHIPMENT FROM BULGARIA

    Energy Technology Data Exchange (ETDEWEB)

    Kelly Cummins; Igor Bolshinsky; Ken Allen; Tihomir Apostolov; Ivaylo Dimitrov

    2009-07-01

    In July 2008, the Global Threat Reduction Initiative and the IRT 2000 research reactor in Sofia, Bulgaria, operated by the Institute for Nuclear Research and Nuclear Energy (INRNE), safely shipped 6.4 kilograms of Russian origin highly enriched uranium (HEU) spent nuclear fuel (SNF) to the Russian Federation. The shipment, which resulted in the removal of all HEU from Bulgaria, was conducted by truck, barge, and rail modes of transport across two transit countries before reaching the final destination at the Production Association Mayak facility in Chelyabinsk, Russia. This paper describes the work, equipment, organizations, and approvals that were required to complete the spent fuel shipment and provides lessons learned that might assist other research reactor operators with their own spent nuclear fuel shipments.

  7. Evaluation of environmental control technologies for commercial uranium nuclear fuel fabrication facilities

    International Nuclear Information System (INIS)

    At present in the United States, there are seven commercial light-water reactor uranium fuel fabrication facilities. Effluent wastes from these facilities include uranium, nitrogen, fluorine, and organic-containing compounds. These effluents may be either discharged to the ambient environment, treated and recycled internally, stored or disposed of on-site, sent off-site for treatment and/or recovery, or sent off-site for disposal (including disposal in low-level waste burial sites). Quantities of waste generated and treatment techniques vary greatly depending on the facility and circuits used internally at the facility, though in general all the fluorine entering the facility as UF6 is discharged as waste. Further studies to determine techniques and procedures that might minimize dose (ALARA) and to give data on possible long-term effects of effluent discharge and waste disposal are needed

  8. Internal exposure to uranium in a pooled cohort of gaseous diffusion plant workers

    International Nuclear Information System (INIS)

    Intakes and absorbed organ doses were estimated for 29 303 workers employed at three former US gaseous diffusion plants as part of a study of cause-specific mortality and cancer incidence in uranium enrichment workers. Uranium urinalysis data (>600 000 urine samples) were available for 58 % of the pooled cohort. Facility records provided uranium gravimetric and radioactivity concentration data and allowed estimation of enrichment levels of uranium to which workers may have been exposed. Urine data were generally recorded with facility department numbers, which were also available in study subjects' work histories. Bioassay data were imputed for study subjects with no recorded sample results (33 % of pooled cohort) by assigning department average urine uranium concentration. Gravimetric data were converted to 24-h uranium activity excretion using department average specific activities. Intakes and organ doses were calculated assuming chronic exposure by inhalation to a 5-μm activity median aerodynamic diameter aerosol of soluble uranium. Median intakes varied between 0.31 and 0.74 Bq d-1 for the three facilities. Median organ doses for the three facilities varied between 0.019 and 0.051, 0.68 and 1.8, 0.078 and 0.22, 0.28 and 0.74, and 0.094 and 0.25 mGy for lung, bone surface, red bone marrow, kidneys, and liver, respectively. Estimated intakes and organ doses for study subjects with imputed bioassay data were similar in magnitude. (authors)

  9. A model for recovery of scrap monolithic uranium molybdenum fuel by electrorefining

    Science.gov (United States)

    Van Kleeck, Melissa A.

    The goal of the Reduced Enrichment for Research and Test Reactors program (RERTR) is toreduce enrichment at research and test reactors, thereby decreasing proliferation risk at these facilities. A new fuel to accomplish this goal is being manufactured experimentally at the Y12 National Security Complex. This new fuel will require its own waste management procedure,namely for the recovery of scrap from its manufacture. The new fuel is a monolithic uraniummolybdenum alloy clad in zirconium. Feasibility tests were conducted in the Planar Electrode Electrorefiner using scrap U-8Mo fuel alloy. These tests proved that a uranium product could be recovered free of molybdenum from this scrap fuel by electrorefining. Tests were also conducted using U-10Mo Zr clad fuel, which confirmed that product could be recovered from a clad version of this scrap fuel at an engineering scale, though analytical results are pending for the behavior of Zr in the electrorefiner. A model was constructed for the simulation of electrorefining the scrap material produced in the manufacture of this fuel. The model was implemented on two platforms, Microsoft Excel and MatLab. Correlations, used in the model, were developed experimentally, describing area specific resistance behavior at each electrode. Experiments validating the model were conducted using scrap of U-10Mo Zr clad fuel in the Planar Electrode Electrorefiner. The results of model simulations on both platforms were compared to experimental results for the same fuel, salt and electrorefiner compositions and dimensions for two trials. In general, the model demonstrated behavior similar to experimental data but additional refinements are needed to improve its accuracy. These refinements consist of a function for surface area at anode and cathode based on charge passed. Several approximations were made in the model concerning areas of electrodes which should be replaced by a more accurate function describing these areas.

  10. Preliminary report on the promise of accelerator-driven natural-uranium-fueled light-water-moderated breeding power reactors

    International Nuclear Information System (INIS)

    A new concept for a power breeder reactor that consists of an accelerator-driven subcritical thermal fission system is proposed. In this system an accelerator provides a high-energy proton beam which interacts with a heavy-element target to produce, via spallation reactions, an intense source of neutrons. This source then drives a natural-uranium-fueled, light-water-moderated and -cooled subcritical blanket which both breeds new fuel and generates heat that can be converted to electrical power. The report given presents a general layout of the resulting Accelerator Driven Light Water Reactor (ADLWR), evaluates its performance, discusses its fuel cycle characteristics, and identifies the potential contributions to the nuclear energy economy this type of power reactor might make. A light-water thermal fission system is found to provide an attractive feature when designed to be source-driven. The equilibrium fissile fuel content that gives the highest energy multiplication is approximately equal to the content of 235U in natural uranium. Consequently, natural-uranium-fueled ADLWRs that are designed to have the highest energy generation per source neutron are also fuel-self-sufficient; that is, their fissile fuel content remains constant with burnup. This feature allows the development of a nuclear energy system that is based on the most highly developed fission technology available (the light water reactor technology) and yet has a simple and safe fuel cycle. ADLWRs will breed on natural uranium, have no doubling time limitation, and be free from the need for uranium enrichment or for the separation of plutonium. It appears that ADLWRs could also be efficiently operated with thorium fuel cycles and with denatured fuel cycles

  11. 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.

  12. Low-Enriched Uranium Fuel Design with Two-Dimensional Grading for the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [ORNL; Primm, Trent [ORNL

    2011-05-01

    An engineering design study of the conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel is ongoing at Oak Ridge National Laboratory. The computational models developed during fiscal year 2010 to search for an LEU fuel design that would meet the requirements for the conversion and the results obtained with these models are documented and discussed in this report. Estimates of relevant reactor performance parameters for the LEU fuel core are presented and compared with the corresponding data for the currently operating HEU fuel core. The results obtained indicate that the LEU fuel design would maintain the current performance of the HFIR with respect to the neutron flux to the central target region, reflector, and beam tube locations under the assumption that the operating power for the reactor fueled with LEU can be increased from the current value of 85 MW to 100 MW.

  13. The prospect of uranium nitride (UN) and mixed nitride fuel (UN-PuN) for pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Syarifah, Ratna Dewi, E-mail: syarifah.physics@gmail.com; Suud, Zaki, E-mail: szaki@fi.itb.ac.id [Nuclear Physics and Biophysics Research Division, Physics Department, Faculty of Mathematics and Natural Science, Bandung Institute of Technology Jalan Ganesha 10, Bandung 40132 (Indonesia)

    2015-09-30

    Design study of small Pressurized Water Reactors (PWRs) core loaded with uranium nitride fuel (UN) and mixed nitride fuel (UN-PuN), Pa-231 as burnable poison, and Americium has been performed. Pa-231 known as actinide material, have large capture cross section and can be converted into fissile material that can be utilized to reduce excess reactivity. Americium is one of minor actinides with long half life. The objective of adding americium is to decrease nuclear spent fuel in the world. The neutronic analysis results show that mixed nitride fuel have k-inf greater than uranium nitride fuel. It is caused by the addition of Pu-239 in mixed nitride fuel. In fuel fraction analysis, for uranium nitride fuel, the optimum volume fractions are 45% fuel fraction, 10% cladding and 45% moderator. In case of UN-PuN fuel, the optimum volume fractions are 30% fuel fraction, 10% cladding and 60% coolant/ moderator. The addition of Pa-231 as burnable poison for UN fuel, enrichment U-235 5%, with Pa-231 1.6% has k-inf more than one and excess reactivity of 14.45%. And for mixed nitride fuel, the lowest value of reactivity swing is when enrichment (U-235+Pu) 8% with Pa-231 0.4%, the excess reactivity value 13,76%. The fuel pin analyze for the addition of Americium, the excess reactivity value is lower than before, because Americium absorb the neutron. For UN fuel, enrichment U-235 8%, Pa-231 1.6% and Am 0.5%, the excess reactivity is 4.86%. And for mixed nitride fuel, when enrichment (U-235+Pu) 13%, Pa-231 0.4% and Am 0.1%, the excess reactivity is 11.94%. For core configuration, it is better to use heterogeneous than homogeneous core configuration, because the radial power distribution is better.

  14. Correlation of radioactive waste treatment costs and the environmental impact of waste effluents in the nuclear fuel cycle: fabrication of high-temperature gas-cooled reactor fuel containing uranium-233 and thorium

    International Nuclear Information System (INIS)

    A cost/benefit study was made to determine the cost and effectiveness of various radioactive waste (radwaste) treatment systems for decreasing the release of radioactive materials from model High-Temperature Gas-Cooled (HTGR) fuel fabrication plants and to determine the radiological impact (dose commitment) of the released materials on the environment. The study is designed to assist in defining the term ''as low as reasonably achievable'' as it applies to these nuclear facilities. The base cases of the two model plants, a fresh fuel fabrication plant and a refabrication plant, are representative of current proposed commercial designs or are based on technology that is being developed to fabricate uranium, thorium, and graphite into fuel elements. The annual capacities of the fresh fuel plant and the refabrication plant are 450 and 245 metric tons of heavy metal (where heavy metal is uranium plus thorium), as charged to about fifty 1000-MW(e) HTGRs. Additional radwaste treatment systems are added to the base case plants in a series of case studies to decrease the amounts of radioactive materials released and to reduce the radiological dose commitment to the population in the surrounding area. The capital and annual costs for the added waste treatment operations and the corresponding reductions in dose commitments are calculated for each case. In the final analysis, the cost/benefit of each case, calculated as additional cost of radwaste system divided by the reduction in dose commitment, is tabulated or the dose commitment is plotted with cost as the variable. The status of each of the radwaste treatment methods is discussed. 48 figures, 74 tables

  15. Correlation of radioactive waste treatment costs and the environmental impact of waste effluents in the nuclear fuel cycle: fabrication of high-temperature gas-cooled reactor fuel containing uranium-233 and thorium

    Energy Technology Data Exchange (ETDEWEB)

    Roddy, J.W.; Blanco, R.E.; Hill, G.S.; Moore, R.E.; Seagren, R.D.; Witherspoon, J.P.

    1976-06-01

    A cost/benefit study was made to determine the cost and effectiveness of various radioactive waste (radwaste) treatment systems for decreasing the release of radioactive materials from model High-Temperature Gas-Cooled (HTGR) fuel fabrication plants and to determine the radiological impact (dose commitment) of the released materials on the environment. The study is designed to assist in defining the term ''as low as reasonably achievable'' as it applies to these nuclear facilities. The base cases of the two model plants, a fresh fuel fabrication plant and a refabrication plant, are representative of current proposed commercial designs or are based on technology that is being developed to fabricate uranium, thorium, and graphite into fuel elements. The annual capacities of the fresh fuel plant and the refabrication plant are 450 and 245 metric tons of heavy metal (where heavy metal is uranium plus thorium), as charged to about fifty 1000-MW(e) HTGRs. Additional radwaste treatment systems are added to the base case plants in a series of case studies to decrease the amounts of radioactive materials released and to reduce the radiological dose commitment to the population in the surrounding area. The capital and annual costs for the added waste treatment operations and the corresponding reductions in dose commitments are calculated for each case. In the final analysis, the cost/benefit of each case, calculated as additional cost of radwaste system divided by the reduction in dose commitment, is tabulated or the dose commitment is plotted with cost as the variable. The status of each of the radwaste treatment methods is discussed. 48 figures, 74 tables.

  16. Review of consequences of uranium hydride formation in N-Reactor fuel elements stored in the K-Basins

    Energy Technology Data Exchange (ETDEWEB)

    Weber, J.W.

    1994-09-28

    The 105-K Basins on the Hanford site are used to store uranium fuel elements and assemblies irradiated in and discharged from N Reactor. The storage cylinders in KW Basin are known to have some broken N reactor fuel elements in which the exposed uranium is slowly reacting chemically with water in the cylinder. The products of these reactions are uranium oxide, hydrogen, and potentially some uranium hydride. The purpose of this report is to document the results f the latest review of potential, but highly unlikely accidents postulated to occur as closed cylinders containing N reactor fuel assemblies are opened under water in the KW basin and as a fuel assembly is raised from the basin in a shipping cask for transportation to the 327 Building for examination as part of the SNF Characterization Program. The postulated accidents reviews in this report are considered to bound all potential releases of radioactivity and hydrogen. These postulated accidents are: (1) opening and refill of a cylinder containing significant amounts of hydrogen and uranium hydride; and (2) draining of the single element can be used to keep the fuel element submerged in water after the cask containing the can and element is lifted from the KW Basin. Analysis shows the release of radioactivity to the site boundary is significantly less than that allowed by the K Basin Safety Evaluation. Analysis further shows there would be no damage to the K Basin structure nor would there be injury to personnel for credible events.

  17. Nuclear energy in Europe: uranium flow modeling and fuel cycle scenario trade-offs from a sustainability perspective.

    Science.gov (United States)

    Tendall, Danielle M; Binder, Claudia R

    2011-03-15

    The European nuclear fuel cycle (covering the EU-27, Switzerland and Ukraine) was modeled using material flow analysis (MFA).The analysis was based on publicly available data from nuclear energy agencies and industries, national trade offices, and nongovernmental organizations. Military uranium was not considered due to lack of accessible data. Nuclear fuel cycle scenarios varying spent fuel reprocessing, depleted uranium re-enrichment, enrichment assays, and use of fast neutron reactors, were established. They were then assessed according to environmental, economic and social criteria such as resource depletion, waste production, chemical and radiation emissions, costs, and proliferation risks. The most preferable scenario in the short term is a combination of reduced tails assay and enrichment grade, allowing a 17.9% reduction of uranium demand without significantly increasing environmental, economic, or social risks. In the long term, fast reactors could theoretically achieve a 99.4% decrease in uranium demand and nuclear waste production. However, this involves important costs and proliferation risks. Increasing material efficiency is not systematically correlated with the reduction of other risks. This suggests that an overall optimization of the nuclear fuel cycle is difficult to obtain. Therefore, criteria must be weighted according to stakeholder interests in order to determine the most sustainable solution. This paper models the flows of uranium and associated materials in Europe, and provides a decision support tool for identifying the trade-offs of the alternative nuclear fuel cycles considered.

  18. 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)

  19. 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.

  20. A plant taxonomic survey of the Uranium City region, Lake Athabasca north shore, emphasizing the naturally colonizing plants on uranium mine and mill wastes and other human-disturbed sites

    International Nuclear Information System (INIS)

    A goal of this study was to acquire more complete baseline data on the existing flora of the Uranium City region, both in natural and human-disturbed sites. Emphasis was given to determining which plant species were naturally revegetating various abandoned uranium mine and mill waste disposal areas, other human-disturbed sites, and ecologically analogous sites. Another goal was to document the occurrence and distribution in the study region of rare and possibly endangered species. A further objective was to suggest regionally-occurring species with potential value for revegetating uranium mine and mill waste sites. Field investigations were carried out in the Uranium City region during August, 1981. During this time 1412 plant collections were made; a total of 366 plant species - trees, shrubs, forbs, graminoids, lichens, and bryophytes were recorded. The report includes an annotated checklist of plant species of the Uranium City region and a reference index of plant taxa indicating species that have high revegetation potential

  1. The uranium market and its characteristics

    International Nuclear Information System (INIS)

    The subject is covered in sections, entitled as shown. Numerical data are indicated in parenthesis. General characteristics of the uranium market, (enrichment plant variables, fuel requirements of a 1000 MWe power plant); demand pattern (enrichment cost relationships), supply pattern; uranium price analysis, production cost (relationship between future uranium requirements and discovery rates necessary), market break-even cost (break-even uranium cost as a function of fossil fuel prices), market value (theoretical and actual supply - demand balance in uranium market, relationship between U3O8 price and world production); geographic and economic distribution of producers and consumers (world resources of uranium, relationship between U308 world production capacity and annual requirements in 1990). (U.K.)

  2. The compatibility of stainless steels with particles and powders of uranium carbide and low-sulphur UCS fuels

    International Nuclear Information System (INIS)

    Slightly hyperstoichiometric (U,Pu)C is a potential nuclear fuel for fast breeder reactors. The excess carbon above the stoichiometric amount results in a higher carbon activity in the fuel, and carbon is transferred to the stainless steel cladding, resulting in embrittlement of the cladding. It is with this problem of carbon transfer from the fuel to the cladding that this thesis is concerned. For practical reasons, UC and not (U,Pu)C was used as the fuel. The theory of decarburisation of carbide fuel and the carburisation of stainless steel, the facilities constructed for the project at the Atomic Energy Board, and the experimental techniques used, including preparation of the fuels, are discussed. The effect of a number of variables of uranium carbide fuel on its compatibility behaviour with stainless steels was investigated, as well as the effect om microstructure and type of stainless steel (304, 304 L and 316) on the rate of carburisation. These studies can be briefly summarised under the following headings: powder-particle size; surface oxidation of uranium carbide; preparation temperature of uranium carbide; low sulfur UCS fuels; uranium sulfide and the microstructure and type of steel. The author concludes that: the effect of surface oxidation and particle size must be taken into account when evaluating out-of-pile tests; the possible effects of surface oxidation must be taken into account when considering vibro-compacted carbide fuels; there is no advantage in replacing a fraction of the carbon atoms by sulphur atoms in slightly hyperstoichiometric carbide fuels, and the type and thermo-mechanical treatment of the stainless steel used as cladding material in a fuel pin is not important as far as the rate of carburisation by the fuel is concerned

  3. Root uptake of uranium by a higher plant model (Phaseolus vulgaris) bioavailability from soil solution

    Energy Technology Data Exchange (ETDEWEB)

    Laroche, L.; Henner, P.; Camilleri, V.; Garnier-Laplace, J. [CEA Cadarache (DEI/SECRE/LRE), Laboratory of Radioecology and Ecotoxicology, Institute for Radioprotection and Nuclear Safety, 13 - Saint-Paul-lez-Durance (France)

    2004-07-01

    Uranium behaviour in soils is controlled by actions and interactions between physicochemical and biological processes that also determine its bioavailability. In soil solution, uranium(+VI) aqueous speciation undergoes tremendous changes mainly depending on pH, carbonates, phosphates and organic matter. In a first approach to identify bioavailable species of U to plants, cultures were performed using hydroponics, to allow an easy control of the composition of the exposure media. The latter, here an artificial soil solution, was designed to control the uranium species in solution. The geochemical speciation code JCHESS using a database compiled from the OECD/NEA thermochemical database project and verified was used to perform the solution speciation calculations. On this theoretical basis, three domains were defined for short-duration well-defined laboratory experiments in simplified conditions: pH 4.9, 5.8 and 7 where predicted dominant species are uranyl ions, hydroxyl complexes and carbonates respectively. For these domains, biokinetics and characterization of transmembrane transport according to a classical Michaelis Menten approach were investigated. The Free Ion Model (or its derived Biotic Ligand Model) was tested to determine if U uptake is governed by the free uranyl species or if other metal complexes can be assimilated. The effect of different variables on root assimilation efficiency and phyto-toxicity was explored: presence of ligands such as phosphates or carbonates and competitive ions such as Ca{sup 2+} at the 3 pH. According to previous experiments, uranium was principally located in roots whatever the pH and no difference in uranium uptake was evidenced between the main growth stages of the plant. Within the 3 studied chemical domains, results from short-term kinetics evidenced a linear correlation between total uranium concentration in bean roots and that in exposure media, suggesting that total uranium in soil solution could be a good predictor

  4. Sustainability of the Chinese nuclear expansion: Natural uranium resources availability, Pu cycle, fuel utilization efficiency and spent fuel management

    International Nuclear Information System (INIS)

    Highlights: • We simulated 4 different future nuclear fuel cycle scenarios. • The ADS and FRs impact on the future Chinese nuclear fuel cycle is studied. • The partition and transmutation option is compared against the simple reprocessing. • The U requirement, Pu flow and MA cycle are key aspects for decision makers. - Abstract: The civil nuclear energy deployment in China is important for future “Nuclear Renaissance” of China and worldwide. Compared to the other nations that developed their nuclear power energy system in last century, China can take advantage of the research and mistakes made by those states in relation to the back-end of the nuclear fuel cycle (NFC). The spent fuel accumulated by decades of operations of civil nuclear power is today a big burden for the industry. China must carefully plan the NFC for a sustainable development of the nuclear energy with special consideration to close the fuel cycle. The present paper addresses the NFC options and implications of a LWR scenario development and of a fast reactor park developed after 2035 and 2050, and covers the historical development of nuclear energy in China (i.e. from the first criticality of the first reactor) to the year 2100. The paper studies the partition and transmutation strategy with the use of accelerator driven system (ADS) to burn the minor actinides (MA) to understand the ADS impact on the NFC and to estimate the number and the necessary deploying schedule of the ADS reactors to limit the minor actinides stock build up. The other aspects taken into consideration for the comparison of the different scenarios are the natural uranium resourced used, the efficiency of fuel utilization, the proliferation and diversion risks associated to each scenario and the overall spent fuel production and flow. The code INFCIS developed by the International Atomic Energy Agency (IAEA) is used in the present study

  5. AIR SHIPMENT OF HIGHLY ENRICHED URANIUM SPENT NUCLEAR FUEL FROM ROMANIA AND LIBYA

    Energy Technology Data Exchange (ETDEWEB)

    Christopher Landers; Igor Bolshinsky; Ken Allen; Stanley Moses

    2010-07-01

    In June 2009 Romania successfully completed the world’s first air shipment of highly enriched uranium (HEU) spent nuclear fuel transported in Type B(U) casks under existing international laws and without special exceptions for the air transport licenses. Special 20-foot ISO shipping containers and cask tiedown supports were designed to transport Russian TUK 19 shipping casks for the Romanian air shipment and the equipment was certified for all modes of transport, including road, rail, water, and air. In December 2009 Libya successfully used this same equipment for a second air shipment of HEU spent nuclear fuel. Both spent fuel shipments were transported by truck from the originating nuclear facilities to nearby commercial airports, were flown by commercial cargo aircraft to a commercial airport in Yekaterinburg, Russia, and then transported by truck to their final destinations at the Production Association Mayak facility in Chelyabinsk, Russia. Both air shipments were performed under the Russian Research Reactor Fuel Return Program (RRRFR) as part of the U.S. National Nuclear Security Administration (NNSA) Global Threat Reduction Initiative (GTRI). The Romania air shipment of 23.7 kg of HEU spent fuel from the VVR S research reactor was the last of three HEU fresh and spent fuel shipments under RRRFR that resulted in Romania becoming the 3rd RRRFR participating country to remove all HEU. Libya had previously completed two RRRFR shipments of HEU fresh fuel so the 5.2 kg of HEU spent fuel air shipped from the IRT 1 research reactor in December made Libya the 4th RRRFR participating country to remove all HEU. This paper describes the equipment, preparations, and license approvals required to safely and securely complete these two air shipments of spent nuclear fuel.

  6. Assessment of failed fuel and tramp uranium based on the activity of fission products in the primary circuit

    International Nuclear Information System (INIS)

    We have proposed a model for the nuclear fuel state of the operating power reactor from the physical characteristics of nuclear fission products which have been produced by nuclear reaction between neutron and uranium-235. The model equation for nuclear fission products release has been split into size independent steps: 1) calculation of the fission products generation inside the solid nuclear fuel, 2) release from the fuel to the fuel surface in three different ways, 3) release between the fuel surface and gap, 4) release from the defective nuclear fuel to the reactor coolant, 5) mass balance in the coolant taking into account the purification rate, 6) separation of fission products sources with two parts, i.e. fuel and tramp uranium. We have solved the equation of the model, calculated the activity of fission products released from the defected fuel to coolant and put the experimental activity data of the nuclear fission products in the primary coolant to determine the number of defective fuel and amount of tramp uranium by using the computer. The measurement and analysis of nuclear fission products in the primary coolant of nuclear power reactors have been carried out at the pressurized water reactor, Korea Nuclear Unit 2, 7 and 8. We have used the iodine isotopes among the nuclides of fission products. The analysis results have been well agreed with the results of diffusion model and of kinetics model. (author). 7 refs, 2 figs, 8 tabs

  7. 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.

  8. Linearity assumption in soil-to-plant transfer factors of natural uranium and radium in Helianthus annuus L

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez, P. Blanco [Departamento de Fisica, Facultad de Ciencias, Universidad de Extremadura, 06071 Badajoz (Spain); Tome, F. Vera [Departamento de Fisica, Facultad de Ciencias, Universidad de Extremadura, 06071 Badajoz (Spain)]. E-mail: fvt@unex.es; Fernandez, M. Perez [Area de Ecologia, Departamento de Fisica, Facultad de Ciencias, Universidad de Extremadura, 06071 Badajoz (Spain); Lozano, J.C. [Laboratorio de Radiactividad Ambiental, Facultad de Ciencias, Universidad de Salamanca, 37008 Salamanca (Spain)

    2006-05-15

    The linearity assumption of the validation of soil-to-plant transfer factors of natural uranium and {sup 226}Ra was tested using Helianthus annuus L. (sunflower) grown in a hydroponic medium. Transfer of natural uranium and {sup 226}Ra was tested in both the aerial fraction of plants and in the overall seedlings (roots and shoots). The results show that the linearity assumption can be considered valid in the hydroponic growth of sunflowers for the radionuclides studied. The ability of sunflowers to translocate uranium and {sup 226}Ra was also investigated, as well as the feasibility of using sunflower plants to remove uranium and radium from contaminated water, and by extension, their potential for phytoextraction. In this sense, the removal percentages obtained for natural uranium and {sup 226}Ra were 24% and 42%, respectively. Practically all the uranium is accumulated in the roots. However, 86% of the {sup 226}Ra activity concentration in roots was translocated to the aerial part.

  9. The determination of uranium distribution homogeneity in the fuel plates with the uranium loading of 4.80 and 5.20 g/cm3 by X-Ray attenuation

    International Nuclear Information System (INIS)

    The calibration of X-Ray intensity of the U3Si2-AI fuel plates with the uranium loading between 3.60 up to 5.20 g/cm3 and varied thickness of AIMgSi1 reference block have been performed. The measurement with changing variable slit diameter and energy of X-Ray attenuation, are produced enough representative X-Ray intensity at 18 mm slit diameter and energy of 43 kV. From the correlation of X-ray intensities vs variation of uranium loading in the fuel plates and thickness of the AIMgSi1 materials, the equivalence of thickness of the AIMgSi1 block to the uranium loading of fuel plates are determined. By assuming that the tolerance of the homogeneity measurement is + 20 % from normal thickness staircase of the AIMgSi1 standard could be determined and than together with fuel plate were scanned to determine the uranium homogeneity. The test result on the U3Si2-AI fuel plates with uranium loading of 4.80 and 5.20 g/cm3 (each 4 fuel plates) indicated that uranium distribution in the fuel plates is relatively homogeneous, with each maximum deviation being 6.30 % and 6.90%. It is showed that measurement method is relatively good, easy, and fast so that this method is suitable to control the uranium homogeneity in the fuel plate. (author)

  10. Characteristics and Disposal Categorization of Solid Radioactive Waste from the Front End of the Uranium Fuel Cycle

    International Nuclear Information System (INIS)

    The proper categorization of radioactive waste forms the basis for defining its disposal method. In particular, it is the basis for defining the disposal policy for solid radioactive waste from the front end of the uranium fuel cycle to identify scientifically its characteristics, in view of the differences in regulatory approach between artificial radioactive waste and NORM waste. The paper examines the disposal policy and practice in China and other countries for solid radioactive waste from the front end of the uranium fuel cycle and discusses the confusion in disposal of the waste as artificial radioactive waste. The radionuclide composition and characteristics of the solid radioactive waste from the front end of the uranium fuel cycle are investigated in detail and a new idea that such waste needs to be disposed of and categorized as NORM waste is proposed. (author)

  11. Thermophysical properties determination of uranium-molybdenum fuel by the flash laser method

    International Nuclear Information System (INIS)

    The conversion of nuclear facilities using high enriched uranium (HEU) to low enriched uranium (LEU) fuels and targets has been worldwide encouraged considering the nuclear non-proliferation treaty. A research program under development at CDTN aims the obtaining of uranium and molybdenum alloys to enable the HEU to LEU conversions. Samples of alloys produced with 3,5,7 and 10 weight percent of molybdenum were prepared and their thermophysical properties were determined at room temperature. The thermal diffusivity was measured by the laser flash method. This conceptually simple method presents advantages in comparison to the various methods used in the past. The thermal conductivity was determined by mathematical model developed by the laboratory. The obtained results show good agreement with the literature data with thermal conductivity values ranging from 7.15 up to 19.51 W ·m-1 ·K-1, the thermal diffusivity from 3.83 x 10-6 up to 6.37 x 10-6 m2·s-1 and the specific heat from 114 up to 184 J·kg-1·K-1. The GUM (Guide to the Expression Uncertainty in Measurement) uncertainty framework and the adaptive Monte Carlos Method (MCM) were used to obtain the associated standard uncertainty with an estimate of the output quantities. (author)

  12. SIMS Analyses of Aerodynamic Fallout from a Uranium-Fueled Test

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, L. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Univ. of California, Berkeley, CA (United States); Knight, K. B. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Matzel, J. E. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Prussin, S. G. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Ryerson, F. J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Kinman, W. S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Zimmer, M. M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hutcheon, I. D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2014-09-09

    Five silicate fallout glass spherules produced in a uranium-fueled, near-surface nuclear test were characterized by secondary ion mass spectrometry, electron probe microanalysis, autoradiography, scanning electron microscopy, and energy-dispersive x ray spectroscopy. Several samples display distinctive compositional heterogeneity suggestive of incomplete mixing, and exhibit heterogeneity in U isotopes with 0.02 < 235U/ 238U < 11.8 among all five samples and 0.02 < 235U/ 238U < 7.81 within a single sample. In two samples, the 235U/ 238U ratio is correlated with major element composition, consistent with the agglomeration of chemically and isotopically distinct molten precursors. Two samples are quasi-homogeneous with respect to composition and uranium isotopic composition, suggesting extensive mixing possibly due longer residence time in the fireball. Correlated variations between 234U, 235U, 236U and 238U abundances point to mixing of end-members corresponding to uranium derived from the device and natural U ( 238U/ 235U = 0.00725) found in soil.

  13. 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.

  14. Post irradiation examinations of 84F-10A capsule containing uranium-plutonium mixed carbide fuels

    International Nuclear Information System (INIS)

    Two fuel pins filled with uranium-plutonium mixed carbide pellets having different stoichiometry, (U,Pu)C1.0 and (U,Pu)C1.1, were encapsulated in 84F-10A and irradiated in JMTR up to 3.0%FIMA at a peak linear power of 59kW/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 sections of fuel pins that the helium gap between the pellets and the cladding tube was completely closed. At the central part of the fuel pellets the number of small pores was decreased and the grain growth was observed compared with the outer zone. (U,Pu)C1.1 pellets showed higher fission gas release ratio than (U,Pu)C1.0 pellets because the former had relatively high open porosity. Although slight carburization was observed near the inner surface of cladding tube the interaction did not affect the fuel performance itself. (author)

  15. The present status for storage of spent metal fuel for the storage design criteria of the reduced uranium metal

    International Nuclear Information System (INIS)

    To development an advanced spent nuclear fuel management technology for countries which have not decided a back-end nuclear fuel cycle policy yet, the development of advanced spent nuclear fuel management, which can reduce the storage volume by a quarter and the storage cooling load by half by the removal of highly radioactive decay-heat elements such as Cs and Sr, is conducting. Therefore, it is necessary that the research for the safe storage of the Reduced Uranium Metal is performed. It is important that the related regulations on storage are reviewed and the present status on storage of metal fuel is looked over. In order to establish the storage design criteria for the safe storage of the Reduced Uranium Metal, thus, it is necessary for not only the related regulations on storage but also the present status on storage of metal fuel to be analyzed.

  16. Optimization of thorium-uranium content in a 54-element fuel bundle for use in a CANDU-SCWR

    International Nuclear Information System (INIS)

    A new 54-element fuel bundle design has been proposed for use in a pressure-tube supercritical water-cooled reactor, a pre-conceptual evolution of existing CANDU reactors. Pursuant to the goals of the Generation IV International Forum regarding advancement in nuclear fuel cycles, optimization of the thorium and uranium content in each ring of fuel elements has been studied with the objectives of maximizing the achievable fuel utilization (burnup) and total thorium content within the bundle, while simultaneously minimizing the linear element ratings and coolant void reactivity. The bundle was modeled within a reactor lattice cell using WIMS-AECL, and the uranium and thorium content in each ring of fuel elements was optimized using a weighted merit function of the aforementioned criteria and a metaheuristic search algorithm. (author)

  17. Uranium and radium in water samples around the Nikola Tesla B lignite-fired power plant - Obrenovac, Serbia

    International Nuclear Information System (INIS)

    This paper deals with the analysis of natural radionuclide content in 23 water samples collected in the vicinity of the Nikola Tesla B thermal power plant, Serbia. All samples were analyzed for 226Ra and uranium isotopes (238U, 234U) activity using radiochemical methods and alpha spectrometry. Obtained results show that the activity concentrations for uranium and radium in the water around the thermal power plant are low when compared to those from areas across Serbia with their enhanced natural uranium and radium content. No important radiological hazard related to uranium and radium activity stored in heap was found. (author)

  18. Uranium and radium in water samples around the Nikola Tesla B lignite-fired power plant - Obrenovac, Serbia

    Directory of Open Access Journals (Sweden)

    Žunić Zora S.

    2011-01-01

    Full Text Available This paper deals with the analysis of natural radionuclide content in 23 water samples collected in the vicinity of the Nikola Tesla B thermal power plant, Serbia. All samples were analyzed for 226Ra and uranium isotopes (238U, 234U activity using radiochemical methods and alpha spectrometry. Obtained results show that the activity concentrations for uranium and radium in the water around the thermal power plant are low when compared to those from areas across Serbia with their enhanced natural uranium and radium content. No important radiological hazard related to uranium and radium activity stored in heap was found.

  19. Calculation of parameters for inspection planning and evaluation: low enriched uranium conversion and fuel fabrication facilities

    International Nuclear Information System (INIS)

    As part of Task C.35 (Calculation of Parameters for Inspection Planning and Evaluation) of the US Program of Technical Assistance to IAEA Safeguards, Pacific Northwest Laboratory has performed some quantitative analyses of IAEA inspection activities at low-enriched uranium (LEU) conversion and fuel fabrication facilities. This report presents the results and conclusions of those analyses. Implementation of IAEA safeguards at LEU conversion and fuel fabrication facilities must take into account a variety of practical problems and constraints. One of the key concerns is the problem of flow verification, especially product verification. The objective of this report is to help put the problem of flow verification in perspective by presenting the results of some specific calculations of inspection effort and probability of detection for various product measurement strategies. In order to provide quantitative information about the advantages and disadvantages of the various strategies, eight specific cases were examined

  20. A two-dimensional, finite-difference model of the oxidation of a uranium carbide fuel pellet

    Science.gov (United States)

    Shepherd, James; Fairweather, Michael; Hanson, Bruce C.; Heggs, Peter J.

    2015-12-01

    The oxidation of spent uranium carbide fuel, a candidate fuel for Generation IV nuclear reactors, is an important process in its potential reprocessing cycle. However, the oxidation of uranium carbide in air is highly exothermic. A model has therefore been developed to predict the temperature rise, as well as other useful information such as reaction completion times, under different reaction conditions in order to help in deriving safe oxidation conditions. Finite difference-methods are used to model the heat and mass transfer processes occurring during the reaction in two dimensions and are coupled to kinetics found in the literature.

  1. Results of the work on development of research reactor fuel element based on high density fuel with decreased enrichment in uranium-235

    International Nuclear Information System (INIS)

    The work is developed to the branch program on decrease of fuel enrichment in the Russian research reactors. The analysis of results of foreign works and own studies on creating dispersion fuel compositions on the basis of high density uranium compounds is accomplished. The uranium alloys U3Si, U3Si alloys, U-Zr-Nb, U6Fe and U6Fe alloys are chosen for application in further developments. Characteristics of these alloys are given from the viewpoint of their behaviour under irradiation

  2. Comparison of potential radiological consequences from a spent-fuel repository and natural uranium deposits

    International Nuclear Information System (INIS)

    A general criterion has been suggested for deep geological repositories containing spent fuel - the repositories should impose no greater radiological risk than due to naturally occurring uranium deposits. The following analysis investigates the rationale of that suggestion and determines whether current expectations of spent-fuel repository performance are consistent with such a criterion. In this study, reference spent-fuel repositories were compared to natural uranium-ore deposits. Comparisons were based on intrinsic characteristics, such as radionuclide inventory, depth, proximity to aquifers, and regional distribution, and actual and potential radiological consequences that are now occurring from some ore deposits and that may eventually occur from repositories and other ore deposits. The comparison results show that the repositories are quite comparable to the natural ore deposits and, in some cases, present less radiological hazard than their natural counterparts. On the basis of the first comparison, placing spent fuel in a deep geologic repository apparently reduces the hazard from natural radioactive materials occurring in the earth's crust by locating the waste in impermeable strata without access to oxidizing conditions. On the basis of the second comparison, a repository constructed within reasonable constraints presents no greater hazard than a large ore deposit. It is recommended that if the naturally radioactive environment is to be used as a basis for a criterion regarding repositories, then this criterion should be carefully constructed. The criterion should be based on the radiological quality of the waters in the immediate region of a specific repository, and it should be in terms of an acceptable potential increase in the radiological content of those waters due to the existence of the repository

  3. Comparison of potential radiological consequences from a spent-fuel repository and natural uranium deposits

    Energy Technology Data Exchange (ETDEWEB)

    Wick, O.J.; Cloninger, M.O.

    1980-09-01

    A general criterion has been suggested for deep geological repositories containing spent fuel - the repositories should impose no greater radiological risk than due to naturally occurring uranium deposits. The following analysis investigates the rationale of that suggestion and determines whether current expectations of spent-fuel repository performance are consistent with such a criterion. In this study, reference spent-fuel repositories were compared to natural uranium-ore deposits. Comparisons were based on intrinsic characteristics, such as radionuclide inventory, depth, proximity to aquifers, and regional distribution, and actual and potential radiological consequences that are now occurring from some ore deposits and that may eventually occur from repositories and other ore deposits. The comparison results show that the repositories are quite comparable to the natural ore deposits and, in some cases, present less radiological hazard than their natural counterparts. On the basis of the first comparison, placing spent fuel in a deep geologic repository apparently reduces the hazard from natural radioactive materials occurring in the earth's crust by locating the waste in impermeable strata without access to oxidizing conditions. On the basis of the second comparison, a repository constructed within reasonable constraints presents no greater hazard than a large ore deposit. It is recommended that if the naturally radioactive environment is to be used as a basis for a criterion regarding repositories, then this criterion should be carefully constructed. The criterion should be based on the radiological quality of the waters in the immediate region of a specific repository, and it should be in terms of an acceptable potential increase in the radiological content of those waters due to the existence of the repository.

  4. Standard model for safety analysis report of hexafluoride power plants from natural uranium

    International Nuclear Information System (INIS)

    The standard model for safety analysis report for hexafluoride production power plants from natural uranium is presented, showing the presentation form, the nature and the degree of detail, of the minimal information required by the Brazilian Nuclear Energy Commission - CNEN. (E.G.)

  5. The logistics and the supply chain in the Juzbado Nuclear Fuel Manufacturing Plant; Cadena logistica en la fabrica de elementos combustibles de Juzbado

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-07-01

    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)

  6. Surface Decontamination of System Components in Uranium Conversion Plant at KAERI

    International Nuclear Information System (INIS)

    A chemical decontamination process using nitric acid solution was selected as in-situ technology for recycle or release with authorization of a large amount of metallic waste including process system components such as tanks, piping, etc., which is generated by dismantling a retired uranium conversion plant at Korea Atomic Energy Research Institute (KAERI). The applicability of nitric acid solution for surface decontamination of metallic wastes contaminated with uranium compounds was evaluated through the basic research on the dissolution of UO2 and ammonium uranyl carbonate (AUC) powder. Decontamination performance was verified by using the specimens contaminated with such uranium compounds as UO2 and AUC taken from the uranium conversion plant. Dissolution rate of UO2 powder is notably enhanced by the addition of H2O2 as an oxidant even in the condition of a low concentration of nitric acid and low temperature compared with those in a nitric acid solution without H2O2. AUC powders dissolve easily in nitric acid solutions until the solution pH attains about 2.5 ∼ 3. Above that solution pH, however, the uranium concentration in the solution is lowered drastically by precipitation as a form of U3(NH3)4O9 . 5H2O. Decontamination performance tests for the specimens contaminated with UO2 and AUC were quite successful with the application of decontamination conditions obtained through the basic studies on the dissolution of UO2 and AUC powders

  7. Microstructural evolution of a uranium-10 wt.% molybdenum alloy for nuclear reactor fuels

    Science.gov (United States)

    Clarke, A. J.; Clarke, K. D.; McCabe, R. J.; Necker, C. T.; Papin, P. A.; Field, R. D.; Kelly, A. M.; Tucker, T. J.; Forsyth, R. T.; Dickerson, P. O.; Foley, J. C.; Swenson, H.; Aikin, R. M.; Dombrowski, D. E.

    2015-10-01

    Low-enriched uranium-10 wt.% molybdenum (LEU-10wt.%Mo) is of interest for the fabrication of monolithic fuels to replace highly-enriched uranium (HEU) dispersion fuels in high performance research and test reactors around the world. In this work, depleted uranium-10 wt.%Mo (DU-10wt.%Mo) is used to simulate the solidification and microstructural evolution of LEU-10wt.%Mo. Electron backscatter diffraction (EBSD) and complementary electron probe microanalysis (EPMA) reveal significant microsegregation present in the metastable γ-phase after solidification. Homogenization is performed at 800 and 1000 °C for times ranging from 1 to 32 h to explore the time-temperature combinations that will reduce the extent of microsegregation, as regions of higher and lower Mo content may influence local mechanical properties and provide preferred regions for γ-phase decomposition. We show for the first time that EBSD can be used to qualitatively assess microstructural evolution in DU-10wt.%Mo after homogenization treatments. Complementary EPMA is used to quantitatively confirm this finding. Homogenization at 1000 °C for 2-4 h may the regions that contain 8 wt.% Mo or lower, whereas homogenization at 1000 °C for longer than 8 h effectively saturates Mo chemical homogeneity, but results in substantial grain growth. The appropriate homogenization time will depend upon additional microstructural considerations, such as grain growth and intended subsequent processing. Higher carbon LEU-10wt.%Mo generally contains more inclusions within the grains and at grain boundaries after solidification. The effect of these inclusions on microstructural evolution (e.g. grain growth) during homogenization and as potential γ-phase decomposition nucleation sites is unclear, but likely requires additional study.

  8. Rapid, cost effective and non destructive determination of concentration of uranium in the process stream samples of uranium oxide plant by gamma spectrometry

    International Nuclear Information System (INIS)

    A rapid, cost effective and nondestructive gamma spectrometric method for the determination of concentration of uranium in the process stream samples of uranium oxide plant has been developed. A well type 3 x 3 NaI(Tl) detector coupled with a multichannel analyser has provided the advantages of greater accuracy and precision in results. Intensity of 185.7 keV peak of 235U is correlated with the concentration of uranium. Results compared well with those obtained by wavelength dispersive x-ray fluorescence spectrometric method and potentiometric method and a relative standard deviation of 2 % is attainable. (author)

  9. Comparative analysis of thorium and uranium fuel for transuranic recycle in a sodium cooled Fast Reactor

    International Nuclear Information System (INIS)

    Highlights: • Thorium as support fertile material for TRU transmutation in Fast Reactors. • Comparative analysis of Th and U based breakeven and burner Fast Reactors. • Thorium fosters significant advantages in terms of safety parameters. • Inherent safety is investigated through quasi-static reactivity and energy balances. • Th use in low-CR Fast Reactors does not reduce fuel decay heat and neutron sources. - Abstract: The present paper compares the reactor physics and transmutation performance of sodium-cooled Fast Reactors (FRs) for TRansUranic (TRU) burning with thorium (Th) or uranium (U) as fertile materials. The 1000 MWt Toshiba-Westinghouse Advanced Recycling Reactor (ARR) conceptual core has been used as benchmark for the comparison. Both burner and breakeven configurations sustained or started with a TRU supply, and assuming full actinide homogeneous recycle strategy, have been developed. State-of-the-art core physics tools have been employed to establish fuel inventory and reactor physics performances for equilibrium and transition cycles. Results show that Th fosters large improvements in the reactivity coefficients associated with coolant expansion and voiding, which enhances safety margins and, for a burner design, can be traded for maximizing the TRU burning rate. A trade-off of Th compared to U is the significantly larger fuel inventory required to achieve a breakeven design, which entails additional blankets at the detriment of core compactness as well as fuel manufacturing and separation requirements. The gamma field generated by the progeny of U-232 in the U bred from Th challenges fuel handling and manufacturing, but in case of full recycle, the high contents of Am and Cm in the transmutation fuel impose remote fuel operations regardless of the presence of U-232

  10. 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

  11. Explosion risks linked to red oils in the spent fuels reprocessing plants

    International Nuclear Information System (INIS)

    This paper presents the risk of explosion associated with reactions between tributyl phosphate (TBP) and its degradation products and nitrates from nitric acid or associated with heavy metals (uranium and plutonium); These reactions may lead to the formation of unstable compounds known as 'red oils'. The feedback explosions linked to the formation of such compounds occurring in spent fuel reprocessing plants round the world, is briefly discussed. The main measures to control these risks, implemented in French factories concerned are also presented. (N.C.)

  12. Establishing a Cost Basis for Converting the High Flux Isotope Reactor from High Enriched to Low Enriched Uranium Fuel

    International Nuclear Information System (INIS)

    Under the auspices of the Global Threat Reduction Initiative Reduced Enrichment for Research and Test Reactors Program, the National Nuclear Security Administration/Department of Energy (NNSA/DOE) has, as a goal, to convert research reactors worldwide from weapons grade to non-weapons grade uranium. The High Flux Isotope Reactor (HFIR) at Oak Ridge National Lab (ORNL) is one of the candidates for conversion of fuel from high enriched uranium (HEU) to low enriched uranium (LEU). A well documented business model, including tasks, costs, and schedules was developed to plan the conversion of HFIR. Using Microsoft Project, a detailed outline of the conversion program was established and consists of LEU fuel design activities, a fresh fuel shipping cask, improvements to the HFIR reactor building, and spent fuel operations. Current-value costs total $76 million dollars, include over 100 subtasks, and will take over 10 years to complete. The model and schedule follows the path of the fuel from receipt from fuel fabricator to delivery to spent fuel storage and illustrates the duration, start, and completion dates of each subtask to be completed. Assumptions that form the basis of the cost estimate have significant impact on cost and schedule.

  13. AREVA invests 610 million euro in new uranium conversion plants

    International Nuclear Information System (INIS)

    AREVA today announced the launch of the Comurhex II project which will see the group build new uranium conversion facilities on the Malvesi site in Narbonne and Tricastin. Through this 610 million euro investment, AREVA aims to maintain its position as world no. 1 for conversion within a context of global nuclear energy. COMURHEX II integrates technological innovations from major R and D programs and return of experience from processes in operation for over forty years. Nuclear safety and reducing the impact on the environment were top priorities when designing the project. These future facilities will also lead to major savings of water and energy consumption and reduce effluents. The groundwork of the Comurhex II project has taken 150,000 hours of engineering over the past three years. Four hundred people will work on the site which will be launched in summer 2007. First industrial production is scheduled for 2012, based on 15,000 metric tons of uranium per year. This figure may be increased to 21,000 tons to meet market requirements

  14. The evaluation, design and construction of the uranium plant for Chemwes Limited

    International Nuclear Information System (INIS)

    The Chemwes uranium plant was designed and constructed within fifteen months; commissioning started during June 1979, and the plant was producing at design capacity four months later. This account highlights the procedures and methods adopted to bring the plant into production. The description of the various phases includes some details of the early evaluation, and the feasibility and optimization studies; of the financing, project management, design, construction, and commissioning of the plant; and of the training of personnel. Some of the mistakes made and the factors that contributed to the success are also listed

  15. An Approach to Reduce Load on the Acid Leaching Circuit of the Commercial Uranium Recovery Plant at Jaduguda, India

    OpenAIRE

    Rao, G. V.; Prakash, S.

    1998-01-01

    The commercial uranium recovery plant at Jaduguda, Bihar, India, currently treats around 900 tonnes of ore per day from the Jaduguda mine, containing around 0.05% U3O8. Subsequent to removal of the sulphide minerals present in the ore by flotation, nearly 95% of the tailings are being treated in the acid leaching circuit to recover the uranium values. Laboratory investigations on the bulk flotation tailings revealed that around 63% of the uranium values are associated with feebly magnetic mat...

  16. Uranium deposit removal from the Oak Ridge Gaseous Diffusion Plant K-25 Building

    International Nuclear Information System (INIS)

    The Oak Ridge Gaseous Diffusion Plant went into operation as the first plant to separate uranium by the gaseous diffusion process. It was built during World War II as part of the U.S. Army Corps of Engineers' Manhattan Project. Its war-time code name was K-25, which was also the name of the first uranium separation building constructed at the installation. The K-25 building was considered an engineering miracle at the time of its construction. Built in a U shape ∼1 mile long and 400 ft wide, it housed complex and unique separation equipment. Despite its size and complexity, it was made fully operational within <2 yr after construction began. The facility operated successfully for more than 20 yr until it was placed in a standby mode in 1964. It is now clear the K-25 gaseous diffusion plant will never again be used to enrich uranium. The U.S. Department of Energy, therefore, has initiated a decontamination and decommission program. This paper discusses various procedures and techniques for addressing critical mass, uranium deposits, and safeguards issues

  17. Extraction of uranium from seawater: chemical process and plant design feasibility study

    International Nuclear Information System (INIS)

    A major assessment was made of the uranium resources in seawater. Several concepts for moving seawater to recover the uranium were investigated, including pumping the seawater and using natural ocean currents or tides directly. The optimal site chosen was on the southeastern Puerto Rico coast, with the south U.S. Atlantic coast as an alternate. The various processes for extracting uranium from seawater were reviewed, with the adsorption process being the most promising at the present time. Of the possible adsorbents, hydrous titanium oxide was found to have the best properties. A uranium extraction plant was conceptually designed. Of the possible methods for contacting the seawater with the adsorbent, a continuous fluidized bed concept was chosen as most practical for a pumped system. A plant recovering 500 tonnes of U3O8 per year requires 5900 cubic meters per second of seawater to be pumped through the adsorbent beds for a 70% overall recovery efficiency. Total cost of the plant was estimated to be about $6.2 billion. A computer model for the process was used for parametric sensitivity studies and economic projections. Several design case variations were developed. Other topics addressed were the impact of co-product recovery, environmental considerations, etc

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

    Directory of Open Access Journals (Sweden)

    LW Figgs

    2013-07-01

    Full Text Available 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 assembled. A job-specific exposure matrix (JEM was used to determine likely toxic metal exposure categories. In addition, radiation film badge dosimeters were used to monitor cumulative external ionizing radiation exposure. International Classification for Disease (ICD codes 9 and 10 were used to identify 147 lung cancer deaths. Logistic and proportional hazards regression were used to estimate lung cancer mortality risk.Results: Lung cancer mortality risk was elevated among workers who experienced external radiation >3.5 mrem and employment duration >12 years.Conclusion: Employees of uranium gaseous diffusion plants carry a higher risk of lung cancer mortality; the mortality is associated with increased radiation exposure and duration of employment.

  19. Extraction of uranium from seawater: chemical process and plant design feasibility study

    Energy Technology Data Exchange (ETDEWEB)

    Campbell, M.H.; Frame, J.M.; Dudey, N.D.; Kiel, G.R.; Mesec, V.; Woodfield, F.W.; Binney, S.E.; Jante, M.R.; Anderson, R.C.; Clark, G.T.

    1979-02-01

    A major assessment was made of the uranium resources in seawater. Several concepts for moving seawater to recover the uranium were investigated, including pumping the seawater and using natural ocean currents or tides directly. The optimal site chosen was on the southeastern Puerto Rico coast, with the south U.S. Atlantic coast as an alternate. The various processes for extracting uranium from seawater were reviewed, with the adsorption process being the most promising at the present time. Of the possible adsorbents, hydrous titanium oxide was found to have the best properties. A uranium extraction plant was conceptually designed. Of the possible methods for contacting the seawater with the adsorbent, a continuous fluidized bed concept was chosen as most practical for a pumped system. A plant recovering 500 tonnes of U/sub 3/O/sub 8/ per year requires 5900 cubic meters per second of seawater to be pumped through the adsorbent beds for a 70% overall recovery efficiency. Total cost of the plant was estimated to be about $6.2 billion. A computer model for the process was used for parametric sensitivity studies and economic projections. Several design case variations were developed. Other topics addressed were the impact of co-product recovery, environmental considerations, etc.

  20. Tramp uranium

    International Nuclear Information System (INIS)

    Many utilities have implemented a no leaker philosophy for fuel performance and actively pursue removing leaking fuel assemblies from their reactor cores whenever a leaking fuel assembly is detected. Therefore, the only source for fission product activity in the RCS when there are no leaking fuel assemblies is tramp uranium. A technique has been developed that strips uranium impurities from ZrCl4. Unless efforts are made to remove natural uranium impurities from reactor materials, the utilities will not be able to reduce the RCS specific 131I activity in PWRs to below the lower limit of ∼1.0 x 10-4 μCi/g

  1. Irradiation performance of uranium-molybdenum alloy dispersion fuels; Desempenho sob irradiacao de elementos combustiveis do tipo U-Mo

    Energy Technology Data Exchange (ETDEWEB)

    Almeida, Cirila Tacconi de

    2005-07-01

    The U-Mo-Al dispersion fuels of Material Test Reactors (MTR) are analyzed in terms of their irradiation performance. The irradiation performance aspects are associated to the neutronic and thermal hydraulics aspects to propose a new core configuration to the IEA-R1 reactor of IPEN-CNEN/SP using U-Mo-Al fuels. Core configurations using U-10Mo-Al fuels with uranium densities variable from 3 to 8 gU/cm{sup 3} were analyzed with the computational programs Citation and MTRCR-IEA R1. Core configurations for fuels with uranium densities variable from 3 to 5 gU/cm{sup 3} showed to be adequate to use in IEA-R1 reactor e should present a stable in reactor performance even at high burn-up. (author)

  2. Standard guide for pyrophoricity/combustibility testing in support of pyrophoricity analyses of metallic uranium spent nuclear fuel

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2007-01-01

    1.1 This guide covers testing protocols for testing the pyrophoricity/combustibility characteristics of metallic uranium-based spent nuclear fuel (SNF). The testing will provide basic data for input into more detailed computer codes or analyses of thermal, chemical, and mechanical SNF responses. These analyses would support the engineered barrier system (EBS) design bases and safety assessment of extended interim storage facilities and final disposal in a geologic repository. The testing also could provide data related to licensing requirements for the design and operation of a monitored retrievable storage facility (MRS) or independent spent fuel storage installation (ISFSI). 1.2 This guide describes testing of metallic uranium and metallic uranium-based SNF in support of transportation (in accordance with the requirements of 10CFR71), interim storage (in accordance with the requirements of 10CFR72), and geologic repository disposal (in accordance with the requirements of 10CFR60/63). The testing described ...

  3. Production Cycle for Large Scale Fission Mo-99 Separation by the Processing of Irradiated LEU Uranium Silicide Fuel Element Targets

    OpenAIRE

    Abdel-Hadi Ali Sameh

    2013-01-01

    Uranium silicide fuels proved over decades their exceptional qualification for the operation of higher flux material testing reactors with LEU elements. The application of such fuels as target materials, particularly for the large scale fission Mo-99 producers, offers an efficient and economical solution for the related facilities. The realization of such aim demands the introduction of a suitable dissolution process for the applied U3Si2 compound. Excellent results are achieved by the oxidiz...

  4. International symposium on uranium raw material for the nuclear fuel cycle: Exploration, mining, production, supply and demand, economics and environmental issues (URAM-2009). Book of abstracts

    International Nuclear Information System (INIS)

    The International Symposium on Uranium Raw Material for the Nuclear Fuel Cycle: Exploration, Mining, Production, Supply and Demand, Economics and Environmental Issues (URAM-2009) addressed all aspects of the uranium fuel cycle, from the availability of raw materials to the long-term sustainability of nuclear power. The revival of the uranium industry in recent years has caused a dramatic increase in uranium exploration and mining activities in several countries. URAM-2009 was intended to bring together scientists, exploration and mining geologists, engineers, operators, regulators and fuel cycle specialists to exchange information and discuss updated research and current issues in uranium geology and deposits, exploration, mining and processing, production economics, and environmental and legal issues. Contributed papers covered uranium markets and economics (including supply and demand); social licensing in the uranium production cycle; uranium exploration (including uranium geology and deposits); uranium mining and processing; environmental and regulatory issues; human resources development. There was a poster session throughout the symposium, as well as an exhibition of topical photographs. A workshop on recent developments in Technical Cooperation Projects relevant to the Uranium Production Cycle area was also organized. On the last day of the symposium, there was an experts' Panel Discussion. The presentations and discussions at URAM-2009 (a) led to a better understanding of the adequacy of uranium sources (both primary and secondary) to meet future demand, (b) provided information on new exploration concepts, knowledge and technologies that will potentially lead to the discovery and development of new uranium resources, (c) described new production technology having the potential to more efficiently and economically exploit new uranium resources; (d) documented the environmental compatibility of uranium production and the overall effectiveness of the final

  5. Criticality accident in uranium fuel processing plant. Progress and reflection of the criticality accident in the uranium fuel processing plant

    International Nuclear Information System (INIS)

    As one year is already passing since forming of the JCO criticality accident, impact given by this accident was so large as to vibrate all of nuclear energy field. This accident was the first instantly forming criticality accident since beginning of peaceful use in nuclear energy in Japan, which formed some severe victims containing two dead and an experienced affair required for evacuation and shelter of the peripheral inhabitants. Direct cause of the instantly forming criticality accident in this accident is simple and clear, and is caused by failure in the most essential technology specific to nuclear energy called by criticality management. And that, it was caused not by instrument accident or human individual error but by recent exceptional blunder in and out of Japan at a point of direct reason on evil violation act due to management organization. And, for the response specific to the nuclear energy field, a drastic reinvestigation on safety filed, a drastic reinvestigation on safety regulation system is also required. On the other hand, in nuclear safety education requiring establishment of safety culture for its foundation, a reflection that it has remained only to moral action to bring a result to suppress power carrying out its practice inversely, was also recognized. And, it is necessary to carry out more efforts and devices for difficulty on management forecast in future in nuclear energy industry not so as to make a system of safety conservation weaker. (G.K.)

  6. Thermal-hydraulic analysis for core conversion to the use of low-enriched uranium fuels in the KUR

    International Nuclear Information System (INIS)

    A feasibility study has been performed on the core conversion to the use of low-enriched uranium (LEU) fuels in the KUR. Five fuel element geometries are studied. For each fuel element, the relation between the pressure drop and the flow rate, critical heat flux, and heat fluxes for the onset of flow instability and the onset of nucleate boiling are calculated using the computer code PLTEMP3.MOD1 which has been developed for this analysis. The effect of fuel material (UAL/sub x/-Al, U3O8-Al and U3Si2-Al) on the peak fuel temperatures is also studied. A particular interest in the mixed core which may be constructed on the way to the use of LEU fuels, the change in the bypass flow rate due to the change in the gap between different fuel elements is investigated

  7. Methods for analysis of uranium-plutonium mixed fuel and transplutonium elements at RIAR (Preprint no. IT-25)

    International Nuclear Information System (INIS)

    Different methods for analysis of the uranium-plutonium mixed nuclear fuel and transplutonium elements are briefly discussed in this paper: coulometry, radiometric techniques, emission spectrography, mass-spectrometry, chromatography, spectrophotometry. The main analytical characteristics of the methods developed are given. (author). 30 refs., 2 tabs

  8. French approach to plant safety reassessment: The benefits of visits to European nuclear fuel fabrication plants

    International Nuclear Information System (INIS)

    With a ten year periodicity, the French nuclear safety authority (DGSNR) and its technical support (IRSN) reassess the safety of the plants under survey. A special part is devoted to internal experience feedback, and the events that occurred during the past working period are carefully analysed. In the case of the French FBFC uranium fuel fabrication plant, it was decided to make a more systematic use of international feedback in the usual reassessment work for non-reactor facilities. A programme has been set up on the basis of peer to peer level bilateral meetings organized with the European safety authorities having similar plants under their responsibility. A special safety-oriented standard questionnaire was previously sent to each and returned. Each bilateral contact consists of a two day visit, which began with discussions about particular topics in the regulation area and methods for an efficient survey. In order to illustrate this, the two authorities visited the nuclear site the second day. In this way, European safety authorities of five countries (Belgium, Germany, United Kingdom, Spain and Sweden) have met, and plants belonging to Framatome-ANP, Westinghouse and ENUSA groups were visited. All the contacted authorities participated and when this programme was completed, it was possible for DGSNR and IRSN to get a more accurate measure of particular reassessment work. The main results of these actions were to 'standardize' the consultancy's judgement and, taking advantage of encountered good safety practices and good safety designs, it was possible to reinforce specific requests during the reassessment work. Moreover, it was an opportunity for the operators to participate in deepening information exchanges about safety topics. (author)

  9. Recent status and future aspect of plate type fuel element technology with high uranium density at NUKEM

    International Nuclear Information System (INIS)

    According to the present state of development, full size test fuel elements with UAl/sub x/, U3O8, and U3Si2 fuel were fabricated at Nukem in production scale. The maximum uranium densities amount to 1.8 g/cm3 for UAl/sub x/, 2.9 g/cm3 for U3O8, and 4.76 g/cm3 for U3Si2. The irradiation performance of these fuel elements is good: Up to the end of September 1982 the following burnups were achieved: 73% with UAl/sub x/, 60% with U3O8, 39% with U3Si2; no defects could be detected. For an economical fuel element production with reduced 235U enrichment chemical uranium recycling methods were developed allowing immediate scrap recovery at minimum waste generation. In addition test plates with UAl/sub x/ and U3O8 fuel were successfully irradiated in the ORR up to a burnup of 75%. The relatively high uranium meat densities of these test plates amount to 2.2 g/cc for UAl/sub x/, and 3.14 g/cc for U3O8 fuel. Apart from plates with standard geometry also plates with increased meat thickness were inserted

  10. Recent status and future aspect of plate type fuel element technology with high uranium density at NUKEM

    International Nuclear Information System (INIS)

    According to the present state of development full size test fuel elements with UAlx, U3O8, and U3Si2 fuel were fabricated at Nukem in production scale. The maximum uranium densities amount to 1.8 g/cc for UAIx, 2.9 g/cc for U3O8, and 4.76 g/cc for U3Si2. The irradiation performance of these fuel elements is good: Up to the end of September 1982 the following burnups were achieved: 73% with UA1x, 60% with U3O8, 39% with U3Si2; no defects could be detected. For an economical fuel element production with reduced 235-U enrichment chemical uranium recycling methods were developed allowing immediate scrap recovery at minimum waste generation. In addition test plates with UAlx and U3O8 fuel were successfully irradiated in the ORR up to a burnup of 75 %. The relatively high uranium meat densities of these test plates amount to 2.2 g/cc for UAI x, and 3.14 g/cc for U3O8 fuel. Apart from plates with standard geometry also plates with increased meat thickness were inserted. (author)

  11. Analysis of nuclear engineering danger factors in NPP fuel cycle (in connection with developing the Russian Federal purposeful program of nuclear and radiation safety)

    International Nuclear Information System (INIS)

    The main steps of nuclear fuel cycle being within the field of monitoring and realization of the Russian Federal purposeful program of nuclear and radiation safety are discussed. These include the following items: uranium and thorium mines, hydrometallurgical plants for mineral uranium and thorium raw materials reprocessing, refining of nuclear materials, uranium hexafluoride, isotope separation plants, reactor fuel manufacturing facilities, transportation of the fuel and intermediate products, spent fuel storage facilities, spent fuel regeneration facilities, radioactive waste storage facilities

  12. Fuel handling and storage systems in nuclear power plants

    International Nuclear Information System (INIS)

    The scope of this Guide includes the design of handling and storage facilities for fuel assemblies from the receipt of fuel into the nuclear power plant until the fuel departs from that plant. The unirradiated fuel considered in this Guide is assumed not to exhibit any significant level of radiation so that it can be handled without shielding or cooling. This Guide also gives limited consideration to the handling and storage of certain core components. While the general design and safety principles are discussed in Section 2 of this Guide, more specific design requirements for the handling and storage of fuel are given in detailed sections which follow the general design and safety principles. Further useful information is to be found in the IAEA Technical Reports Series No. 189 ''Storage, Handling and Movement of Fuel and Related Components at Nuclear Power Plants'' and No. 198 ''Guide to the Safe Handling of Radioactive Wastes at Nuclear Power Plants''. However, the scope of the Guide does not include consideration of the following: (1) The various reactor physics questions associated with fuel and absorber loading and unloading into the core; (2) The design aspects of preparation of the reactor for fuel loading (such as the removal of the pressure vessel head for a light water reactor) and restoration after loading; (3) The design of shipping casks; (4) Fuel storage of a long-term nature exceeding the design lifetime of the nuclear power plant; (5) Unirradiated fuel containing plutonium

  13. Hot Hydrogen Testing of Tungsten-Uranium Dioxide (W-UO2) CERMET Fuel Materials for Nuclear Thermal Propulsion

    Science.gov (United States)

    Hickman, Robert; Broadway, Jeramie

    2014-01-01

    CERMET fuel materials are being developed at the NASA Marshall Space Flight Center for a Nuclear Cryogenic Propulsion Stage. Recent work has resulted in the development and demonstration of a Compact Fuel Element Environmental Test (CFEET) System that is capable of subjecting depleted uranium fuel material samples to hot hydrogen. A critical obstacle to the development of an NCPS engine is the high-cost and safety concerns associated with developmental testing in nuclear environments. The purpose of this testing capability is to enable low-cost screening of candidate materials, fabrication processes, and further validation of concepts. The CERMET samples consist of depleted uranium dioxide (UO2) fuel particles in a tungsten metal matrix, which has been demonstrated on previous programs to provide improved performance and retention of fission products1. Numerous past programs have utilized hot hydrogen furnace testing to develop and evaluate fuel materials. The testing provides a reasonable simulation of temperature and thermal stress effects in a flowing hydrogen environment. Though no information is gained about radiation damage, the furnace testing is extremely valuable for development and verification of fuel element materials and processes. The current work includes testing of subscale W-UO2 slugs to evaluate fuel loss and stability. The materials are then fabricated into samples with seven cooling channels to test a more representative section of a fuel element. Several iterations of testing are being performed to evaluate fuel mass loss impacts from density, microstructure, fuel particle size and shape, chemistry, claddings, particle coatings, and stabilizers. The fuel materials and forms being evaluated on this effort have all been demonstrated to control fuel migration and loss. The objective is to verify performance improvements of the various materials and process options prior to expensive full scale fabrication and testing. Post test analysis will

  14. Cimarrones y palenques en las provincias al norte del Nuevo Reino de Granada siglo XVII

    Directory of Open Access Journals (Sweden)

    María Cristina Navarrete P.

    2001-01-01

    Full Text Available A finales del siglo XVII, ya existían comunidades de cimarrones en varias provincias de la jurisdicción de la Audiencia del Nuevo Reino de Granada. Hacia 1598, en Zaragoza, y en 1607, en Remedios, provincia de Antioquia, esclavos negros de las rancherías mineras se levantaron violentamente y fortificaron en palenques, causando perjuicios en el comercio, la vida de las ciudades y la labor de las minas. Otro de los más importantes alzamientos de esclavos fue el dirigido por Domingo Biohó quien se refugió con sus seguidores en la Ciénaga de la Matuna. Asimismo, las Sierras de María y la Serranía de San Lucas fueron espacios de establecimiento de palenques. Los cimarrones intentaron reconstruir organizaciones sociales con base en la herencia cultural africana, en la permanencia en las haciendas y minas al contacto con amos, capataces y compañeros de circunstancias y, en las nuevas condiciones que les ofrecía la vida de los palenques.

  15. 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...

  16. Fuel ethanol production using nuclear-plant steam

    International Nuclear Information System (INIS)

    In the United States, the production of fuel ethanol from corn for cars and light trucks has increased from about 6 billion liters per year in 2000 to 19 billion liters per year in 2006. A third of the world's liquid fuel demands could ultimately be obtained from biomass. The production of fuel ethanol from biomass requires large quantities of steam. For a large ethanol plant producing 380 million liters of fuel ethanol from corn per year, about 80 MW(t) of 1-MPa (∼180 deg. C) steam is required. Within several decades, the steam demand for ethanol plants in the United States is projected to be tens of gigawatts, with the worldwide demand being several times larger. This market may become the largest market for cogeneration of steam from nuclear electric power plants. There are strong incentives to use steam from nuclear power plants to meet this requirement. The cost of low-pressure steam from nuclear power plants is less than that of natural gas, which is now used to make steam in corn-to-ethanol plants. Steam from nuclear power plants reduces greenhouse gases compared with steam produced from fossil fuels. While ethanol is now produced from sugarcane and corn, the next-generation ethanol plants will use more abundant cellulose feedstocks. It is planned that these plants will burn the lignin in the cellulosic feedstocks to provide the required steam. Lignin is the primary non-sugar-based component in cellulosic biomass that can not be converted to ethanol. Low-cost steam from nuclear plants creates the option of converting the lignin to other liquid fuels and thus increase the liquid fuel production per unit of biomass. Because liquid fuel production from biomass is ultimately limited by the availability of biomass, steam from nuclear plants can ultimately increase the total liquid fuels produced from biomass. (author)

  17. Long term developments in irradiated natural uranium processing costs. Optimal size and siting of plants; Perspectives a long terme des couts de traitement de l'uranium naturel irradie. Tailles et localisations optimales des usines

    Energy Technology Data Exchange (ETDEWEB)

    Thiriet, L. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Oger, C.; Vaumas, P. de [Saint-Gobain Nucleaire, 92 - Courbevoie (France)

    1964-07-01

    The aim of this paper is to help solve the problem of the selection of optimal sizes and sites for spent nuclear fuel processing plants associated with power capacity programmes already installed. Firstly, the structure of capital and running costs of irradiated natural uranium processing plants is studied, as well as the influence of plant sizes on these costs and structures. Shipping costs from the production site to the plant must also be added to processing costs. An attempt to reach a minimum cost for the production of a country or a group of countries must therefore take into account both the size and the location of the plants. The foreseeable shipping costs and their structure (freight, insurance, container cost and depreciation), for spent natural uranium are indicated. Secondly, for various annual spent fuel reprocessing programmes, the optimal sizes and locations of the plants are determined. The sensitivity of the results to the basic assumptions relative to processing costs, shipping costs, the starting up year of the plant programme and the length of period considered, is also tested. - this rather complex problem, of a combinative nature, is solved through dynamic programming methods. - It is shown that these methods can also be applied to the problem of selecting the optimal sizes and locations of processing plants for MTR type fuel elements, related to research reactor programmes, as well as to future plutonium element processing plants related to breeder reactors. Thirdly, the case where yearly extraction of the plutonium contained in the irradiated natural uranium is not compulsory is examined; some stockpiling of the fuel is then allowed some years, entailing delayed processing. The load factor of such plants is thus greatly improved with respect to that of plants where the annual plutonium demand is strictly satisfied. By including spent natural uranium stockpiling costs an optimal rhythm of introduction and optimal sizes for spent fuel

  18. Effect of transplutonium doping on approach to long-life core in uranium-fueled PWR

    Energy Technology Data Exchange (ETDEWEB)

    Peryoga, Yoga; Saito, Masaki; Artisyuk, Vladimir [Tokyo Inst. of Tech. (Japan). Research Lab. for Nuclear Reactors; Shmelev, Anatolii [Moscow Engineering Physics Institute, Moscow (Russian Federation)

    2002-08-01

    The present paper advertises doping of transplutonium isotopes as an essential measure to improve proliferation-resistance properties and burnup characteristics of UOX fuel for PWR. Among them {sup 241}Am might play the decisive role of burnable absorber to reduce the initial reactivity excess while the short-lived nuclides {sup 242}Cm and {sup 244}Cm decay into even plutonium isotopes, thus increasing the extent of denaturation for primary fissile {sup 239}Pu in the course of reactor operation. The doping composition corresponds to one discharged from a current PWR. For definiteness, the case identity is ascribed to atomic percentage of {sup 241}Am, and then the other transplutonium nuclide contents follow their ratio as in the PWR discharged fuel. The case of 1 at% doping to 20% enriched uranium oxide fuel shows the potential of achieving the burnup value of 100 GWd/tHM with about 20% {sup 238}Pu fraction at the end of irradiation. Since so far, americium and curium do not require special proliferation resistance measures, their doping to UOX would assist in introducing nuclear technology in developing countries with simultaneous reduction of accumulated minor actinides stockpiles. (author)

  19. Role of uranium speciation on its bioaccumulation, transfer and toxicity in plants. Application to phyto-remediation

    International Nuclear Information System (INIS)

    Uranium is both a radiological and a chemical toxic, which naturally occurs in the environment as a trace element. Metal accumulation and distribution in plants is modulated by speciation. The aim of this PhD work was thus to assay uranium accumulation, intra planta repartition and toxicity according to its speciation in solution. Acquired knowledge will be applied in phyto-remediation technologies. We exposed three plant species (sunflower, oilseed rape and wheat) to a panel of hydroponic media containing one or two predominant uranium chemical forms. After exposition in these various contaminated media, we evaluated uranium content in plant organs by ICP-MS. In order to investigate uranium repartition and localization at organ/tissue and cellular scales, we carried out four complementary imaging techniques. The uranium repartition within soluble and membrane fractions in roots and shoot was assayed after fractionation and separation through a chromatography column. In parallel, we used X-ray absorption spectroscopy to determine the molecular-level structure of chemical species formed by uranium in exposure media and plant samples. Finally, we explored toxic effects of uranium on plant growth and metabolism. Our results revealed three schema of accumulation according to the uranium speciation in the exposure medium: when exposed to UO22+ free ion, root accumulation is high, but uranium transfer to the shoots is limited. Uranium is immobilized by adsorption on root surface and precipitation on root cell walls, associated with phosphorus and calcium. The existence of uranium-binding proteins is also suggested. When complexed with phosphate, root accumulation is considerably reduced and translocation becomes negligible. Uranium is precipitated as described above. Conversely, complexation with carbonate or citrate reduces root accumulation but drastically increases translocation to the shoots. If some uranyl phosphate precipitates are still found in root and shoot, a

  20. Chemical analysis used in nuclear fuels reprocessing of uranium and thorium

    International Nuclear Information System (INIS)

    An overall review of the analytical chemistry in nuclear fuel reprocessing is done. In Purex and Thorex process flowsheets, the analyses required to the control of the process, balance and accountability of fissile and fertile materials, and final product specification are pointed out. Some analytical methods applied to the determination of uranium, plutonium, thorium, nitric acid, tributylphosphate and fission products are described. Specific features of the analytical laboratories are presented. The radioactivity level of the samples requires facilities as shielded cells and glove boxes, and handling by remote control. Finally it is reported an application of one analytical method to evaluate thorium content in organic and aqueous solutions, in cold tests of Thorex process. These tests were performed at CDTN/NUCLEBRAS. (author)

  1. RA-3 reactor core with uranium silicide fuel elements P-07 type

    International Nuclear Information System (INIS)

    Following the studies on the utilization of fuel elements (FE) containing uranium silicide, core of the RA-3 was analyzed with several calculation models. At first, the present situation, i.e. the core charged with normal FE (U3O8), has been analyzed to validate the simulation methodology comparing with experimental results and to establish reference data to 5 and 10 MW able to be compared with future new situations. Also, CITVAP's nuclear data libraries to be used in irradiation experiment planning were completed. The results were satisfactory and were applied to the study of the core containing P-07 FE [U3Si2], in face of a future core change. Comparing with the performance of the U3O8FE, the silicides ones show the following advantages: - average burnup: 45 % greater; -extraction burnup increase 12 %; and, -the residence time [in full power days] could be a 117 % greater. (author)

  2. Environmental impacts of fossil-fuel and nuclear power plants

    International Nuclear Information System (INIS)

    Large power plants burning fossil fuels generate emissions with a high content of sulphur dioxide and a content of noxious aerosols and radioisotopes whose radioactivity exceeds the limits set for nuclear power plants. The main problem of nuclear power plants is to secure radiation safety namely in case of an accident even though the probability of such an event is very small. The most complicated problems are related to the treatment of spent fuel, its transport, processing and storage. (B.H.)

  3. Uranium hexafluoride handling

    International Nuclear Information System (INIS)

    The United States Department of Energy, Oak Ridge Field Office, and Martin Marietta Energy Systems, Inc., are co-sponsoring this Second International Conference on Uranium Hexafluoride Handling. The conference is offered as a forum for the exchange of information and concepts regarding the technical and regulatory issues and the safety aspects which relate to the handling of uranium hexafluoride. Through the papers presented here, we attempt not only to share technological advances and lessons learned, but also to demonstrate that we are concerned about the health and safety of our workers and the public, and are good stewards of the environment in which we all work and live. These proceedings are a compilation of the work of many experts in that phase of world-wide industry which comprises the nuclear fuel cycle. Their experience spans the entire range over which uranium hexafluoride is involved in the fuel cycle, from the production of UF6 from the naturally-occurring oxide to its re-conversion to oxide for reactor fuels. The papers furnish insights into the chemical, physical, and nuclear properties of uranium hexafluoride as they influence its transport, storage, and the design and operation of plant-scale facilities for production, processing, and conversion to oxide. The papers demonstrate, in an industry often cited for its excellent safety record, continuing efforts to further improve safety in all areas of handling uranium hexafluoride

  4. Uranium hexafluoride handling. Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1991-12-31

    The United States Department of Energy, Oak Ridge Field Office, and Martin Marietta Energy Systems, Inc., are co-sponsoring this Second International Conference on Uranium Hexafluoride Handling. The conference is offered as a forum for the exchange of information and concepts regarding the technical and regulatory issues and the safety aspects which relate to the handling of uranium hexafluoride. Through the papers presented here, we attempt not only to share technological advances and lessons learned, but also to demonstrate that we are concerned about the health and safety of our workers and the public, and are good stewards of the environment in which we all work and live. These proceedings are a compilation of the work of many experts in that phase of world-wide industry which comprises the nuclear fuel cycle. Their experience spans the entire range over which uranium hexafluoride is involved in the fuel cycle, from the production of UF{sub 6} from the naturally-occurring oxide to its re-conversion to oxide for reactor fuels. The papers furnish insights into the chemical, physical, and nuclear properties of uranium hexafluoride as they influence its transport, storage, and the design and operation of plant-scale facilities for production, processing, and conversion to oxide. The papers demonstrate, in an industry often cited for its excellent safety record, continuing efforts to further improve safety in all areas of handling uranium hexafluoride. Selected papers were processed separately for inclusion in the Energy Science and Technology Database.

  5. Spent-fuel-storage studies at the Barnwell Nuclear Fuel Plant. Studies and research concerning BNFP

    International Nuclear Information System (INIS)

    This report contains the results of various studies and demonstrations related to advanced spent-fuel-storage techniques which were performed at the Barnwell Nuclear Fuel Plant (BNFP) in 1982. The demonstrations evaluated various technical aspects of fuel disassembly and canning and dry-storage techniques. The supporting studies examined thermal limitations and criticality concerns

  6. Screening of plant species for phytoremediation of uranium, thorium, barium, nickel, strontium and lead contaminated soils from a uranium mill tailings repository in South China.

    Science.gov (United States)

    Li, Guang-yue; Hu, Nan; Ding, De-xin; Zheng, Ji-fang; Liu, Yu-long; Wang, Yong-dong; Nie, Xiao-qin

    2011-06-01

    The concentrations of uranium, thorium, barium, nickel, strontium and lead in the samples of the tailings and plant species collected from a uranium mill tailings repository in South China were analyzed. Then, the removal capability of a plant for a target element was assessed. It was found that Phragmites australis had the greatest removal capabilities for uranium (820 μg), thorium (103 μg) and lead (1,870 μg). Miscanthus floridulus had the greatest removal capabilities for barium (3,730 μg) and nickel (667 μg), and Parthenocissus quinquefolia had the greatest removal capability for strontium (3,920 μg). In this study, a novel coefficient, termed as phytoremediation factor (PF), was proposed, for the first time, to assess the potential of a plant to be used in phytoremediation of a target element contaminated soil. Phragmites australis has the highest PFs for uranium (16.6), thorium (8.68), barium (10.0) and lead (10.5). Miscanthus floridulus has the highest PF for Ni (25.0). Broussonetia papyrifera and Parthenocissus quinquefolia have the relatively high PFs for strontium (28.1 and 25.4, respectively). On the basis of the definition for a hyperaccumulator, only Cyperus iria and Parthenocissus quinquefolia satisfied the criteria for hyperaccumulator of uranium (36.4 μg/g) and strontium (190 μg/g), and could be the candidates for phytoremediation of uranium and strontium contaminated soils. The results show that the PF has advantage over the hyperaccumulator in reflecting the removal capabilities of a plant for a target element, and is more adequate for assessing the potential of a plant to be used in phytoremediation than conventional method.

  7. Preparation for commissioning of nuclear plant with reference to British Nuclear Fuels Plc fuel handling plant project

    International Nuclear Information System (INIS)

    The new Fuel Handing Plant at British Nuclear Fuels Sellafield is part of a Pound 550M complex which provides facilities for the receipt, storage and mechanical preparation of both magnox and A.G.R. fuel. The plant is very large and complex with considerable use of computer based process control systems, providing for physical and nuclear safety. The preparation of such plant for ''active'' commissioning necessitates a great many physical checks and technical evaluations in support of its safety case. This paper describes arrangements for plant commissioning checks, against the regulatory framework and explains the physical preparations necessary for their timely accomplishment. (author)

  8. Uranium, its impact on the national and global energy mix; and its history, distribution, production, nuclear fuel-cycle, future, and relation to the environment

    Science.gov (United States)

    Finch, Warren Irvin

    1997-01-01

    The many aspects of uranium, a heavy radioactive metal used to generate electricity throughout the world, are briefly described in relatively simple terms intended for the lay reader. An adequate glossary of unfamiliar terms is given. Uranium is a new source of electrical energy developed since 1950, and how we harness energy from it is explained. It competes with the organic coal, oil, and gas fuels as shown graphically. Uranium resources and production for the world are tabulated and discussed by country and for various energy regions in the United States. Locations of major uranium deposits and power reactors in the United States are mapped. The nuclear fuel-cycle of uranium for a typical light-water reactor is illustrated at the front end-beginning with its natural geologic occurrence in rocks through discovery, mining, and milling; separation of the scarce isotope U-235, its enrichment, and manufacture into fuel rods for power reactors to generate electricity-and at the back end-the reprocessing and handling of the spent fuel. Environmental concerns with the entire fuel cycle are addressed. The future of the use of uranium in new, simplified, 'passively safe' reactors for the utility industry is examined. The present resource assessment of uranium in the United States is out of date, and a new assessment could aid the domestic uranium industry.

  9. BASIC program to compute uranium density and void volume fraction in laboratory-scale uranium silicide aluminum dispersion plate-type fuel

    International Nuclear Information System (INIS)

    BASIC program simple and easy to operate has been developed to compute uranium density and void volume fraction for laboratory-scale uranium silicide aluminum dispersion plate-type fuel, so called miniplate. An example of the result of calculation is given in order to demonstrate how the calculated void fraction correlates with the microstructural distribution of the void in a miniplate prepared in our laboratory. The program is also able to constitute data base on important parameters for miniplates from experimentally-determined values of density, weight of each constituent and dimensions of miniplates. Utility programs pertinent to the development of the BASIC program are also given which run in the popular MS-DOS environment. All the source lists are attached and brief description for each program is made. (author)

  10. Uptake of uranium by native aquatic plants: potential for bioindication and phytoremediation

    Directory of Open Access Journals (Sweden)

    Favas P. J. C.

    2013-04-01

    Full Text Available The work presented here is a part the on going study on the uraniferous geochemical province of Central Portugal in which, the use of aquatic plants as indicators of uranium contamination is being probed using aquatic plants emphasizing their potential use in the emerging phytotechnologies. Even though we have observed very low concentration of U in the fresh waters of the studied sites we found a set of vegetable species with the ability to accumulate U in concentrations which are orders of magnitude higher than the surrounding environment. We have observed that Apium nodiflorum, Callitriche stagnalis, Lemna minor and Fontinalis antipyretica accumulated significant amounts of uranium, whereas Oenanthe crocata excluded U. These results indicate substantial scope for proper radiophytoremediation and phytosociological investigation exploiting the native flora. These species show great potential for phytoremediation because they are endemic and easy to grow in their native conditions. A. nodiflorum and C. stagnalis have high bioproductivity and yield good biomass.

  11. Cost update technology, safety, and costs of decommissioning a reference uranium hexafluoride conversion plant

    Energy Technology Data Exchange (ETDEWEB)

    Miles, T.L.; Liu, Y.

    1995-08-01

    The purpose of this study is to update the cost estimates developed in a previous report, NUREG/CR-1757 (Elder 1980) for decommissioning a reference uranium hexafluoride conversion plant from the original mid-1981 dollars to values representative of January 1993. The cost updates were performed by using escalation factors derived from cost index trends over the past 11.5 years. Contemporary price quotes wee used for costs that have increased drastically or for which is is difficult to find a cost trend. No changes were made in the decommissioning procedures or cost element requirements assumed in NUREG/CR-1757. This report includes only information that was changed from NUREG/CR-1757. Thus, for those interested in detailed descriptions and associated information for the reference uranium hexafluoride conversion plant, a copy of NUREG/CR-1757 will be needed.

  12. Progress and prospects for phosphoric acid fuel cell power plants

    Energy Technology Data Exchange (ETDEWEB)

    Bonville, L.J.; Scheffler, G.W.; Smith, M.J. [International Fuel Cells Corp., South Windsor, CT (United States)

    1996-12-31

    International Fuel Cells (IFC) has developed the fuel cell power plant as a new, on-site power generation source. IFC`s commercial fuel cell product is the 200-kW PC25{trademark} power plant. To date over 100 PC25 units have been manufactured. Fleet operating time is in excess of one million hours. Individual units of the initial power plant model, the PC25 A, have operated for more than 30,000 hours. The first model {open_quotes}C{close_quotes} power plant has over 10,000 hours of operation. The manufacturing, application and operation of this power plant fleet has established a firm base for design and technology development in terms of a clear understanding of the requirements for power plant reliability and durability. This fleet provides the benchmark against which power plant improvements must be measured.

  13. A study on possible use of Urtica dioica (common nettle) plants as uranium (234U, 238U) contamination bioindicator near phosphogypsum stockpile

    OpenAIRE

    Olszewski, Grzegorz; Boryło, Alicja; Skwarzec, Bogdan

    2015-01-01

    The aim of this study was to determine uranium concentrations in common nettle (Urtica dioica) plants and corresponding soils samples which were collected from the area of phosphogypsum stockpile in Wiślinka (northern Poland). The uranium concentrations in roots depended on its concentrations in soils. Calculated BCF and TF values showed that soils characteristics and air deposition affect uranium absorption and that different uranium species have different affinities to U. dioica plants. The...

  14. Design of a 252Cf neutron assay system for evaluation at the Savannah River Plant fuel fabrication facility

    International Nuclear Information System (INIS)

    A nondestructive assay (NDA) unit will be evaluated at the Savannah River Plant (SRP) reactor fuel fabrication facility for measurement of a range of highly enriched uranium materials. The unit employs cyclic neutron interrogation with a Cf-252 neutron source followed by delayed-neutron counting to assay the U-235 content of fuel alloys with up to 2.4 kg U-235 per item in addition to scrap and waste with a lower U-235 content. The accuracy goal for the majority of the measurements is 1 to 3 percent

  15. Uranium and radium in water samples around the Nikola Tesla B lignite-fired power plant - Obrenovac, Serbia

    OpenAIRE

    Žunić Zora S.; Mietelski Jerzy W.; Radanović Sanja B.; Kierepko Renata; Ciotoli Giancarlo; Čeliković Igor T.; Ujić Predrag N.; Kisić Dragica M.; Bartyzel Miroslaw; Bogacz Joanna; Udovičić Vladimir I.; Simović Rodoljub D.

    2011-01-01

    This paper deals with the analysis of natural radionuclide content in 23 water samples collected in the vicinity of the Nikola Tesla B thermal power plant, Serbia. All samples were analyzed for 226Ra and uranium isotopes (238U, 234U) activity using radiochemical methods and alpha spectrometry. Obtained results show that the activity concentrations for uranium and radium in the water around the thermal power plant are low when compared to those from areas across Serbia with their enhance...

  16. Biogeochemical investigation in south eastern Andhra Pradesh: the distribution of rare earths, thorium and uranium in plants and soils

    Energy Technology Data Exchange (ETDEWEB)

    Raju, K.K.; Raju, A.N. [Sri Venkateswara Univ., Tirupati (India). Dept. of Geology

    2000-09-01

    The concentration of rare earth elements (REE), thorium and uranium were determined by inductively coupled plasma mass spectrometry (ICP-MS) in the plant species, Pterocarpus santalinus, P. marsupium and P. dalbergioides, and the soils on which they were growing. Higher concentrations of lanthanum (La), cerium (Ce) were observed in both plants and soils. Large amounts of thorium and uranium were found in the soil. In all tree species, the concentration of REEs were higher in the heartwood than the leaves. The heartwood of P. santalinus accumulated larger quantities of uranium (average concentration of 1.22 ppm) and thorium (mean value of 2.57 ppm) than the other two species. (orig.)

  17. Biogeochemical investigation in south eastern Andhra Pradesh: the distribution of rare earths, thorium and uranium in plants and soils

    International Nuclear Information System (INIS)

    The concentration of rare earth elements (REE), thorium and uranium were determined by inductively coupled plasma mass spectrometry (ICP-MS) in the plant species, Pterocarpus santalinus, P. marsupium and P. dalbergioides, and the soils on which they were growing. Higher concentrations of lanthanum (La), cerium (Ce) were observed in both plants and soils. Large amounts of thorium and uranium were found in the soil. In all tree species, the concentration of REEs were higher in the heartwood than the leaves. The heartwood of P. santalinus accumulated larger quantities of uranium (average concentration of 1.22 ppm) and thorium (mean value of 2.57 ppm) than the other two species. (orig.)

  18. 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...

  19. 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...

  20. 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...

  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. Recent activities and trends in the uranium market

    International Nuclear Information System (INIS)

    Implementing the large number of nuclear power plant projects worldwide presupposes a considerable increase in the production of natural uranium. Preparations have been made: Higher uranium prices stimulate investments into future mines and into uranium exploration. In some countries, the uranium industry is undergoing structural changes so as to be able to meet future requirements. The terms and conditions laid down in long-term uranium supply contracts (prices and fixed delivery volumes) provide present and future producers with the necessary security in investing and planning. The electricity utilities have accepted the shift from a former 'buyer's market' to a 'seller's market' and adapted their uranium supply strategies accordingly. Numerous uranium mines, most of them small, with relatively low uranium ore concentrations, are under construction or in the commissioning phase. However, as secondary sources (fuels not made up of fresh uranium) are gradually coming to an end, many more uranium deposits need to be found and developed to commercial maturity in order to ensure uranium supply also on the long term. The steadily growing industries in the front end and the back end of the fuel cycle have intensified concerns about the non-proliferation of nuclear fuels. However, political considerations with respect to proliferation resistant uranium supply strategies have met with scepticism right from the outset. (orig.)

  3. Development of UO2/PuO2 dispersed in uranium matrix CERMET fuel system for fast reactors

    International Nuclear Information System (INIS)

    CERMET fuel with either PuO2 or enriched UO2 dispersed in uranium metal matrix has a strong potential of becoming a fuel for the liquid metal cooled fast breeder reactors (LMR’s). In fact it may act as a bridge between the advantages and disadvantages associated with the two extremes of fuel systems (i.e. ceramic fuel and metallic fuel) for fast reactors. At Bhabha Atomic Research Centre (BARC), R and D efforts are on to develop this CERMET fuel by powder metallurgy route. This paper describes the development of flow sheet for preparation of UO2 dispersed in uranium metal matrix pellets for three different compositions i.e. U–20 wt%UO2, U–25 wt%UO2 and U–30 wt%UO2. It was found that the sintered pellets were having excellent integrity and their linear mass was higher than that of carbide fuel pellets used in Fast Breeder Test Reactor programme (FBTR) in India. The pellets were characterized by X-ray diffraction (XRD) technique for phase analysis and lattice parameter determination. The optical microstructures were developed and reported for all the three different U–UO2 compositions.

  4. Plant-scale anodic dissolution of unirradiated IFR fuel pins

    International Nuclear Information System (INIS)

    This report discusses anodic dissolution which is a major operation in the pyrometallurgical process for recycling spent metal fuels from the Integral Fast Reactor (IFR), an advanced reactor design developed at Argonne National Laboratory. This process involves electrorefining the heavy metals (uranium and plutonium) from chopped, steel-clad fuel segments. The heavy metals are electrotransported from anodic dissolution baskets to solid and liquid cathodes in a molten salt electrolyte (LiCl-KCI) at 500 degrees C. Uranium is recovered on a solid cathode mandrel, while a uranium-plutonium mixture is recovered in a liquid cadmium cathode. The anode configuration consists of four baskets mounted on an anode shaft. These baskets provide parallel circuits in the electrolyte and salt flow through the chopped fuelbed as the baskets are rotated. The baskets for the engineering-scale tests were sized to contain up to 2.5 kg of heavy metal. Anodic dissolution of 10 kg batches of chopped, steel-clad simulated tuel (U-10% Zr and U-Zr-Fs alloy) was demonstrated

  5. Considerations for handling failed fuel at the Barnwell Nuclear Fuel Plant

    International Nuclear Information System (INIS)

    The impact of failed fuel receipt on reprocessing operations is qualitatively described. It appears that extended storage of fuel, particularly with advanced storage techniques, will increase the quantity of failed fuel, the nature and possibly the configuration of the fuel. The receipt of failed fuel at the BNFP increases handling problems, waste volumes, and operator exposure. If it is necessary to impose special operating precautions to minimize this impact, a loss in plant throughput will result. Hence, ideally, the reprocessing plant operator would take every reasonable precaution so that no failed fuel is received. An alternative policy would be to require that failed fuel be placed in a sealed canister. In the latter case the canister must be compatible with the shipping cask and suitable for in-plant storage. A required inspection of bare fuel would be made at the reactor prior to shipping off-site. This would verify fuel integrity. These requirements are obviously idealistic. Due to the current uncertain status of reprocessing and the need to keep reactors operating, business or governmental policy may be enacted resulting in the receipt of a negotiated quantity of non-standard fuel (including failed fuel). In this situation, BNFP fuel receiving policy based soley on fuel cladding integrity would be difficult to enforce. There are certain areas where process incompatibility does exist and where a compromise would be virtually impossible, e.g., canned fuel for which material or dimensional conflicts exist. This fuel would have to be refused or the fuel would require recanning prior to shipment. In other cases, knowledge of the type and nature of the failure may be acceptable to the operator. A physical inspection of the fuel either before shipment or after the cask unloading operation would be warranted. In this manner, concerns with pool contamination can be identified and the assembly canned if deemed necessary

  6. Air strikes on uranium enrichment plants as potential sources of radioecological danger

    Directory of Open Access Journals (Sweden)

    Živanov Dragan

    2007-01-01

    Full Text Available According to the Non Proliferation Treaty (NPT, the signatory countries are not forbidden to preform uranium enrichment for peaceful purposes. However, if there is a justified doubt that the uranium enrichment is performed with the aim to produce nuclear weapons, this certainly causes great concern. In this case, the international community can apply pressure to a certain country if it determines that the country does not want to cease activities of making its own nuclear weapons. The international community pressure on the country can be intesified until its political leadership is not made to question and cease all activities of producing nuclear weapons. This pressure can be political, economic, and as a last resort-military. As a gesture of goodwill the country can stop the uranium enrichment process. In this way, the country shows that it finally gives up the intention to produce nuclear weapons. However, when military pressure is applied, i.e. military strikes (air strikes for example on nuclear plants used for uranium enrichment, this certainly creates a risk of releasing radioactivity into the environment. That is why the aim of this paper is to signal this very fact. Using military force in these cases leads to additional radioactive contamination of the environment, so this way of solving conflicts should be avoided within the international community.

  7. Radon transport from uranium mill tailings via plant transpiration. Final report

    International Nuclear Information System (INIS)

    Radon exhalation by vegetation planted on bare or soil-covered uranium mill wastes was studied based on an assumption that radon transport from soil to atmosphere via plants takes place in the transpiration stream. Results show that radon exhalation by plants is inversely related to water transpired, primarily a dilution effect. Radon released appeared directly related to leaf area, suggesting that radon is carried into the plant by mass flow in water; however, once within the plant, radon very likely diffuses through the entire leaf cuticle, while water vapor diffuses primarily through open stomates. Application of a computerized model for water transpiration to radon exhalation is not immediately useful until the role of water in radon transport is defined throughout the continuum from rooting medium to the atmosphere. Until then, a simple calculation based on leaf area index and Ra-226 concentration in the rooting medium can provide an estimate of radon release from revegetated wastes containing radium

  8. Use of Pilot Plants for Developing Used Nuclear Fuel Recycling Facilities

    International Nuclear Information System (INIS)

    EnergySolutions and its teaming partners are working with the US Department of Energy (DOE) to develop processes, equipment and facilities for recycling used nuclear fuel (UNF). Recycling significantly reduces the volume of wastes that ultimately will be consigned to the National Geologic Repository, enables the re-use in new fuel of the valuable uranium and plutonium in the UNF, and allows the long-lived minor actinides to be treated separately so they do not become long term heat emitters in the Repository. A major requirement of any new UNF recycling facility is that pure plutonium is not separated anywhere in the process, so as to reduce the nuclear proliferation attractiveness of the facility. EnergySolutions and its team partner the UK National Nuclear Laboratory (NNL) have developed the NUEX process to achieve this and to handle appropriately the treatment of other species such as krypton, tritium, neptunium and technetium. NUEX is based on existing successful commercial UNF recycling processes deployed in the UK, France and imminently in Japan, but with a range of modifications to the flowsheet to keep some uranium with the plutonium at all times and to minimize aerial and liquid radioactive discharges. NNL's long-term experience in developing the recycling and associated facilities at the Sellafield site in the UK, and its current duties to support technically the operation of the Thermal Oxide Reprocessing Plant (THORP) at Sellafield provides essential input to the design of the US NUEX-based facility. Development work for THORP and other first-of-kind nuclear plants employed miniature scale fully radioactive through large scale inactive pilot plants. The sequence of development work that we have found most successful is to (i) perform initial process development at small (typically 1/5000) scale in gloveboxes using trace active materials, (ii) demonstrate the processes at the same small scale with actual irradiated fuel in hot cells and (iii) demonstrate

  9. Development of a computer systems for operational data acquisition of uranium isotopic enrichment pilot plant

    International Nuclear Information System (INIS)

    A pilot plant for uranium enrichment using the jet nozzle process was transfered from Federal Republic of Germany to Brazil, to train Brazilian technicist in its operation and to improve the process. This pilot plant is monitored by a data acquisition system and the possibility of faulty events would cause serious dificulties, as far as maintenance is concerned (for instance, unvailable special components). It is described the development of a new system, which is proposed in order to minimize difficulties with maintenance that utilizes in the assembling integrated circuits of large scale of integration. It is controlled by a microcomputer. (Author)

  10. Uranium Raw Material for the Nuclear Fuel Cycle: Exploration, Mining, Production, Supply and Demand, Economics and Environmental Issues (URAM-2009). Proceedings of an International Symposium

    International Nuclear Information System (INIS)

    This IAEA symposium is a third in a series which began in 2000 to discuss issues related to uranium raw materials. The symposia covered all areas of the uranium production cycle — including uranium geology, exploration, mining; milling and refining of uranium concentrates; and safety, environmental, social, training and regulatory issues — and reported on uranium supply and demand, and market scenarios. The first symposium was held in October 2000 — a time of extremely depressed market prices for uranium and of mines being closed — and primarily addressed environmental and safety issues in the uranium production cycle. By the time the second symposium was held in June 2005, the uranium market had started to improve after nearly two decades of depressed activity because of increased demand due to rising expectations for nuclear power expansion. Since then, there has been a dramatic rise in the uranium spot price, which in turn has promoted a significant increase in uranium exploration activities all over the world. The international symposium on Uranium Raw Material for the Nuclear Fuel Cycle (URAM-2009) was held at the IAEA, Vienna, 22–26 June 2009, at a time when nuclear energy was emerging as a viable alternative to meet the ever increasing demand of electricity in a sustainable manner, without degrading the environment. However, the global recession and credit crunch could impact the growth of the uranium industry. Since 2000, the identified uranium resource base has grown by more than 75%, exploration efforts have continued to increase in greenfield as well as brownfield sites, annual uranium production has risen, and the issue of social licensing and uranium stewardship has become increasingly important for public acceptance of the uranium industry. Some 210 delegates from 33 States and four international organizations participated in the symposium. In total, 120 technical papers were presented in the oral and poster sessions, and an exhibition on

  11. Barnwell Nuclear Fuels Plant applicability study. Volume III. Appendices

    International Nuclear Information System (INIS)

    Volume III suppliees supporting information to assist Congress in making a decision on the optimum utilization of the Barnwell Nuclear Fuels Plant. Included are applicable fuel cycle policies; properties of reference fuels; description and evaluation of alternative operational (flue cycle) modes; description and evaluation of safeguards systems and techniques; description and evaluation of spiking technology; waste and waste solidification evaluation; and Department of Energy programs relating to nonproliferation

  12. Cost and quality of fuels for electric plants 1993

    Energy Technology Data Exchange (ETDEWEB)

    1994-07-01

    The Cost and Quality of Fuels for Electric Utility Plants (C&Q) presents an annual summary of statistics at the national, Census division, State, electric utility, and plant levels regarding the quantity, quality, and cost of fossil fuels used to produce electricity. The purpose of this publication is to provide energy decision-makers with accurate and timely information that may be used in forming various perspectives on issues regarding electric power.

  13. Uranium resource assessments

    International Nuclear Information System (INIS)

    The objective of this investigation is to examine what is generally known about uranium resources, what is subject to conjecture, how well do the explorers themselves understand the occurrence of uranium, and who are the various participants in the exploration process. From this we hope to reach a better understanding of the quality of uranium resource estimates as well as the nature of the exploration process. The underlying questions will remain unanswered. But given an inability to estimate precisely our uranium resources, how much do we really need to know. To answer this latter question, the various Department of Energy needs for uranium resource estimates are examined. This allows consideration of whether or not given the absence of more complete long-term supply data and the associated problems of uranium deliverability for the electric utility industry, we are now threatened with nuclear power plants eventually standing idle due to an unanticipated lack of fuel for their reactors. Obviously this is of some consequence to the government and energy consuming public. The report is organized into four parts. Section I evaluates the uranium resource data base and the various methodologies of resource assessment. Part II describes the manner in which a private company goes about exploring for uranium and the nature of its internal need for resource information. Part III examines the structure of the industry for the purpose of determining the character of the industry with respect to resource development. Part IV arrives at conclusions about the emerging pattern of industrial behavior with respect to uranium supply and the implications this has for coping with national energy issues

  14. Uranium industry annual 1996

    International Nuclear Information System (INIS)

    The Uranium Industry Annual 1996 (UIA 1996) provides current statistical data on the US uranium industry's activities relating to uranium raw materials and uranium marketing. The UIA 1996 is prepared for use by the Congress, Federal and State agencies, the uranium and nuclear electric utility industries, and the public. Data on uranium raw materials activities for 1987 through 1996 including exploration activities and expenditures, EIA-estimated reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities for 1994 through 2006, including purchases of uranium and enrichment services, enrichment feed deliveries, uranium fuel assemblies, filled and unfilled market requirements, uranium imports and exports, and uranium inventories are shown in Chapter 2. A feature article, The Role of Thorium in Nuclear Energy, is included. 24 figs., 56 tabs

  15. Uranium industry annual 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-04-01

    The Uranium Industry Annual 1996 (UIA 1996) provides current statistical data on the US uranium industry`s activities relating to uranium raw materials and uranium marketing. The UIA 1996 is prepared for use by the Congress, Federal and State agencies, the uranium and nuclear electric utility industries, and the public. Data on uranium raw materials activities for 1987 through 1996 including exploration activities and expenditures, EIA-estimated reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities for 1994 through 2006, including purchases of uranium and enrichment services, enrichment feed deliveries, uranium fuel assemblies, filled and unfilled market requirements, uranium imports and exports, and uranium inventories are shown in Chapter 2. A feature article, The Role of Thorium in Nuclear Energy, is included. 24 figs., 56 tabs.

  16. The effects of fuel microstructure evolution on thermal performance in fast reactor type uranium-plutonium oxide fuel pins up to high burnup

    International Nuclear Information System (INIS)

    Higher power density and larger thermal gradients in fast reactor type uranium-plutonium oxide fuel pins than in water reactor type fuel rods induces the drastic fuel microstructure evolutions, which are useful for estimating the thermal condition at power in the fuel pellets. Recently, fuel-to-cladding gap reopen phenomena is paid attention to due to the influences on the heat transfer behaviors. The objectives of this work are to clarify the dominant factors for fuel microstructure evolution and investigate the gap reopen phenomena on the fuel pin thermal performance up to high burnup. The results of irradiation tests accumulated in JOYO Mk-II core, FFTF, Phenix were exhaustively collected and compiled to the database in this work. Then, the influences of irradiation conditions on the extent of microstructure evolution, such as central void, columnar grain, gas bubble region, densified region, dark ring, and residual gap (including gap reopen phenomena), were reviewed and investigated. Also the thermal behaviors in fuel pins were modeled to apply for the thermal analysis based on the recent knowledge referred to water reactor fuel experiences. It was revealed that the gap reopen phenomena related directly with the fission gas release behaviors in JOYO Mk-II J2 type driver fuel pins whose fabrication, irradiation, and post-irradiation conditions were well characterized. The results of one dimensional thermal analysis at the transverse sections were proved that gap conductance were improved by the condensed material which fills the gap even after the gap reopen phenomena. (author)

  17. 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

  18. Economical benefits for the use of slightly enriched fuel elements at the Atucha-I nuclear power plant

    International Nuclear Information System (INIS)

    The fuel represents a very important factor in the operative cost of the Atucha I nuclear power plant. This cost is drastically reduced with the use of fuel elements of slightly enriched uranium. The annual saving is analyzed with actual values for fuel elements with an enrichment of 0.85% by weight of U-235. With the reactor core in equilibrium state the annual saving achieved is approximately 7.5-10 u$s. According to the present irradiation plan, the benefit for the transition period is studied. An analysis of the sensitivity to differential increments in factors determining the cost of fuel elements or to changes in manufacturing losses is also performed, calculating its effect on the waste, the storage of irradiated elements and the amount of UO2 required. (Author)

  19. LIBS Spectral Data for a Mixed Actinide Fuel Pellet Containing Uranium, Plutonium, Neptunium and Americium

    Energy Technology Data Exchange (ETDEWEB)

    Judge, Elizabeth J. [Los Alamos National Laboratory; Berg, John M. [Los Alamos National Laboratory; Le, Loan A. [Los Alamos National Laboratory; Lopez, Leon N. [Los Alamos National Laboratory; Barefield, James E. [Los Alamos National Laboratory

    2012-06-18

    Laser-induced breakdown spectroscopy (LIBS) was used to analyze a mixed actinide fuel pellet containing 75% UO{sub 2}/20% PuO{sub 2}/3% AmO{sub 2}/2% NpO{sub 2}. The preliminary data shown here is the first report of LIBS analysis of a mixed actinide fuel pellet, to the authors knowledge. The LIBS spectral data was acquired in a plutonium facility at Los Alamos National Laboratory where the sample was contained within a glove box. The initial installation of the glove box was not intended for complete ultraviolet (UV), visible (VIS) and near infrared (NIR) transmission, therefore the LIBS spectrum is truncated in the UV and NIR regions due to the optical transmission of the window port and filters that were installed. The optical collection of the emission from the LIBS plasma will be optimized in the future. However, the preliminary LIBS data acquired is worth reporting due to the uniqueness of the sample and spectral data. The analysis of several actinides in the presence of each other is an important feature of this analysis since traditional methods must chemically separate uranium, plutonium, neptunium, and americium prior to analysis. Due to the historic nature of the sample fuel pellet analyzed, the provided sample composition of 75% UO{sub 2}/20% PuO{sub 2}/3% AmO{sub 2}/2% NpO{sub 2} cannot be confirm without further analytical processing. Uranium, plutonium, and americium emission lines were abundant and easily assigned while neptunium was more difficult to identify. There may be several reasons for this observation, other than knowing the exact sample composition of the fuel pellet. First, the atomic emission wavelength resources for neptunium are limited and such techniques as hollow cathode discharge lamp have different dynamics than the plasma used in LIBS which results in different emission spectra. Secondly, due to the complex sample of four actinide elements, which all have very dense electronic energy levels, there may be reactions and

  20. Midwest Fuel Recovery Plant. Environmental impact appraisal related to license amendment

    International Nuclear Information System (INIS)

    The environmental aspects of the proposed amendment to increase the spent fuel receipt and storage capacity at the Morris Operation facility from 100 metric tons of uranium to 750 metric tons of uranium by redesigning storage baskets and racks are discussed

  1. Draft environmental statement related to the Energy Fuels Nuclear, Inc., White Mesa Uranium Project (San Juan County, Utah)

    International Nuclear Information System (INIS)

    The proposed action is the issuance of a Source Material License to Energy Fuels Nuclear, Inc., for the construction and operation of the proposed White Mesa Uranium Project (Utah) with a product (U3O8) production limited to 7.3 x 105 kg (1.6 x 106 lb) per year. Possible environmental impacts and adverse effects were identified. Conditions for the protection of the environment are set forth before the license can be issued

  2. Fuel performance of licensed nuclear power plants through 1974

    International Nuclear Information System (INIS)

    General aspects of fuel element design and specific design data for typical Pressurized and Boiling Water Reactors are presented. Based on a literature search, failure modes and specific failures incurred through December 31, 1974 are described, together with a discussion of those problems which have had a significant impact upon plant operation. The relationship between fuel element failures and the resultant coolant activity/radioactive gaseous effluents upon radiation exposure, plant availability and capacity factors, economic impact, and waste management, are discussed. An assessment was made regarding the generation, availability, and use of fuel performance data

  3. Occurrence of Acidithiobacillus ferrooxidans and Acidithiobacillus thiooxidans in uranium mine-Caldas uranium mining and extraction plant, Brazil (CUMEP)

    International Nuclear Information System (INIS)

    The sulfated minerals present in mining areas may cause serious environmental problems due to the action of chemolithotrophic bacteria from genus Acithiobacillus, represented mainly by Acithiobacillus ferrooxidans and Acithiobacillus thiooxidans. These microorganisms are able to oxidize mineral sulfates, elementary sulfur and ferrous ion (A. ferrooxidans), as well are capable of mobilizing radionuclide as uranium to the environment. In this context, this study aimed at investigating the occurrence and the fluctuation of A. ferrooxidans and A. thiooxidans populations within the mine effluents, tailing dam and waste rocks of the Caldas Uranium Mining arid Extraction Plant (CUMEP) in Minas Gerais State - Brazil. Samples from 16 sites were evenly taken monthly in the CUMEP, during 28 months. The oxi-reduction potential, pH and temperature values were determined at the Radioecology Laboratory. The Most Probable Number technique was applied using a series of five tubes for selective counting of A. ferrooxidans and A. thiooxidans. Each sample was submitted to serial dilutions using Tween 80 and sterilized water (pH=2.0) and subsequently transferred into assay tubes containing T and K with ferrous ion and also elementary sulfur, as energy source, for detection of A. ferrooxidans and A. thiooxidans, respectively. Populations of A. ferrooxidans and A. thiooxidans presented seasonal quantitative fluctuations at the different studied sites. A. ferrooxidans showed higher or equal frequency to that observed for A. thiooxidans; as consequence, they were considered the predominant bacteria in this environment. In the majority of the sites, the highest values for the frequency and counting of A. ferrooxidans and A. thiooxidans were observed during the rainy period (October to March). The relative seasonal behavior when several variables are evaluated simultaneously indicated that, due to the high values of oxi-reduction potential, the low values of pH, the detection of the highest

  4. The French commercial plant for reprocessing and vitrification of spent reactor fuel

    International Nuclear Information System (INIS)

    The basic principles of the system for the reprocess ing of spent reactor fuel on a commercial basis at the plant at Marcoule in France are described. After four months cooling the canning is stripped mechanically and the fuel dissolved in boiling nitric acid. Using T B P the uranium and plutonium are separated from the fission products. The plutonium is then separated by valency change and aqueous extraction. The uranium is transferred to a refinery and stored as hexahydrate. The plutonium is concentrated and precipitated as oxalate, dried and converted via oxide to fluoride, thence to metal. The fission products are concentrated and converted to glass form in a process dependent on the reactor type, the burnup and their chemical processing. The two processes developed, the'pot' process and the continuous process, are described in some detail. The final disposal of the radioactive wastes is not yet decided, and temporary storage facilities have therefore been built, with a capacity for 30 years of reprocessing. (JIW)

  5. Uranium 2009 resources, production and demand

    CERN Document Server

    Organisation for Economic Cooperation and Development. Paris

    2010-01-01

    With several countries currently building nuclear power plants and planning the construction of more to meet long-term increases in electricity demand, uranium resources, production and demand remain topics of notable interest. In response to the projected growth in demand for uranium and declining inventories, the uranium industry – the first critical link in the fuel supply chain for nuclear reactors – is boosting production and developing plans for further increases in the near future. Strong market conditions will, however, be necessary to trigger the investments required to meet projected demand. The "Red Book", jointly prepared by the OECD Nuclear Energy Agency and the International Atomic Energy Agency, is a recognised world reference on uranium. It is based on information compiled in 40 countries, including those that are major producers and consumers of uranium. This 23rd edition provides a comprehensive review of world uranium supply and demand as of 1 January 2009, as well as data on global ur...

  6. The market for utility-scale fuel cell plants

    Science.gov (United States)

    Watanabe, Yasuo; Matsumoto, Masaru; Takasu, Kazuhiko

    This paper is devoted to a survey of the current technology and future market for utility-scale fuel cell plants. The phosphoric acid fuel cell (PAFC) is entering into the stage where it is practically available for use with natural gas. Large capacity plants such as 11, 5 and 1 MW have been installed and operated in Italy and Japan. Their efficiency ranges from 36 to 42%. The molten carbonate fuel cell (MCFC) is in the demonstrating stage, both the fuel cell and the balance-of-plant (BOP) for natural gas. Demonstration plants of 2 and 1 MW have been under construction in the USA and Japan. Their efficiency will range from 40 to 50%. The solid oxide fuel cell (SOFC) is in the experimental stage around 100 kW for co-generation. Its conceptual system design has been conducted for both centralized and dispersed power plant in a cooperation with Westinghouse and NEDO. A market survey is now considered on the basis that future fuel cells will run for around 40 000 h in a stable manner with competitive performance. The market for fuel cells will be roughly at 2000 MW in Japan by the year 2010. Half of them will be installed for electric companies on the utility scale. The market will be shared between PAFC and MCFC by 10 and 90%, respectively. Current technologies have not reached the stage to precisely forecast when fuel cells will be entering into the market on a utility scale. At the present time, it is worthwhile to consider how the technological and economic requirements will be definitely achieved. After achieving these requirements, fuel cells will be positively introduced and socially accepted as the best energy converting option to save energy and environmental impact. Further efforts will be devoted to meeting the market from the technological and economic aspects.

  7. Preconceptual design studies and cost data of depleted uranium hexafluoride conversion plants

    International Nuclear Information System (INIS)

    One of the more important legacies left with the Department of Energy (DOE) after the privatization of the United States Enrichment Corporation is the large inventory of depleted uranium hexafluoride (DUF6). The DOE Office of Nuclear Energy, Science and Technology (NE) is responsible for the long-term management of some 700,000 metric tons of DUF6 stored at the sites of the two gaseous diffusion plants located at Paducah, Kentucky and Portsmouth, Ohio, and at the East Tennessee Technology Park in Oak Ridge, Tennessee. The DUF6 management program resides in NE's Office of Depleted Uranium Hexafluoride Management. The current DUF6 program has largely focused on the ongoing maintenance of the cylinders containing DUF6. However, the long-term management and eventual disposition of DUF6 is the subject of a Programmatic Environmental Impact Statement (PEIS) and Public Law 105-204. The first step for future use or disposition is to convert the material, which requires construction and long-term operation of one or more conversion plants. To help inform the DUF6 program's planning activities, it was necessary to perform design and cost studies of likely DUF6 conversion plants at the preconceptual level, beyond the PEIS considerations but not as detailed as required for conceptual designs of actual plants. This report contains the final results from such a preconceptual design study project. In this fast track, three month effort, Lawrence Livermore National Laboratory and Bechtel National Incorporated developed and evaluated seven different preconceptual design cases for a single plant. The preconceptual design, schedules, costs, and issues associated with specific DUF6 conversion approaches, operating periods, and ownership options were evaluated based on criteria established by DOE. The single-plant conversion options studied were similar to the dry-conversion process alternatives from the PEIS. For each of the seven cases considered, this report contains information on

  8. Technico-Economical study of retreated uranium reenrichment

    International Nuclear Information System (INIS)

    Spent fuel from nuclear power plants is reprocessed at La Hague reprocessing plant in France. Recovered and stored reprocessed uranium has an energy potential unutilized so far. A modelisation is proposed in this paper for evaluating the economic interest reprocessed uranium reenrichment for using it again in a power plant. After briefly recalling the fuel cycle in light water reactors and reprocessed uranium specificities, a mathematical model for multi-isotope enrichment gives a differential system governing isotopic separation. Different solutions are proposed and compared. A. de la Garza analytical model's is retained. An economic value is attributed to reprocessed uranium. Results are presented as curves for determining the sensitivity of this value to simulation parameters (natural uranium cost, enrichment required by the electricity board etc.)

  9. Impact of recovered uranium cycle on the natural uranium production cycle and the environment

    International Nuclear Information System (INIS)

    The requirements by which future reactor and fuel cycle concepts must be judged are following: - properly utilize natural resources and national capabilities; - maximize the economic benefits; - effectively demonstrate the safety of fuel cycle facilities, and gain government and public approval for the enterprise; - satisfy national and international policies and goals; - contribute to sustainable energy supply. The ability to combine these five requirements ensures the success of the best options. Fuel utilization in thermal reactors can be improved in three ways: - lower the tails assay in the depleted stream of enrichment plants; - utilization of higher burnup fuel; - recycle plutonium. Recovered Uranium (RU) Cycle is a way to improve Slightly Enriched Uranium resulted from LWR spent fuel reprocessing which has 0.9-1.2% 235 U (dependent of the fuel history: reprocessing, burn up, reactor type) comparatively with 0.72% 235 U in natural Uranium. An international collaboration between Korea Atomic Energy Research Institute (KAERI), Atomic Energy of Canada Limited (AECL) and British Nuclear Fuel plc (BNFL) to use RU was developed. Since 1991, KAERI and AECL have introduced the Canadian Flexible (CANFLEX) fuel concept. A very attractive alternative to use RU in CANDU Reactors appears. Theoretically the quantity of 25,000 t (Europe and Japan) of RU would provide sufficient fuel for 500 CANDU reactor years of operation, knowing that the annual refueling requirement for a RU fuel burnup 13 MWd/KgU is around 50 t/y in comparison with 85 t/y for Natural Uranium (NU). Hereby, it is not necessary to mine about 42,500 t grade NU. In conclusion Recovered Uranium fuel cycle can be a very good option for the future of nuclear power in Romania. Moreover, waste resulted from uranium mining, waste resulted from uranium grade obtaining will disappear and financial costs, zones with nuclear activities and population exposed to irradiation will decrease. Also, the costs for fresh

  10. Correlation of radioactive waste treatment costs and the environmental impact of waste effluents in the nuclear fuel cycle for use in establishing ''as low as practicable'' guides: fabrication of light-water reactor fuel from enriched uranium dioxide

    International Nuclear Information System (INIS)

    A cost-benefit study was made to determine the cost and effectiveness of radioactive waste (radwaste) treatment systems for decreasing the release of radioactive materials from a model enriched-uranium, light-water reactor (LWR) fuel fabrication plant, and to determine the radiological impact (dose commitment) of the released materials on the environment. The study is designed to assist in defining the term ''as low as practicable'' in relation to limiting the release of radioactive materials from nuclear facilities. The base case model plant is representative of current plant technology and has an annual capacity of 1500 metric tons of LWR fuel. Additional radwaste treatment equipment is added to the base case plants in a series of case studies to decrease the amounts of radioactive materials released and to reduce the radiological dose commitment to the population in the surrounding area. The cost for the added waste treatment operations and the corresponding dose commitment are calculated for each case. In the final analysis, radiological dose is plotted vs the annual cost for treatment of the radwastes. The status of the radwaste treatment methods used in the case studies is discussed. Some of the technology used in the advanced cases is in an early stage of development and is not suitable for immediate use. The methodology used in estimating the costs and the radiological doses, detailed calculations, and tabulations are presented in Appendix A and ORNL-4992. (U.S.)

  11. Environmental releases from fuel cycle facility: part 1: radionuclide resuspension vs. stack releases on ambient airborne uranium and thorium levels

    International Nuclear Information System (INIS)

    Airborne activity levels of uranium and thorium series were measured in the vicinity (1.1 km) of a uranium (UF4) processing plant, located in Malvési, south of France. Regarding its impact on the environment, this facility is characterized by its routine atmospheric releases of uranium and by the emission of radionuclide-labelled particles from a storage pond filled with waste water or that contain dried sludge characterized by traces of plutonium and thorium (230Th). This study was performed during a whole year (November 2009–November 2010) and based on weekly aerosol sampling. Thanks to ICP-MS results, it was possible to perform investigations of uranium and thorium decay product concentration in the air. The number of aerosol filters sampled (50) was sufficient to establish a relationship between airborne radionuclide variations and the wind conditions. As expected, the more the time spent in the plume, the higher the ambient levels. The respective contributions of atmospheric releases and resuspension from local soil and waste ponds on ambient dust load and uranium-bearing aerosols were estimated. Two shutdown periods dedicated to facility servicing made it possible to estimate the resuspension contribution and to specify its origin (local or regional) according to the wind direction and remote background concentration. Airborne uranium mainly comes from the emission stack and, to a minor extent (∼20%), from wind resuspension of soil particles from the surrounding fields and areas devoted to waste storage. Moreover, weighed activity levels were clearly higher during operational periods than for shutdown periods. - Highlights: • Airborne activity levels of uranium and thorium were determined at about 1 km from the smokestack of a UF4 facility. • During shutdown periods all the airborne activity can be ascribed to local resuspension of formerly deposited radionuclides. • During production periods, the respective contribution of resuspension

  12. Off gas processing device for degreasing furnace for uranium/plutonium mixed oxide fuel

    International Nuclear Information System (INIS)

    A low melting ingredient capturing-cooling trap connected to a degreasing sintering furnace by way of sealed pipelines, a burning/decomposing device for decomposing high melting ingredient gases discharged from the cooling trap by burning them and a gas sucking means for forming the flow of off gases are contained in a glovebox, the inside pressure of which is kept negative. Since the degreasing sintering furnace for uranium/plutonium mixed oxide fuels is disposed outside of the glovebox, operation can be performed safely without greatly increasing the scale of the device, and the back flow of gases is prevented easily by keeping the pressure in the inside of the glovebox negative. Further, a heater is disposed at the midway of the sealed pipelines from the degreasing sintering furnace to the cooling trap, the temperature is kept high to prevent deposition of low melting ingredients to prevent clogging of the sealed pipelines. Further, a portion of the pipelines is made extensible in the axial direction to eliminate thermal stresses caused by temperature change thereby enabling to extend the life of the sealed pipelines. (N.H.)

  13. The Comparison Of Silicon Analysis For The Uranium Silicide Fuel Using Spectrophotometrical And Gravimetrical Methods

    International Nuclear Information System (INIS)

    The analysis of silicon content in the uranium silicide fuel spectro-photometrical and gravimetrical method have been performed. The nitrous oxide-acetylene was used in the atomic absorption spectrophotometry (AAS) on the wave length of 251.6 nm, and the mixture of ammonium hepta molybdate complexes and SnC12 as reductor were applied during analysis by UV-VIS spectrophotometry (UV-VIS) on the wave length of 757.5 mm. The reagent of HCLO4 and HNO3 were used for determining Si content by gravimetrical methods. The results of this comparison is as follows: the accuracy result is around 96.37 % + 0.24 % for the Si concentration up to 300 ppm (the AAS), is 138.60 % = 0.43 % for the Si concentration range between 0.1-1.5 ppm (UV-VIS), and is 51.13 % + 0.8 % for 1 gram of Si (gravimetry). The results also show that the lowest analytical error is obtained by AAS method

  14. The concept of the use of recycled uranium for increasing the degree of security of export deliveries of fuel for light-water reactors

    International Nuclear Information System (INIS)

    The present paper deals with investigation of the possibilities for reducing the risk of proliferation of fissionable materials by means of increasing the degree of protection of fresh fuel intended for light-water reactors against unsanctioned use in the case of withdrawal of a recipient country of deliveries from IAEA safeguards. It is shown that the use of recycled uranium for manufacturing export nuclear fuel makes transfer of nuclear material removed from the fuel assemblies for weapons purposes difficult because of the presence of isotope 232U, whose content increases when one attempts to enrich uranium extracted from fresh fuel. In combination with restricted access to technologies for isotope separation by means of establishing international centers for uranium enrichment, this technical measure can significantly reduce the risk of proliferation associated with export deliveries of fuel made of low-enriched uranium. The assessment of a maximum level of contamination of nuclear material being transferred by isotope 232U for the given isotope composition of the initial fuel is obtained. The concept of further investigations of the degree of security of export deliveries of fuel assemblies with recycled uranium intended for light-water reactors is suggested.

  15. A Consistent Comparative Study of Advanced Sodium-cooled Fast Burner Cores loaded with Thorium and Uranium-based Metallic Fuels

    International Nuclear Information System (INIS)

    We considered uranium-based metallic fuel of TRU-U-10Zr for driver fuel and thorium was considered as blanket because thorium blanket produces less amount of TRU than uranium blanket and use of thorium blanket leads to smaller sodium void worth than the use of uranium blanket due to the fact that the η-value increases much less with energy for 233U than for 239Pu and 232Th is less fissile than 238U. However, these cores using thorium blanket still have a large amount of TRU production from the driver fuels because the driver fuels contain a large amount of depleted uranium which leads to the production of TRU through neutron capture. The objective of this work is to consistently compare the neutronic performances of advanced sodium cooled fast reactor cores loaded with thorium and uraniumbased metallic fuels as driver fuel for TRU burning. Our main emphasis is given on the analyses of the differences in the core performance parameters. For consistent comparison, we used the same core configuration and all the same design parameters except for the fact that depleted uranium in uraniumbased fuel is replaced with thorium. We considered the cores having no thorium blanket and the cores having thorium blanket that were designed in our previous works

  16. Capital and operating costs of irradiated natural uranium reprocessing plants; Couts d'investissement et d'exploitation des usines de retraitement de l'uranium naturel irradie

    Energy Technology Data Exchange (ETDEWEB)

    Thiriet, L.; Jouannaud, C.; Couture, J.; Duboz, J. [Commissariat a l' Energie Atomique (France). Centre d' Etudes Nucleaires; Oger, C. [Saint Gobain Nucleaire (France)

    1966-07-01

    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) [French] La communication expose d'abord une methode d'analyse des couts d'investissement des usines de retraitement de l'uranium naturel irradie (inspiree de celles de LANG et de BACH, bien connues dans l'industrie petroliere) et de leurs couts d'exploitation (selon leur nature economique). Cette methode permet d'etablir des structures types de couts de ces usines et de comparer les realisations industrielles et les projets. Elle facilite l'exploration des consequences previsibles du progres technique. On indique un certain nombre de resultats obtenus, concernant la sensibilite des couts d'investissement de ces usines aux differents parametres techniques definissant le combustible et leur confrontation selon les pays ou aires economiques envisages. On montre enfin comment doit pouvoir s'exprimer l'influence de la taille des usines sur leur cout d'investissement. (auteur)

  17. Arid lands plants as feedstocks for fuels and chemicals

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmann, J.J.

    1983-01-01

    The purpose of this paper is to review the recent research on arid-adapted plants that have potential as producers of fuels or chemicals. The major focus will be on plant species that appear to have commercial value. Research on guayule (Parthenium argentatum) and jojoba (Simmondsia chinensis) will be mentioned only briefly, since these plants have been discussed extensively in the literature, and excellent reviews are already in existence. In this review the literature on arid-adapted plants that have potential uses for solid fuels, liquid fuels, and chemical feedstocks is summarized, followed by an overview of the research directions and types of development that are needed in order for bio-energy production systems to reach the commercial stage. 127 references.

  18. Preparation of High-Density Uranium-Silicide U3Sl2-Uss: Effects of Preirradiation Heat Treatment on As-Cast Ingot Fuel Plates

    International Nuclear Information System (INIS)

    Heat treatment experiments upon U3Si2-Uss ingot have been camed out to obtain free uranium particle size improvement which is required to enhance the U-Al inter-diffusion reaction in the fuel plate meat. . Heat treatment experiments upon fuel plates containing dispersion of U3Si2-Uss in Al matrix have also been carried out to study the effect of temperature and treatment duration on the extent of inter-diffusion reaction between free uranium particle and aluminium matrix in the fuel plate meat. Both the experiments indicate that a drastic size improvement has occurred with the U3Si2 as well as free uranium particles upon heat treatment at controlled temperature between the U3Si2 peritectic and peritectoid temperatures and that the inter-diffusion reaction between free uranium and Al matrix occurs quite significantly at temperatures higher than that ordinarily used in the fabrication procedure

  19. The environmental behaviour of uranium and thorium

    International Nuclear Information System (INIS)

    Uranium and thorium have had many uses in the past, and their present and potential use as nuclear fuels in energy production is very significant. Both elements, and their daughter products, are of environmental interest because they may have effects from the time of mining to the time of ultimate disposal of used nuclear fuel. To assess the impact on the environment of man's use and disposal of uranium and thorium, we must know the physical, chemical and biological behaviour of these elements. This report summarizes the literature, updating and extending earlier reviews pertaining to uranium and thorium. The radiological properties, chemistry, forms of occurrence in nature, soil interactions, as well as distribution coefficients and mode of transport are discussed for both elements. In addition, uranium and thorium concentrations in plants, plant transfer coefficients, concentrations in soil organisms and methods of detection are summarized. (auth)

  20. Application of neutron well coincidence counting for plutonium determination in mixed oxide fuel fabrication plant

    International Nuclear Information System (INIS)

    Mixed oxide (MOX) fuel is an alternative to conventional enriched uranium oxide fuel in thermal reactors. Indian interest in plutonium recycle in thermal reactors is primarily due to the need to develop alternative indigenous fuel for two boiling water reactors (BWR) at Tarapur, which are designed to use imported light enriched uranium fuel. A few MOX assemblies have been fabricated and loaded into the reactors. Neutron well coincidence counting (NWCC) system has been successfully employed to check the enrichments of PuO2 in MOX blends. NWCC has also been successfully applied in developing dry recycling process of clean rejected oxide (CRO) and dirty rejected oxide (DRO). (author)

  1. Mill tailings disposal and environmental monitoring at the Ningyo-Toge uranium processing pilot plant

    International Nuclear Information System (INIS)

    The tailings from the uranium processing pilot plant with a maximum ore processing capacity of 50 t/d are transferred to a tailings dam. The overflow from the dam is chemically treated and through settling ponds, sand filters to be discharged into a river. The concentrations of U, 226Ra, pH, S.S., COD, Fe, Mn, Cl and F were monitored periodically and they were all below the control values. The results of monitoring on the river bed and rice paddy soil showed no signs of accumulation of U and 226Ra in it

  2. The New Generation of Uranium In Situ Recovery Facilities: Design Improvements Should Reduce Radiological Impacts Relative to First Generation Uranium Solution Mining Plants

    International Nuclear Information System (INIS)

    In the last few years, there has been a significant increase in the demand for Uranium as historical inventories have been consumed and new reactor orders are being placed. Numerous mineralized properties around the world are being evaluated for Uranium recovery and new mining / milling projects are being evaluated and developed. Ore bodies which are considered uneconomical to mine by conventional methods such as tunneling or open pits, can be candidates for non-conventional recovery techniques, involving considerably less capital expenditure. Technologies such as Uranium In Situ Leaching / In Situ Recovery (ISL / ISR - also referred to as 'solution mining'), have enabled commercial scale mining and milling of relatively small ore pockets of lower grade, and are expected to make a significant contribution to overall world wide uranium supplies over the next ten years. Commercial size solution mining production facilities have operated in the US since the mid 1970's. However, current designs are expected to result in less radiological wastes and emissions relative to these 'first' generation plants (which were designed, constructed and operated through the 1980's). These early designs typically used alkaline leach chemistries in situ including use of ammonium carbonate which resulted in groundwater restoration challenges, open to air recovery vessels and high temperature calcining systems for final product drying vs the 'zero emissions' vacuum dryers as typically used today. Improved containment, automation and instrumentation control and use of vacuum dryers in the design of current generation plants are expected to reduce production of secondary waste byproduct material, reduce Radon emissions and reduce potential for employee exposure to uranium concentrate aerosols at the back end of the milling process. In Situ Recovery in the U.S. typically involves the circulation of groundwater, fortified with oxidizing (gaseous oxygen e.g) and complexing agents (carbon

  3. Practical Gamma Counting of Unirradiated Uranium-235

    International Nuclear Information System (INIS)

    During the fabrication of reactor fuel elements it is necessary to have assurance regarding the accuracy of the fabricator's uranium assignment within the specified tolerances. Destructive analysis of random samples is both expensive and time-consuming. Where the uranium-bearing components are suitable for gamma counting, a non-destructive method of assay can be used with greater efficiency and equal accuracy. The particular method described was used for checking fuel cores of nominal 30 wt.% enriched uranium in aluminium measuring about two inches square by 0.080 in. and 0.160 in. thick. The equipment was a basic Nal scintillation counter equipped with a single-channel analyser. The analyser, however, was operated with a very wide window covering both the 90-keV and 184-keV peaks characteristic of uranium-235. In practice, the threshold level acid the window opening, were adjusted to give the optimum maximum count rate as indicated by a ratemeter. The counting of a fuel core was then performed with the Nal crystal essentially unshielded and located several inches above the fuel core. The counting time was adjusted to yield a total count in the range of 105 to 106 in order to minimize the counting error. Effects due to variations in the counting geometry and to non-uniform uranium distribution were minimized by the relatively large separation of the crystal from the fuel core. Effects due to shifting of analyser window were minimized by use of a wide opening. To compensate for possible non-uniform uranium distribution through the thickness of a fuel core, each core was counted on both sides. The total count obtained in this manner was directly proportioned to the uranium-235 content of the fuel core. In application, the counting equipment was set up in the fabricator's plant and a number of production fuel cores were counted. The plotting of the total counts against the fabricator's uranium-235 assignment revealed an unexpected error in the fabricator's system

  4. Elements of radioprotection culture at Nuclear Fuel Plant at Pitesti, Romania

    International Nuclear Information System (INIS)

    Culture as an element of human existence is characterized primarily by accumulating, skills, attitudes, behavior and promotion to something. Nuclear security culture has developed a desire to prepare to cope with nuclear risk, born with the discovery of radioactivity. Culture of radiation protection is required for the implementation of safety and security in the protection against ionizing radiation. In the activity of producing nuclear fuel for CANDU based on natural and depleted uranium stakeholders are developing a specific radiation protection culture due the existing radiological risks that can lead to exposure to ionizing radiation. For Nuclear Fuel Plant Pitesti (FCN) where nuclear material is both as bulk and as itemized form, individual radiological monitoring as well as the working environment are established in Radiological Security Manual (MSR), in accordance with national legislation and international. All employees FCN are classified as A and B classes of exposure to ionizing radiation. Classification is based on work performed and the degree of exposure and specific training for radiological safety is sized for each category. External workers, contractors and visitors are aware of the entry in the plant on the risks of exposure to ionizing radiation and are properly trained. Basics of radiation protection culture as individual monitoring, communication, rigorous and prudent approach, pro-active action activity are transmitted continuously. In addition the radiation protection culture existed there from the beginning of the commissioning of the plant, but its elements still remain scattered in knowledge, attitudes and behaviors of employees. (authors)

  5. Production Cycle for Large Scale Fission Mo-99 Separation by the Processing of Irradiated LEU Uranium Silicide Fuel Element Targets

    Directory of Open Access Journals (Sweden)

    Abdel-Hadi Ali Sameh

    2013-01-01

    Full Text Available Uranium silicide fuels proved over decades their exceptional qualification for the operation of higher flux material testing reactors with LEU elements. The application of such fuels as target materials, particularly for the large scale fission Mo-99 producers, offers an efficient and economical solution for the related facilities. The realization of such aim demands the introduction of a suitable dissolution process for the applied U3Si2 compound. Excellent results are achieved by the oxidizing dissolution of the fuel meat in hydrofluoric acid at room temperature. The resulting solution is directly behind added to an over stoichiometric amount of potassium hydroxide solution. Uranium and the bulk of fission products are precipitated together with the transuranium compounds. The filtrate contains the molybdenum and the soluble fission product species. It is further treated similar to the in-full scale proven process. The generated off gas stream is handled also as experienced before after passing through KOH washing solution. The generated alkaline fluoride containing waste solution is noncorrosive. Nevertheless fluoride can be selectively bonded as in soluble CaF2 by addition of a mixture of solid calcium hydroxide calcium carbonate to the sand cement mixture used for waste solidification. The generated elevated amounts of LEU remnants can be recycled and retargeted. The related technology permits the minimization of the generated fuel waste, saving environment, and improving processing economy.

  6. A technique to measure fuel oil viscosity in a fuel power plant.

    Science.gov (United States)

    Delgadillo, Miguel Angel; Ibargüengoytia, Pablo H; García, Uriel A

    2016-01-01

    The viscosity measurement and control of fuel oil in power plants is very important for a proper combustion. However, the conventional viscometers are only reliable for a short period of time. This paper proposes an on-line analytic viscosity evaluation based on energy balance applied to a piece of tube entering the fuel oil main heater and a new control strategy for temperature control. This analytic evaluation utilizes a set of temperature versus viscosity graphs were defined during years of analysis of fuel oil in Mexican power plants. Also the temperature set-point for the fuel oil main heater output is obtained by interpolating in the corresponding graph. Validation tests of the proposed analytic equations were carried out in the Tuxpan power plant in Veracruz, Mexico.

  7. Accumulation of uranium by aquatic plants in field conditions: Prospects for phytoremediation

    Energy Technology Data Exchange (ETDEWEB)

    Favas, Paulo J.C., E-mail: pjcf@utad.pt [School of Life Sciences and the Environment, University of Trás-os-Montes e Alto Douro, 5001-801 Vila Real (Portugal); IMAR-CMA Marine and Environmental Research Centre, Faculty of Sciences and Technology, University of Coimbra, 3001-401 Coimbra (Portugal); Pratas, João [Department of Earth Sciences, Faculty of Sciences and Technology, University of Coimbra, 3001-401 Coimbra (Portugal); IMAR-CMA Marine and Environmental Research Centre, Faculty of Sciences and Technology, University of Coimbra, 3001-401 Coimbra (Portugal); Varun, Mayank; D' Souza, Rohan; Paul, Manoj S. [Department of Botany, St. John' s College, Agra 282 002 (India)

    2014-02-01

    A study was undertaken to determine Uranium concentrations in water and aquatic plants in the uraniferous region of Beiras, Central Portugal. Samples were collected from running water (n = 200) at places where aquatic species were observed. Plant samples were collected from 28 species of submerged, free-floating and rooted emergent plants including 2 bryophytes and 1 pteridophyte. Uranium concentrations in surface waters ranged from 0.23 to 1217 μg L{sup −1}. The aquatic plant species studied, including several previously untested species, exhibited the ability to accumulate U in concentrations many times that of the ambient water. In general submerged plants exhibited higher U content followed by rooted emergent and free floating species. The highest U concentrations were observed in the bryophyte Fontinalis antipyretica (up to 4979 mg kg{sup −1}) followed by Callitriche stagnalis (1963 mg kg{sup −1}), Callitriche hamulata (379 mg kg{sup −1}), Ranunculus peltatus subsp. saniculifolius (243 mg kg{sup −1}), Callitriche lusitanica (218 mg kg{sup −1}), and Ranunculus trichophyllus (65.8 mg kg{sup −1}). In two out of three rooted emergent species U seemed to be preferentially partitioned in rhizome/roots with highest rhizome U content recorded in Typha latifolia (380 mg kg{sup −1}). Among the free-floating species, the highest U content (42.5 mg kg{sup −1}) was seen in Lemna minor. The bryophyte F. antipyretica and Callitrichaceae members seem to be promising candidates for the development of phytofiltration methodologies based on U accumulation, abundance and biomass production. - Highlights: • Exploration of U contamination extent in uraniferous province of Central Portugal • A group of previously untested species with the ability to accumulate U was assessed • U accumulation patterns in the species indicate their potential in bioindication and phytoremediation of U-contaminated water.

  8. 226Ra bioavailability to plants at the Urgeiriça uranium mill tailings site.

    Science.gov (United States)

    Madruga, M J; Brogueira, A; Alberto, G; Cardoso, F

    2001-01-01

    Large amounts of solid wastes (tailings) resulting from the exploitation and treatment of uranium ore at the Urgeiriça 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 these plants, some shrubs (Cytisus spp.) are growing in some of the dams. The objective of this study is to determine the 226Ra bioavailability from uranium mill tailings by quantifying the total and available fraction of radium in the tailings and to estimate its transfer to plants growing on the tailing piles. Plant and tailing samples were randomly collected and the activity concentration of 226Ra in plants (aerial part and roots) and tailings was measured by gamma-spectrometry. The exchangeable fraction of radium in tailings 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 226Ra uptake by plants show that 226Ra concentration ratios for eucalyptus and pines decrease at low 226Ra concentrations in the tailings 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 tailings, 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, at a 95% confidence level, 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. PMID:11379070

  9. BIO-MONITORING FOR URANIUM USING STREAM-SIDE TERRESTRIAL PLANTS AND MACROPHYTES

    Energy Technology Data Exchange (ETDEWEB)

    Caldwell, E.; Duff, M.; Hicks, T.; Coughlin, D.; Hicks, R.; Dixon, E.

    2012-01-12

    This study evaluated the abilities of various plant species to act as bio-monitors for environmental uranium (U) contamination. Vegetation and soil samples were collected from a U processing facility. The water-way fed from facility storm and processing effluents was the focal sample site as it represented a primary U transport mechanism. Soils and sediments from areas exposed to contamination possessed U concentrations that averaged 630 mg U kg{sup -1}. Aquatic mosses proved to be exceptional accumulators of U with dry weight (dw) concentrations measuring as high as 12500 mg U kg{sup -1} (approximately 1% of the dw mass was attributable to U). The macrophytes (Phragmites communis, Scripus fontinalis and Sagittaria latifolia) were also effective accumulators of U. In general, plant roots possessed higher concentrations of U than associated upper portions of plants. For terrestrial plants, the roots of Impatiens capensis had the highest observed levels of U accumulation (1030 mg kg{sup -1}), followed by the roots of Cyperus esculentus and Solidago speciosa. The concentration ratio (CR) characterized dry weight (dw) vegetative U levels relative to that in associated dw soil. The plant species that accumulated U at levels in excess of that found in the soil were: P. communis root (CR, 17.4), I. capensis root (CR, 3.1) and S. fontinalis whole plant (CR, 1.4). Seven of the highest ten CR values were found in the roots. Correlations with concentrations of other metals with U were performed, which revealed that U concentrations in the plant were strongly correlated with nickel (Ni) concentrations (correlation: 0.992; r-squared: 0.984). Uranium in plant tissue was also strongly correlated with strontium (Sr) (correlation: 0.948; r-squared: 0.899). Strontium is chemically and physically similar to calcium (Ca) and magnesium (Mg), which were also positively-correlated with U. The correlation with U and these plant nutrient minerals, including iron (Fe), suggests that active

  10. Thermal performance of fast reactor type uranium-plutonium oxide fuel pins at beginning-life conditions

    International Nuclear Information System (INIS)

    A new thermal conductivity correlation, named as ''PNC'98 equation'', of fast reactor type uranium-plutonium oxide (FR-MOX) fuel was proposed by one of the authors last year. Since maximum operating temperature of fast reactor fuel is generally much higher than that of light water reactor fuel, the predictability of the correlations for FR-MOX fuel at higher temperatures, especially, above 2300K, become more important in case of the fuel pin thermal performance analysis. Although the thermal conductivity measurements at higher temperatures are very scares, power-to-melt (PTM) experiments at beginning-of-life conditions can compensate for the lack of the measurements and be preferable to verify the correlation up to melting point. The PNC'98 equation was applied to the PTM test results irradiated in experimental fast reactor JOYO in order to examine its predictability up to melting point by integral method from a viewpoint of the heat transfer behavior across fuel-to-cladding gap through the fuel pin thermal analysis. Fuel pin irradiation behaviors, such as microstructure evolution and the porosity effect on the heat conduction in fuel pellet, were carefully modeled with some related physical properties based on the latest knowledge in order to minimize the uncertainty of the analysis. The calculated heat transfer behavior across the gap derived from temperature increments in the fuel pin is reasonably consistent to the past studies. This implies that the combinations of the PNC'98 equation with the model are highly predictable for the temperature distribution in the PTM test fuel pins and that the applicability of the PNC'98 equation is verified up to melting point. (author)

  11. Riddle of depleted uranium

    International Nuclear Information System (INIS)

    Depleted Uranium (DU) is the waste product of uranium enrichment from the manufacturing of fuel rods for nuclear reactors in nuclear power plants and nuclear power ships. DU may also results from the reprocessing of spent nuclear reactor fuel. Potentially DU has both chemical and radiological toxicity with two important targets organs being the kidney and the lungs. DU is made into a metal and, due to its availability, low price, high specific weight, density and melting point as well as its pyrophoricity; it has a wide range of civilian and military applications. Due to the use of DU over the recent years, there appeared in some press on health hazards that are alleged to be due to DU. In these paper properties, applications, potential environmental and health effects of DU are briefly reviewed

  12. CO2 capture from oxy-fuel combustion power plants

    OpenAIRE

    Hu, Yukun

    2011-01-01

    To mitigate the global greenhouse gases (GHGs) emissions, carbon dioxide (CO2) capture and storage (CCS) has the potential to play a significant role for reaching mitigation target. Oxy-fuel combustion is a promising technology for CO2 capture in power plants. Advantages compared to CCS with the conventional combustion technology are: high combustion efficiency, flue gas volume reduction, low fuel consumption, near zero CO2 emission, and less nitrogen oxides (NOx) formation can be reached sim...

  13. Dimensional stability of low enriched uranium silicide plate-type fuel for research reactors at transient conditions

    International Nuclear Information System (INIS)

    This paper describes the result of transient experiments using low enriched uranium silicide plate-type fuel for research reactors. The pulse irradiation was carried out at Nuclear Safety Research Reactor (NSRR) in Japan Atomic Energy Research Institute. The results obtained were: (1) At fuel plate temperature of below 400degC, a good dimensional stability of the tested fuel was kept. No fuel failure occurred. (2) At a plate temperature of about 540degC, a local crack was initiated on the Al-3% Mg alloy cladding. Once the cladding temperature exceeded the melting point of 640degC, the fuel plate was degraded much by increased bowing and cracking of the denuded fuel meat occurred after relocation of molten Al cladding. Despite of these degradation, neither fragmentation of the fuel plate nor mechanical energy generation occurred up to the cladding temperature of 971degC. (3) At the temperatures of around 925degC, the reaction of silicide particles with molten Al in the matrix and that of cladding occurred, forming Al riched U (Al, Si) compounds and Si riched (U, Si) compounds at the outermost surface of the silicide particles. (author)

  14. Conversion and Evaluation of the University of Massachusetts Lowell Research Reactor From High-Enriched To Low-Enriched Uranium Fuel

    International Nuclear Information System (INIS)

    The process for converting the University of Massachusetts Lowell Research Reactor (UMLRR) from high-enrichment uranium (HEU) fuel to low-enrichment uranium (LEU) fuel began in 1988. Several years of design reviews, computational modeling, and thermal hydraulic analyses resulted in a preliminary reference core design and configuration based on 20 standard, MTR-type, flat-plate, 19.75% enriched, uranium silicide (u3Si2) fuel elements. A final safety analysis for the fuel conversion was submitted to the Nuclear Regulatory Commission (NRC) in 1993. The NRC made two additional requests for additional information and supplements were submitted in 1994 and 1997. The new UMLRR Reactor Supervisor initiated an effort to change the LEU reference core configuration to eliminate a complicated control rod modification needed for the smaller core

  15. DOE small scale fuel alcohol plant design

    Energy Technology Data Exchange (ETDEWEB)

    LaRue, D.M.; Richardson, J.G.

    1980-01-01

    The Department of Energy, in an effort to facilitate the deployment of rural-based ethanol production capability, has undertaken this effort to develop a basic small-scale plant design capable of producing anhydrous ethanol. The design, when completed, will contain all necessary specifications and diagrams sufficient for the construction of a plant. The design concept is modular; that is, sections of the plant can stand alone or be integrated into other designs with comparable throughput rates. The plant design will be easily scaled up or down from the designed flow rate of 25 gallons of ethanol per hour. Conversion factors will be provided with the final design package to explain scale-up and scale-down procedures. The intent of this program is to provide potential small-scale producers with sound information about the size, engineering requirements, costs and level of effort in building such a system.

  16. Environmental releases from fuel cycle facility: part 1: radionuclide resuspension vs. stack releases on ambient airborne uranium and thorium levels.

    Science.gov (United States)

    Masson, Olivier; Pourcelot, Laurent; Boulet, Béatrice; Cagnat, Xavier; Videau, Gérard

    2015-03-01

    Airborne activity levels of uranium and thorium series were measured in the vicinity (1.1 km) of a uranium (UF4) processing plant, located in Malvési, south of France. Regarding its impact on the environment, this facility is characterized by its routine atmospheric releases of uranium and by the emission of radionuclide-labelled particles from a storage pond filled with waste water or that contain dried sludge characterized by traces of plutonium and thorium ((230)Th). This study was performed during a whole year (November 2009-November 2010) and based on weekly aerosol sampling. Thanks to ICP-MS results, it was possible to perform investigations of uranium and thorium decay product concentration in the air. The number of aerosol filters sampled (50) was sufficient to establish a relationship between airborne radionuclide variations and the wind conditions. As expected, the more the time spent in the plume, the higher the ambient levels. The respective contributions of atmospheric releases and resuspension from local soil and waste ponds on ambient dust load and uranium-bearing aerosols were estimated. Two shutdown periods dedicated to facility servicing made it possible to estimate the resuspension contribution and to specify its origin (local or regional) according to the wind direction and remote background concentration. Airborne uranium mainly comes from the emission stack and, to a minor extent (∼20%), from wind resuspension of soil particles from the surrounding fields and areas devoted to waste storage. Moreover, weighed activity levels were clearly higher during operational periods than for shutdown periods.

  17. Environmental releases from fuel cycle facility: part 1: radionuclide resuspension vs. stack releases on ambient airborne uranium and thorium levels.

    Science.gov (United States)

    Masson, Olivier; Pourcelot, Laurent; Boulet, Béatrice; Cagnat, Xavier; Videau, Gérard

    2015-03-01

    Airborne activity levels of uranium and thorium series were measured in the vicinity (1.1 km) of a uranium (UF4) processing plant, located in Malvési, south of France. Regarding its impact on the environment, this facility is characterized by its routine atmospheric releases of uranium and by the emission of radionuclide-labelled particles from a storage pond filled with waste water or that contain dried sludge characterized by traces of plutonium and thorium ((230)Th). This study was performed during a whole year (November 2009-November 2010) and based on weekly aerosol sampling. Thanks to ICP-MS results, it was possible to perform investigations of uranium and thorium decay product concentration in the air. The number of aerosol filters sampled (50) was sufficient to establish a relationship between airborne radionuclide variations and the wind conditions. As expected, the more the time spent in the plume, the higher the ambient levels. The respective contributions of atmospheric releases and resuspension from local soil and waste ponds on ambient dust load and uranium-bearing aerosols were estimated. Two shutdown periods dedicated to facility servicing made it possible to estimate the resuspension contribution and to specify its origin (local or regional) according to the wind direction and remote background concentration. Airborne uranium mainly comes from the emission stack and, to a minor extent (∼20%), from wind resuspension of soil particles from the surrounding fields and areas devoted to waste storage. Moreover, weighed activity levels were clearly higher during operational periods than for shutdown periods. PMID:25613358

  18. Uptake of uranium, thorium and radium isotopes by plants growing in dam impoundment Tasotkel and the Lower Shu region (Kazakhstan)

    Energy Technology Data Exchange (ETDEWEB)

    Matveyeva, Ilona; Burkitbayev, Mukhambetkali [al-Farabi Kazakh National University, Almaty (Kazakhstan). Faculty of Chemistry and Chemical Technology; Jacimovic, Radojko [Jozef Stefan Institute, Ljubljana (Slovenia). Dept. of Environmental Sciences; Planinsek, Petra; Smodis, Borut [Jozef Stefan Institute, Ljubljana (Slovenia). Dept. of Environmental Sciences; Jozef Stefan International Postgraduate School, Ljubljana (Slovenia)

    2016-04-01

    The activity concentrations of isotopes of uranium, thorium and radium-226 in dominant species of plants (Xantium strumarium, Phragmites communis, Artemisia nitrosa and Artemisia serotina) growing on the territories contaminated by uranium industry of Kazakhstan (close to dam impoundment Tasotkel and the Lower Shu region) are presented. The obtained data showed the significant variations of activity concentrations of isotopes of uranium, thorium and radium-226 in above ground parts. The concentrations of most of the investigated radionuclides in the root system are higher than in the aboveground parts; it can be explained by root barrier. It was found that the highest root barrier has Xantium strumarium, especially for uranium isotopes. The concentration ratios of radionuclides were calculated, and as the result it was found that the highest accumulation ability in the investigated region has Artemisia serotina.

  19. Uptake of uranium, thorium and radium isotopes by plants growing in dam impoundment Tasotkel and the Lower Shu region (Kazakhstan)

    International Nuclear Information System (INIS)

    The activity concentrations of isotopes of uranium, thorium and radium-226 in dominant species of plants (Xantium strumarium, Phragmites communis, Artemisia nitrosa and Artemisia serotina) growing on the territories contaminated by uranium industry of Kazakhstan (close to dam impoundment Tasotkel and the Lower Shu region) are presented. The obtained data showed the significant variations of activity concentrations of isotopes of uranium, thorium and radium-226 in above ground parts. The concentrations of most of the investigated radionuclides in the root system are higher than in the aboveground parts; it can be explained by root barrier. It was found that the highest root barrier has Xantium strumarium, especially for uranium isotopes. The concentration ratios of radionuclides were calculated, and as the result it was found that the highest accumulation ability in the investigated region has Artemisia serotina.

  20. ND online software development for data acquisition of replacement operations of fuel assemblies of Atucha I Nuclear power plant

    International Nuclear Information System (INIS)

    The ND Online software was developed in order to acquire data on a real-time basis of the refueling operations at the Atucha I nuclear power plant. The fuel elements containing slightly enriched uranium dioxide are located in the nuclear reactor core inside the cooling channels. The refueling operations are made periodically while the reactor is operating at full power. The acquired signals during the refueling operations are: pressure, force and position of the fuel element. In order to improve the safety and availability of the installation, monitoring of the refueling operations is important for the early detection of anomalies related to the fuel element itself, the cooling channels or the refueling machine (author)

  1. HIGH ENERGY LIQUID FUELS FROM PLANTS

    Energy Technology Data Exchange (ETDEWEB)

    Nemethy, E. K.; Otvos, J. W.; Calvin, M.

    1980-10-01

    The heptane extract of Euphorbia lathyris has a low oxygen content and a heat valve of 42 MJ/kg which is comparable to that of crude oil (44 MJ/kg). These qualities indicate a potential for use as fuel or chemical feedstock material. Therefore we have investigated the chemical composition of this fraction in some detail. Since the amoun of the methanol fraction is quite substantial we have also identified the major components of this fraction.

  2. Functional design criteria for the 242-A evaporator and PUREX [Plutonium-Uranium Extraction] Plant condensate interim retention basin

    International Nuclear Information System (INIS)

    This document contains the functional design criteria for a 26- million-gallon retention basin and 10 million gallons of temporary storage tanks. The basin and tanks will be used to store 242-A Evaporator process condensate, the Plutonium-Uranium Extraction (PUREX) Plant process distillate discharge stream, and the PUREX Plant ammonia scrubber distillate stream. Completion of the project will allow both the 242-A Evaporator and the PUREX Plant to restart. 4 refs

  3. Uranium Determination in Samples from Decommissioning of Nuclear facilities Related to the First Stage of Nuclear Fuel Cycle

    International Nuclear Information System (INIS)

    An adequate workplace monitoring must be carried out during the decommissioning activities, to ensure the protection of workers involved in these tasks. In addition, a large amount of waste materials are generated during the decommissioning of nuclear facilities. Clearance levels are established by regulatory authorities and are normally quite low. The determination of those activity concentration levels become more difficult when it is necessary to quantify alpha emitters such as uranium, especially when complex matrices are involved. Several methods for uranium determination in samples obtained during the decommissioning of a facility related to the first stage of the nuclear fuel cycle are presented in this work. Measurements were carried out by laboratory techniques. In situ gamma spectrometry was also used to perform measurements on site. A comparison among the different techniques was also done by analysing the results obtained in some practical applications. (Author)

  4. The spectrographic analysis of plutonium oxide or mixed plutonium oxide/uranium oxide fuel pellets by the dried residue technique

    International Nuclear Information System (INIS)

    An emission spectrographic method for the quantitative determination of metallic impurities in plutonium oxide and mixed plutonium oxide/uranium oxide is described. The fuel is dissolved in nitric acid and the plutonium and/or uranium extracted with tributyl phosphate. A small aliquot of the aqueous residue is dried on a 'mini' pyrolitic graphite plate and excited by high voltage AC spark in an oxygen atmosphere. Spectra are recorded in a region which has been specially selected to record simultaneously lines of boron and cadmium in the 2nd order and all the other elements of interest in the 1st order. Indium is used as an internal standard. The excitation of very small quantities of the uraniumm/plutonium free residue by high voltage spark, together with three separate levels of containment reduce the hazards to personnel and the environment to a minimum with limited effect on sensitivity and accuracy of the results. (auth)

  5. Idaho Chemical Processing Plant Spent Fuel and Waste Management Technology Development Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1993-09-01

    The Department of Energy (DOE) has received spent nuclear fuel (SNF) at the Idaho Chemical Processing Plant (ICPP) for interim storage and reprocessing since 1953. Reprocessing of SNF has resulted in an existing inventory of 1.5 million gallons of radioactive sodium-bearing liquid waste and 3800 cubic meters (m{sup 3}) of calcine, in addition to the 768 metric tons (MT) of SNF and various other fuel materials in inventory. To date, the major activity of the ICPP has been the reprocessing of SNF to recover fissile uranium; however, recent changes in world events have diminished the demand to recover and recycle this material. As a result, DOE has discontinued reprocessing SNF for uranium recovery, making the need to properly manage and dispose of these and future materials a high priority. In accordance with the Nuclear Waste Policy Act (NWPA) of 1982, as amended, disposal of SNF and high-level waste (HLW) is planned for a geological repository. Preparation of SNF, HLW, and other radioactive wastes for disposal may include mechanical, physical, and/or chemical processes. This plan outlines the program strategy of the ICPP Spent Fuel and Waste Management Technology Development Program (SF&WMTDP) to develop and demonstrate the technology required to ensure that SNF and radioactive waste will properly stored and prepared for final disposal. Program elements in support of acceptable interim storage and waste minimization include: developing and implementing improved radioactive waste treatment technologies; identifying and implementing enhanced decontamination and decommissioning techniques; developing radioactive scrap metal (RSM) recycle capabilities; and developing and implementing improved technologies for the interim storage of SNF.

  6. Natural phenomena hazards evaluation of equipment and piping of Gaseous Diffusion Plant Uranium Enrichment Facility

    International Nuclear Information System (INIS)

    In support of the Gaseous Diffusion Plant Safety Analysis Report Upgrade program (GDP SARUP), a natural phenomena hazards evaluation was performed for the main process equipment and piping in the uranium enrichment buildings at Paducah and Portsmouth gaseous diffusion plants. In order to reduce the cost of rigorous analyses, the evaluation methodology utilized a graded approach based on an experience data base collected by SQUG/EPRI that contains information on the performance of industrial equipment and piping during past earthquakes. This method consisted of a screening walkthrough of the facility in combination with the use of engineering judgment and simple calculations. By using these screenings combined with evaluations that contain decreasing conservatism, reductions in the time and cost of the analyses were significant. A team of experienced seismic engineers who were trained in the use of the DOE SQUG/EPRI Walkdown Screening Material was essential to the success of this natural phenomena hazards evaluation

  7. Some design and operating aspects of the Ranger uranium mine treatment plant

    International Nuclear Information System (INIS)

    Environmental considerations were key factors in the design of the Ranger Uranium Mines treatment plant. The mine is located adjacent to the Kakadu National Park and has an average rainfall of 1.6m per annum. No contaminated water or liquid effluents are to be released from the project area and thus water management is a key design and operating fact. Particulate and gas emission criteria influenced design as did occupational hygiene factors (dust, radon, housekeeping, maintenance access). Equipment selection and engineering standards were conservative and resulted in the plant attaining design performance in less than three months from the date of commissioning. A number of mechanical and operational problems were experienced. However, none of these problems have had a significant effect on production

  8. Advanced coal gasifier-fuel cell power plant systems design

    Science.gov (United States)

    Heller, M. E.

    1983-01-01

    Two advanced, high efficiency coal-fired power plants were designed, one utilizing a phosphoric acid fuel cell and one utilizing a molten carbonate fuel cell. Both incorporate a TRW Catalytic Hydrogen Process gasifier and regenerator. Both plants operate without an oxygen plant and without requiring water feed; they, instead, require makeup dolomite. Neither plant requires a shift converter; neither plant has heat exchangers operating above 1250 F. Both plants have attractive efficiencies and costs. While the molten carbonate version has a higher (52%) efficiency than the phosphoric acid version (48%), it also has a higher ($0.078/kWh versus $0.072/kWh) ten-year levelized cost of electricity. The phosphoric acid fuel cell power plant is probably feasible to build in the near term: questions about the TRW process need to be answered experimentally, such as weather it can operate on caking coals, and how effective the catalyzed carbon-dioxide acceptor will be at pilot scale, both in removing carbon dioxide and in removing sulfur from the gasifier.

  9. History of Uranium-233(233U)Processing at the Rocky Flats Plant. In support of the RFETS Acceptable Knowledge Program

    International Nuclear Information System (INIS)

    This report documents the processing of Uranium-233 at the Rocky Flats Plant (Rocky Flats Environmental Technology Site). The information may be used to meet Waste Isolation Pilot Plant (WIPP) Waste Acceptance Criteria (WAC)and for determining potential Uranium-233 content in applicable residue waste streams

  10. Relationship Between 210Pb Concentrations in Solid Wastes and Plants from Uranium Mill Tailings

    International Nuclear Information System (INIS)

    Full text: The exploitation and treatment of uranium ore at the Urgeirica mine (north of Portugal) have led to the accumulation of large amounts of soild wastes (mill tailings) in dams (tailing ponds). These solid wastes containing natural radionuclides could be dispersed by the action of atmospheric agents and contaminate the environment. To minimize the dispersion of these radionuclides into the environment the dams were revegetated with pines (Pinus pinea) and eucalyptes (Eucalyptus globolus). The objective of this study is to know the 210Pb behaviour in what concerns its transfer from the uranium mill tailings to these plants, through the analysis of relationships between 210Pb concentrations in the solid wastes and the plants. Solid wastes and plant samples were randomly collected at the dams and the 210Pb activity concentration in solid wastes and plant (aerial part and roots) samples were determined by gamma spectrometry. The results obtained for pines show a good correlation between 210Pb concentrations in the solid wastes and roots. No correlation was found to 210Pb concentrations in the solid wastes and needles. The 210Pb concentration data for eucalyptes show a quite good correlation between 210Pb concentrations in the solid wastes and leaves. Concentration ratio data, solid wastes/roots and solid wastes/needles for pines are on the same order of magnitude. The 210Pb uptake by pines (roots and needles) and eucalyptes (leaves) show that 210Pb concentration ratios decrease at low 210Pb concentrations in the solid wastes and appear relatively constant at higher 210Pb concentrations in the solid wastes. Data presented for 210Pb in this paper, will be compared with those obtained for 226Ra on the same samples. (author)

  11. Literature review: Phytoaccumulation of chromium, uranium, and plutonium in plant systems

    Energy Technology Data Exchange (ETDEWEB)

    Hossner, L.R.; Loeppert, R.H.; Newton, R.J. [Texas A& M Univ., College Station, TX (United States); Szaniszlo, P.J. [Univ. of Texas, Austin, TX (United States)

    1998-05-01

    Phytoremediation is an integrated multidisciplinary approach to the cleanup of contaminated soils, which combines the disciplines of plant physiology, soil chemistry, and soil microbiology. Metal hyperaccumulator plants are attracting increasing attention because of their potential application in decontamination of metal-polluted soils. Traditional engineering technologies may be too expensive for the remediation of most sites. Removal of metals from these soils using accumulator plants is the goal of phytoremediation. The emphasis of this review has been placed on chromium (Cr), plutonium (Pu), and uranium (U). With the exception of Cr, these metals and their decay products exhibit two problems, specifically, radiation dose hazards and their chemical toxicity. The radiation hazard introduces the need for special precautions in reclamation beyond that associated with non-radioactive metals. The uptake of beneficial metals by plants occurs predominantly by way of channels, pores, and transporters in the root plasma membrane. Plants characteristically exhibit a remarkable capacity to absorb what they need and exclude what they don`t need. But most vascular plants absorb toxic and heavy metals through their roots to some extent, though to varying degrees, from negligible to substantial. Sometimes absorption occurs because of the chemical similarity between beneficial and toxic metals. Some plants utilize exclusion mechanisms, where there is a reduced uptake by the roots or a restricted transport of the metal from root to shoot. At the other extreme, hyperaccumulator plants absorb and concentrate metals in both roots and shoots. Some plant species endemic to metalliferous soils accumulate metals in percent concentrations in the leaf dry matter.

  12. Impact of uranium concentration reduction in side plates of the fuel elements of IEA-R1 reactor on neutronic and thermal hydraulic analyses

    International Nuclear Information System (INIS)

    This master thesis presents a study to verify the impact of the uranium concentration reduction in the side plates of the reactor IEA-R1 fuel elements on the neutronic and thermal-hydraulic analyses. To develop such study, a previous IPEN-CNEN/SP research was reproduced by simulating the fuel elements burn-up, with side plate uranium density reduced to 50, 60 and 70% of the standard fuel element plates. This research begins with the neutronic analysis using the computer code HAMMER and the first step consists in the calculation of the cross section of all materials presented at the reactor core, with their initial concentration; the second step consists in the calculation of the fast and thermal neutron group fluxes and power densities for fuel elements using the computer code CITATION. HAMMER output data is used as input data. Once the neutronic analysis is finished and the most critical fuel elements with highest power density have been defined, the thermal-hydraulics analysis begins. This analysis uses MCTR-IEA-R1 thermal-hydraulics model, which equations are solved by commercial code EES. Thermalhydraulics analysis input is the power density data calculated by CITATION: it is considered the highest power density on each fuel element, where there is a higher energy release and, consequently, higher temperatures. This data is used on energy balance equations to calculate temperatures on critical fuel element regions. Reactor operation comparison for three different uranium densities on fuel side plates is presented. Uranium density reduction contributes to the cladding surface temperature to remain below the established limit, as reactor operation safety requirement and it does not affect significantly fuel element final burn-up nor reactor reactivity. The reduction of uranium in the side plates of the fuel elements of the IEA-R1 showed to be a viable option to avoid corrosion problems due to high temperatures. (author)

  13. Research and development projects, new processes binded at the stop of a fuel fabrication plant (Nukem-A)

    International Nuclear Information System (INIS)

    This research work is aimed at the assessment of new procedures in the framework of the decommissioning of the NUKEM-A facility, a plant used for over 30 years for the fabrication of Material Test Reactor (MTR) and Thorium High Temperature Reactor (THTR) fuel elements. Important issues in this work are the preparation of detailed uranium and thorium contamination distribution maps in walls and floors, the execution of various dismantling and decontamination operations under health physics control, the large-scale treatment of arising primary waste and the minimization of secondary waste

  14. Opening remarks [International symposium on uranium production and raw materials for the nuclear fuel cycle - Supply and demand, economics, the environment and energy security

    International Nuclear Information System (INIS)

    The International Atomic Energy Agency takes a great deal of pride in the role that it played in the early 1990s in bringing together the international uranium community to begin to develop a comprehensive database on uranium supply and to openly exchange ideas on uranium geology, production technology and economics and environmental issues associated with uranium production. One of the most valuable biannual publications jointly issued by OECD/NEA and the IAEA is URANIUM: Resources, Production and demand popularly know as Red Book. The Joint OECD/NEA - IAEA Uranium Group also supports the activities in the front-end of the nuclear fuel cycle. The Agency is committed to serve as a forum for discussion on all aspects of uranium supply and demand and hence this symposium. The ultimate goal of the uranium industry can be stated rather simply - providing an adequate supply of uranium that can be delivered to the market place at competitive prices by environmentally sound production practices. Achieving that goal is, however, anything but simple. The industry faces a number of challenges that can best be met through exchanging ideas at meetings such as this Symposium. New exploration techniques may help to broaden the geographic diversity of the uranium resource base. Diversity of supply in turn can help offset the risk arising out of unforeseen delay in executing uranium mining projects in some areas. Sharing information on new production technology can potentially lead to improved project economics and an increase in the low-cost resource base. Case histories ranging from pre-production licensing to final site rehabilitation are invaluable guides to the time frames and costs associated with environmental planning and compliance. An adequate supply of uranium is essential to the future of nuclear power. This Symposium is intended to bring together the broadly diverse expertise needed to assure that future

  15. Fabrication procedures for manufacturing high uranium concentration dispersion fuel elements; Procedimentos de fabricacao de elementos combustiveis a base de dispersoes com alta concentracao de uranio

    Energy Technology Data Exchange (ETDEWEB)

    Souza, J.A.B.; Durazzo, M., E-mail: jasouza@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2010-07-01

    IPEN developed and made available for routine production the technology for manufacturing dispersion type fuel elements for use in research reactors. However, the fuel produced at IPEN is limited to the uranium concentration of 3.0 gU/cm{sup 3} by using the U{sub 3}Si{sub 2}-Al dispersion. Increasing the uranium concentration of the fuel is interesting by the possibility of increasing the reactor core reactivity and lifetime of the fuel. It is possible to increase the concentration of uranium in the fuel up to the technological limit of 4.8 gU/cm{sup 3} for the U{sub 3}Si{sub 2}-Al dispersion, which is well placed around the world. This new fuel will be applicable in the new Brazilian-Multipurpose Reactor RMB. This study aimed to develop the manufacturing process of high uranium concentration fuel, redefining the procedures currently used in the manufacture of IPEN. This paper describes the main procedures adjustments that will be necessary. (author)

  16. Chemical aspects of the precise and accurate determination of uranium and plutonium from nuclear fuel solutions

    International Nuclear Information System (INIS)

    A method for the simultaneous or separate determination of uranium and plutonium has been developed. The method is based on the sorption of uranium and plutonium as their chloro complexes on Dowex 1x10 column. When separate uranium and plutonium fractions are desired, plutonium ions are reduced to Pu (III) and eluted, after which the uranium ions are eluted with dilute HCl. Simultaneous stripping of a mass ratio U/Pu approximately 1 fraction for mass spectrometric measurements is achieved by proper choice of eluant HC1 concentration. Special attention was paid to the obtaining of americium free plutonium fractions. The distribution coefficient measurements showed that at 12.5-M HCl at least 30 % of americium ions formed anionic chloro complexes. The chemical aspects of isotopic fractionation in a multiple filament thermal ionization source were also investigated. Samples of uranium were loaded as nitrates, chlorides, and sulphates and the dependence of the measured uranium isotopic ratios on the chemical form of the loading solution as well as on the filament material was studied. Likewise the dependence of the formation of uranium and its oxide ions on various chemical and instrumental conditions was investigated using tungsten and rhenium filaments. Systematic errors arising from the chemical conditions are compared with errors arising from the automatic evaluation of of spectra. (author)

  17. Cost and availability of gadolinium for nuclear fuel reprocessing plants

    International Nuclear Information System (INIS)

    Gadolinium is currently planned for use as a soluble neutron poison in nuclear fuel reprocessing plants to prevent criticality of solutions of spent fuel. Gadolinium is relatively rare and expensive. The present study was undertaken therefore to estimate whether this material is likely to be available in quantities sufficient for fuel reprocessing and at reasonable prices. It was found that gadolinium, one of 16 rare earth elements, appears in the marketplace as a by-product and that its present supply is a function of the production rate of other more prevalent rare earths. The potential demand for gadolinium in a fuel reprocessing facility serving a future fast reactor industry amounts to only a small fraction of the supply. At the present rate of consumption, domestic supplies of rare earths containing gadolinium are adequate to meet national needs (including fuel reprocessing) for over 100 years. With access to foreign sources, US demands can be met well beyond the 21st century. It is concluded therefore that the supply of gadolinium will quite likely be more than adequate for reprocessing spent fuel for the early generation of fast reactors. The current price of 99.99% pure gadolinium oxide lies in the range $50/lb to $65/lb (1984 dollars). By the year 2020, in time for reprocessing spent fuel from an early generation of large fast reactors, the corresponding values are expected to lie in the $60/lb to $75/lb (1984 dollars) price range. This increase is modest and its economic impact on nuclear fuel reprocessing would be minor. The economic potential for recovering gadolinium from the wastes of nuclear fuel reprocessing plants (which use gadolinium neutron poison) was also investigated. The cost of recycled gadolinium was estimated at over twelve times the cost of fresh gadolinium, and thus recycle using current recovery technology is not economical. 15 refs., 4 figs., 11 tabs

  18. Nuclear fuel

    International Nuclear Information System (INIS)

    It is expected that nuclear power generation will reach 49 million kW in 1985 and 129 million kW in 1995, and the nuclear fuel having to be supplied and processed will increase in proportion to these values. The technical problems concerning nuclear fuel are presented on the basis of the balance between the benefit for human beings and the burden on the human beings. Recently, especially the downstream of nuclear fuel attracts public attention. Enriched uranium as the raw material for light water reactor fuel is almost monopolized by the U.S., and the technical information has not been published for fear of the diversion to nuclear weapons. In this paper, the present situations of uranium enrichment, fuel fabrication, transportation, reprocessing and waste disposal and the future problems are described according to the path of nuclear fuel cycle. The demand and supply of enriched uranium in Japan will be balanced up to about 1988, but afterwards, the supply must rely upon the early establishment of the domestic technology by centrifugal separation method. No problem remains in the fabrication of light water reactor fuel, but for the fabrication of mixed oxide fuel, the mechanization of the production facility and labor saving are necessary. The solution of the capital risk for the construction of the second reprocessing plant is the main problem. Japan must develop waste disposal techniques with all-out efforts. (Kako, I.)

  19. The element technology of clean fuel alcohol plant construction

    Energy Technology Data Exchange (ETDEWEB)

    Lee, D.S; Lee, D.S. [Sam-Sung Engineering Technical Institute (Korea, Republic of); Choi, C.Y [Seoul National University, Seoul (Korea, Republic of)] [and others

    1996-02-01

    The fuel alcohol has been highlighted as a clean energy among new renewable energy sources. However, the production of the fuel alcohol has following problems; (i)bulk distillate remains is generated and (ii) benzene to be used as a entertainer in the azeotropic distillation causes the environmental problem. Thus, we started this research on the ground of preserving the cleanness in the production of fuel alcohol, a clean energy. We examined the schemes of replacing the azotropic distillation column which causes the problems with MSDP(Molecular Sieve Dehydration Process) system using adsorption technology and of treating the bulk distillate remains to be generated as by-products. In addition, we need to develop the continuous yea station technology for the continuous operation of fuel alcohol plant as a side goal. Thus, we try to develop a continuous ethanol fermentation process by high-density cell culture from tapioca, a industrial substrate, using cohesive yeast. For this purpose, we intend to examine the problem of tapioca, a industrial substrate, where a solid is existed and develop a new process which can solve the problem. Ultimately, the object of this project is to develop each element technology for the construction of fuel alcohol plant and obtain the ability to design the whole plant. (author) 54 refs., 143 figs., 34 tabs.

  20. Preconceptual design studies and cost data of depleted uranium hexafluoride conversion plants

    Energy Technology Data Exchange (ETDEWEB)

    Jones, E

    1999-07-26

    One of the more important legacies left with the Department of Energy (DOE) after the privatization of the United States Enrichment Corporation is the large inventory of depleted uranium hexafluoride (DUF6). The DOE Office of Nuclear Energy, Science and Technology (NE) is responsible for the long-term management of some 700,000 metric tons of DUF6 stored at the sites of the two gaseous diffusion plants located at Paducah, Kentucky and Portsmouth, Ohio, and at the East Tennessee Technology Park in Oak Ridge, Tennessee. The DUF6 management program resides in NE's Office of Depleted Uranium Hexafluoride Management. The current DUF6 program has largely focused on the ongoing maintenance of the cylinders containing DUF6. However, the long-term management and eventual disposition of DUF6 is the subject of a Programmatic Environmental Impact Statement (PEIS) and Public Law 105-204. The first step for future use or disposition is to convert the material, which requires construction and long-term operation of one or more conversion plants. To help inform the DUF6 program's planning activities, it was necessary to perform design and cost studies of likely DUF6 conversion plants at the preconceptual level, beyond the PEIS considerations but not as detailed as required for conceptual designs of actual plants. This report contains the final results from such a preconceptual design study project. In this fast track, three month effort, Lawrence Livermore National Laboratory and Bechtel National Incorporated developed and evaluated seven different preconceptual design cases for a single plant. The preconceptual design, schedules, costs, and issues associated with specific DUF6 conversion approaches, operating periods, and ownership options were evaluated based on criteria established by DOE. The single-plant conversion options studied were similar to the dry-conversion process alternatives from the PEIS. For each of the seven cases considered, this report contains

  1. Initial evaluation of dry storage issues for spent nuclear fuels in wet storage at the Idaho Chemical Processing Plant

    Energy Technology Data Exchange (ETDEWEB)

    Guenther, R J; Johnson, Jr, A B; Lund, A L; Gilbert, E R [and others

    1996-07-01

    The Pacific Northwest Laboratory has evaluated the basis for moving selected spent nuclear fuels in the CPP-603 and CPP-666 storage pools at the Idaho Chemical Processing Plant from wet to dry interim storage. This work is being conducted for the Lockheed Idaho Technologies Company as part of the effort to determine appropriate conditioning and dry storage requirements for these fuels. These spent fuels are from 22 test reactors and include elements clad with aluminum or stainless steel and a wide variety of fuel materials: UAl{sub x}, UAl{sub x}-Al and U{sub 3}O{sub 8}-Al cermets, U-5% fissium, UMo, UZrH{sub x}, UErZrH, UO{sub 2}-stainless steel cermet, and U{sub 3}O{sub 8}-stainless steel cermet. The study also included declad uranium-zirconium hydride spent fuel stored in the CPP-603 storage pools. The current condition and potential failure mechanisms for these spent fuels were evaluated to determine the impact on conditioning and dry storage requirements. Initial recommendations for conditioning and dry storage requirements are made based on the potential degradation mechanisms and their impacts on moving the spent fuel from wet to dry storage. Areas needing further evaluation are identified.

  2. Initial evaluation of dry storage issues for spent nuclear fuels in wet storage at the Idaho Chemical Processing Plant

    International Nuclear Information System (INIS)

    The Pacific Northwest Laboratory has evaluated the basis for moving selected spent nuclear fuels in the CPP-603 and CPP-666 storage pools at the Idaho Chemical Processing Plant from wet to dry interim storage. This work is being conducted for the Lockheed Idaho Technologies Company as part of the effort to determine appropriate conditioning and dry storage requirements for these fuels. These spent fuels are from 22 test reactors and include elements clad with aluminum or stainless steel and a wide variety of fuel materials: UAlx, UAlx-Al and U3O8-Al cermets, U-5% fissium, UMo, UZrHx, UErZrH, UO2-stainless steel cermet, and U3O8-stainless steel cermet. The study also included declad uranium-zirconium hydride spent fuel stored in the CPP-603 storage pools. The current condition and potential failure mechanisms for these spent fuels were evaluated to determine the impact on conditioning and dry storage requirements. Initial recommendations for conditioning and dry storage requirements are made based on the potential degradation mechanisms and their impacts on moving the spent fuel from wet to dry storage. Areas needing further evaluation are identified

  3. Arizona Public Service - Alternative Fuel (Hydrogen) Pilot Plant Design Report

    Energy Technology Data Exchange (ETDEWEB)

    James E. Francfort

    2003-12-01

    Hydrogen has promise to be the fuel of the future. Its use as a chemical reagent and as a rocket propellant has grown to over eight million metric tons per year in the United States. Although use of hydrogen is abundant, it has not been used extensively as a transportation fuel. To assess the viability of hydrogen as a transportation fuel and the viability of producing hydrogen using off-peak electric energy, Pinnacle West Capital Corporation (PNW) and its electric utility subsidiary, Arizona Public Service (APS) designed, constructed, and operates a hydrogen and compressed natural gas fueling station—the APS Alternative Fuel Pilot Plant. This report summarizes the design of the APS Alternative Fuel Pilot Plant and presents lessons learned from its design and construction. Electric Transportation Applications prepared this report under contract to the U.S. Department of Energy’s Advanced Vehicle Testing Activity. The Idaho National Engineering and Environmental Laboratory manages these activities for the Advanced Vehicle Testing Activity.

  4. Addendum to the Book of Abstracts for the international symposium on uranium raw material for the nuclear fuel cycle: Exploration, mining, production, supply and demand, economics and environmental issues (URAM-2009)

    International Nuclear Information System (INIS)

    The International Symposium on Uranium Raw Material for the Nuclear Fuel Cycle: Exploration, Mining, Production, Supply and Demand, Economics and Environmental Issues (URAM-2009) addressed all aspects of the uranium fuel cycle, from the availability of raw materials to the long-term sustainability of nuclear power. The revival of the uranium industry in recent years has caused a dramatic increase in uranium exploration and mining activities in several countries. URAM-2009 was intended to bring together scientists, exploration and mining geologists, engineers, operators, regulators and fuel cycle specialists to exchange information and discuss updated research and current issues in uranium geology and deposits, exploration, mining and processing, production economics, and environmental and legal issues. This addendum contains a key note presentation on nuclear energy and sustainability, and seven poster presentations on uranium exploration, geology and uranium mining and processing. All presentations are indexed individually

  5. Preliminary experience and near future utilization programmes of the MPR-30 fueled by LEU [low enriched uranium

    International Nuclear Information System (INIS)

    The MTR type reactor MPR-30 G.A. Siwabessy, located at PUSPIPTEK Serpong has recently reached its first criticality. This multipurpose reactor is supposed to be the first MTR type reactor in the world that is designed and constructed to be fueled by low enriched uranium. Preliminary experience covering the approach to the first criticality and the excess reactivity loading as well as some thermal hydraulics and power ascension tests are briefly presented and discussed. The near future utilization programmes during and after commissioning are also presented. (Author)

  6. Standard model for safety analysis report of fuel reprocessing plants

    International Nuclear Information System (INIS)

    A standard model for a safety analysis report of fuel reprocessing plants is established. This model shows the presentation format, the origin, and the details of the minimal information required by CNEN (Comissao Nacional de Energia Nuclear) aiming to evaluate the requests of construction permits and operation licenses made according to the legislation in force. (E.G.)

  7. China's First Orimulsion-Fueled Power Plant Constructed in Zhanjiang

    Institute of Scientific and Technical Information of China (English)

    Xing Jianxiao

    2003-01-01

    @@ A ceremony commencement of Orimulsion-fueled electric power generation project was held in Zhanjiang City,Guangdong Province, on November 26. Vice Governor of Guangdong Province You Ningfeng, Vice President of PetroChina and other officials from the related departments attended the foundation-laying ceremony for the power plant.

  8. Cost and quality of fuels for electric utility plants, 1992

    Energy Technology Data Exchange (ETDEWEB)

    1993-08-02

    This publication presents an annual summary of statistics at the national, Census division, State, electric utility, and plant levels regarding the quantity, quality, and cost of fossil fuels used to produce electricity. The purpose of this publication is to provide energy decision-makers with accurate and timely information that may be used in forming various perspectives on issues regarding electric power.

  9. Standard model for safety analysis report of fuel fabrication plants

    International Nuclear Information System (INIS)

    A standard model for a safety analysis report of fuel fabrication plants is established. This model shows the presentation format, the origin, and the details of the minimal information required by CNEN (Comissao Nacional de Energia Nuclear) aiming to evaluate the requests of construction permits and operation licenses made according to the legislation in force. (E.G.)

  10. Cost and quality of fuels for electric utility plants, 1994

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-14

    This document presents an annual summary of statistics at the national, Census division, State, electric utility, and plant levels regarding the quantity, quality, and cost of fossil fuels used to produce electricity. Purpose of this publication is to provide energy decision-makers with accurate, timely information that may be used in forming various perspectives on issues regarding electric power.

  11. Contribution to the study of nuclear fuel materials with a metallic uranium base; Contribution a l'etude des materiaux combustibles nucleaires a base d'uranium metallique

    Energy Technology Data Exchange (ETDEWEB)

    Englander, M. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1957-11-15

    In a power reactor destined to supply industrially recoverable thermal energy, the most economical source of heat still consists of natural metallic uranium. However, the nuclear fuel material, most often employed in the form of rods of 20 to 40 mm diameter, is subjected to a series of stresses which lead to irreversible distortions usually incompatible with the substructure of the reactor. As a result the fuel material must possess at the outset a certain number of qualities which must be determined. Investigations have therefore been carried out, first on the technological characters peculiar to each of the three allotropic phases of pure uranium metal, and on their interactions on the stabilisation of the material which consists of either cast uranium or uranium pile-treated in the {gamma} phase. (author) [French] Dans un reacteur de puissance destine a fournir de l'energie thermique industriellement recuperable, la source de chaleur la plus economique reste constituee par de l'uranium metallique naturel. Or, le materiau combustible nucleaire, employe le plus souvent sous forme de barreaux de 20 a 40 mm de diametre, se trouve soumis a un ensemble de contraintes qui provoque des deformations irreversibles, le plus souvent incompatibles avec l'infrastructure du reacteur. Par consequent, le materiau combustible doit presenter a l'origine un certain nombre de qualites qu'il est necessaire de determiner. Aussi a-t-on d'abord etudie les caracteres technologiques propres a chacune des trois phases allotropiques de l'uranium-metal pur et leurs interactions sur la stabilisation du materiau constitue soit par de l'uranium coule, soit par de l'uranium traite en pile en phase {gamma}. (auteur)

  12. Radioactive contamination at nuclear fuel cycle facilities. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Bernhardt, D.E.; Grant, M.W.; Rich, D.C.; Jensen, C.M.; Macbeth, P.J.

    1982-01-01

    This report presents information to characterize uranium fuel cycle facilities (excluding reactors), levels of contamination at those facilities, and volumes and activity of wastes associated with their decontamination and decommissioning (DandD). It is one of a series of reports providing information to assist the U.S. Environmental Protection Agency in setting standards and guidelines for permissible residual levels of radioactivity from DandD. The categories of facilities covered by this report are: Uranium mines, Uranium mills, Uranium hexafluoride conversion plants, Fuel fabrication plants, including both low and high enriched uranium and mixed oxide facilities. Both active and inactive facilities are identified. The highest volumes of DandD waste (hundreds of millions of cubic meters) are associated with uranium mines, while the highest amounts of radioactivity are a result of DandD at fuel reprocessing plants.

  13. Standard test method for analysis of isotopic composition of uranium in nuclear-grade fuel material by quadrupole inductively coupled plasma-mass spectrometry

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2000-01-01

    1.1 This test method is applicable to the determination of the isotopic composition of uranium (U) in nuclear-grade fuel material. The following isotopic weight percentages are determined using a quadrupole inductively coupled plasma-mass spectrometer (Q-ICP-MS): 233U, 234U, 235U, 236U, and 238U. The analysis can be performed on various material matrices after acid dissolution and sample dilution into water or dilute nitric (HNO3) acid. These materials include: fuel product, uranium oxide, uranium oxide alloys, uranyl nitrate (UNH) crystals, and solutions. The sample preparation discussed in this test method focuses on fuel product material but may be used for uranium oxide or a uranium oxide alloy. Other preparation techniques may be used and some references are given. Purification of the uranium by anion-exchange extraction is not required for this test method, as it is required by other test methods such as radiochemistry and thermal ionization mass spectroscopy (TIMS). This test method is also described i...

  14. Integrating fuel cell power systems into building physical plants

    Energy Technology Data Exchange (ETDEWEB)

    Carson, J. [KCI Technologies, Inc., Hunt Valley, MD (United States)

    1996-12-31

    This paper discusses the integration of fuel cell power plants and absorption chillers to cogenerate chilled water or hot water/steam for all weather air conditioning as one possible approach to building system applications. Absorption chillers utilize thermal energy in an absorption based cycle to chill water. It is feasible to use waste heat from fuel cells to provide hydronic heating and cooling. Performance regimes will vary as a function of the supply and quality of waste heat. Respective performance characteristics of fuel cells, absorption chillers and air conditioning systems will define relationships between thermal and electrical load capacities for the combined systems. Specifically, this paper develops thermodynamic relationships between bulk electrical power and cooling/heating capacities for combined fuel cell and absorption chiller system in building applications.

  15. Pilot uranium lysimeter studies at the Oak Ridge Y-12 plant

    International Nuclear Information System (INIS)

    A field lysimeter test facility has been constructed at the Oak Ridge Y-12 plant to evaluate land burial of wastes containing depleted uranium. The principal objective in the construction of such a facility is to provide a means for monitoring waste leachate characteristics over time, in particular uranium concentrations in leachate. The design of the field lysimeter test facility allows, via the portals along the side walls of the lysimeter, the collection of leachate as a function of depth in the lysimeter. The methodology to collect leachate from within the field lysimeter has not been clearly defined. Thus, before wastes were loaded into the field lysimeter facility, a pilot lysimeter study was initiated to test several design concepts for the collection of in situ leachate. The primary objective of this pilot study was to demonstrate the feasibility and quality assurance of proposed instrumentation used to monitor leachate generation and characteristics in the full-scale field lysimeter. Secondary objectives included gaining experience in the handling/packing of wastes, installation/operation of the leachate collection devices, and waste leachate characterization

  16. Pilot uranium lysimeter studies at the Oak Ridge Y-12 plant

    Energy Technology Data Exchange (ETDEWEB)

    Francis, C.W.; Hyder, L.K.; Howard, S.C.; Cline, J.E.; Clapp, R.B.

    1993-08-01

    A field lysimeter test facility has been constructed at the Oak Ridge Y-12 plant to evaluate land burial of wastes containing depleted uranium. The principal objective in the construction of such a facility is to provide a means for monitoring waste leachate characteristics over time, in particular uranium concentrations in leachate. The design of the field lysimeter test facility allows, via the portals along the side walls of the lysimeter, the collection of leachate as a function of depth in the lysimeter. The methodology to collect leachate from within the field lysimeter has not been clearly defined. Thus, before wastes were loaded into the field lysimeter facility, a pilot lysimeter study was initiated to test several design concepts for the collection of in situ leachate. The primary objective of this pilot study was to demonstrate the feasibility and quality assurance of proposed instrumentation used to monitor leachate generation and characteristics in the full-scale field lysimeter. Secondary objectives included gaining experience in the handling/packing of wastes, installation/operation of the leachate collection devices, and waste leachate characterization

  17. Assessment of enriched uranium storage safety issues at the Oak Ridge Y-12 Plant

    International Nuclear Information System (INIS)

    This document is an assessment of the technical safety issues pertaining to the storage of EU at the Oak Ridge Y-12 Plant. The purpose of the assessment is to serve as the basis for defining the technical standards for storage of EU at Y-12. A formal assessment of the Y-12 materials acceptance criteria for EU is currently being conducted by a task force cochaired by B. G. Eddy of DOE Oak Ridge Operations and S. 0. Cox of Y-12 Defense Programs. The mission of this technical assessment for storage is obviously dependent on results of the acceptance assessment. Clearly, the two efforts require coordination to avoid inconsistencies. In addition, both these Assessments must be consistent with the Environmental Assessment for EU storage at Y-12.1 Both the Storage Assessment and the Criteria for Acceptance must take cognizance of the fact that a portion of the EU to be submitted for storage in the future is expected to be derived from foreign sources and to include previously irradiated uranium containing significant levels of transuranics, radioactive daughter products, and unstable uranium isotopes that do not occur in the EU stream of the DOE weapons complex. National security considerations may dictate that these materials be accepted despite the fact that they fail to conform to the Acceptance Criteria. This document will attempt to address the complexities inherent in this situation

  18. Assessment of enriched uranium storage safety issues at the Oak Ridge Y-12 Plant

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-08-01

    This document is an assessment of the technical safety issues pertaining to the storage of EU at the Oak Ridge Y-12 Plant. The purpose of the assessment is to serve as the basis for defining the technical standards for storage of EU at Y-12. A formal assessment of the Y-12 materials acceptance criteria for EU is currently being conducted by a task force cochaired by B. G. Eddy of DOE Oak Ridge Operations and S. 0. Cox of Y-12 Defense Programs. The mission of this technical assessment for storage is obviously dependent on results of the acceptance assessment. Clearly, the two efforts require coordination to avoid inconsistencies. In addition, both these Assessments must be consistent with the Environmental Assessment for EU storage at Y-12.1 Both the Storage Assessment and the Criteria for Acceptance must take cognizance of the fact that a portion of the EU to be submitted for storage in the future is expected to be derived from foreign sources and to include previously irradiated uranium containing significant levels of transuranics, radioactive daughter products, and unstable uranium isotopes that do not occur in the EU stream of the DOE weapons complex. National security considerations may dictate that these materials be accepted despite the fact that they fail to conform to the Acceptance Criteria. This document will attempt to address the complexities inherent in this situation.

  19. Technology, Safety and Costs of Decommissioning a Reference Uranium Hexafluoride Conversion Plant

    Energy Technology Data Exchange (ETDEWEB)

    Elder, H. K.

    1981-10-01

    Safety and cost information is developed for the conceptual decommissioning of a commercial uranium hexafluoride conversion (UF{sub 6}) plant. Two basic decommissioning alternatives are studied to obtain comparisons between cost and safety impacts: DECON, and passive SAFSTOR. A third alternative, DECON of the plant and equipment with stabilization and long-term care of lagoon wastes. is also examined. DECON includes the immediate removal (following plant shutdown) of all radioactivity in excess of unrestricted release levels, with subsequent release of the site for public use. Passive SAFSTOR requires decontamination, preparation, maintenance, and surveillance for a period of time after shutdown, followed by deferred decontamination and unrestricted release. DECON with stabilization and long-term care of lagoon wastes (process wastes generated at the reference plant and stored onsite during plant operation} is also considered as a decommissioning method, although its acceptability has not yet been determined by the NRC. The decommissioning methods assumed for use in each decommissioning alternative are based on state-of-the-art technology. The elapsed time following plant shutdown required to perform the decommissioning work in each alternative is estimated to be: for DECON, 8 months; for passive SAFSTOR, 3 months to prepare the plant for safe storage and 8 months to accomplish deferred decontamination. Planning and preparation for decommissioning prior to plant shutdown is estimated to require about 6 months for either DECON or passive SAFSTOR. Planning and preparation prior to starting deferred decontamination is estimated to require an additional 6 months. OECON with lagoon waste stabilization is estimated to take 6 months for planning and about 8 months to perform the decommissioning work. Decommissioning cost, in 1981 dollars, is estimated to be $5.91 million for OECON. For passive SAFSTOR, preparing the facility for safe storage is estimated to cost $0

  20. Uranium fate in wetland mesocosms: Effects of plants at two iron loadings with different pH values

    Science.gov (United States)

    Small-scale continuous flow wetland mesocosms (~0.8 L) were used to evaluate how plant roots under different iron loadings affect uranium (U) mobility. When significant concentrations of ferrous iron (Fe) were present at circumneutral pH values, U concentrations in root exposed ...

  1. Materials safeguards and accountability in the low enriched uranium conversion-fabrication sector of the fuel cycle

    International Nuclear Information System (INIS)

    Today materials accounting in the low enriched conversion-fabrication sector of the LWR fuel cycle is of increased importance. Low enriched uranium is rapidly becoming a precious metal with current dollar values in the range of one dollar per gram comparing with gold and platinum at 7-8 dollars per gram. In fact, people argue that its dollar value exceeds its safeguards value. Along with this increased financial incentive for better material control, the nuclear industry is faced with the impending implementation of international safeguards and increased public attention over its ability to control nuclear materials. Although no quantity of low enriched uranium (LEU) constitutes a practical nuclear explosive, its control is important to international safeguards because of plutonium production or further enrichment to an explosive grade material. The purpose of the paper is to examine and discuss some factors in the area of materials safeguards and accountability as they apply to the low enriched uranium conversion-fabrication sector. The paper treats four main topics: basis for materials accounting; our assessment of the proposed new IAEA requirements; adequacy of current practices; and timing and direction of future modifications

  2. Update on uranium-molybdenum fuel foil fabrication development activities at the Y-12 National Security Complex in 2007

    International Nuclear Information System (INIS)

    In support of the RERTR Program, efforts are underway at Y-12 to develop and validate a production oriented, monolithic uranium molybdenum (U-Mo) foil fabrication process adaptable for potential implementation in a manufacturing environment. These efforts include providing full-scale prototype depleted and enriched U-Mo foils in support of fuel qualification testing. The work has three areas of focus; develop and demonstrate a feasible foil fabrication process utilizing depleted uranium-molybdenum (DU-Mo) source material, transition these production techniques to enriched uranium (EU-Mo) source material, and evaluate full-scale implementation of the developed production techniques. In 2006, Y-12 demonstrated successful fabrication of full-size DU-10Mo foils. In 2007, Y-12 activities were expanded to include continued DU-Mo foil fabrication with a focus on process refinement, source material impurity effects (specifically carbon), and the feasibility of physical vapor deposition (PVD) on DU-10Mo mini-foils. FY2007 activities also included a transition to EU-Mo and fabrication of full-size enriched foils. The purpose of this report is to update the RERTR audience on Y-12 efforts in 2007 that support the overall RERTR Program goals. (author)

  3. Determination of uranium in urine samples of fuel element fabrication workers by beta-delayed neutron counting

    Science.gov (United States)

    Gabelmann, H.; Lerch, M.; Kratz, K.-L.; Rudolph, W.

    1984-06-01

    Within the health physics examination of fuel element fabrication workers, the control of uranium incorporation is of importance. This is commonly performed by the determination of the alpha activity concentration of uranium excreted in the urine. However, since the chemical separation procedure and the preparation of alpha-counting samples are complicated and time-consuming, this method may imply restrictions on the routine control of large numbers of persons. Therefore, we have investigated the applicability of measuring the beta-delayed neutrons from thermal neutron induced fission of the 235U in the urine samples. The uranium was separated by coprecipitation with Fe(OH) 3 from the urine samples and irradiated in a rabbit system of the Mainz TRIGA reactor. The neutrons were counted with a 3He long counter. The detection limit of 0.3 to 0.9 pCi 1 -1 is comparable to that of alpha spectrometry, but the time required for one sample, from preparation to data evaluation is less than 25 min.

  4. The New Generation of Uranium In Situ Recovery Facilities: Design Improvements should Reduce Radiological Impacts Relative to First Generation Uranium Solution Mining Plants

    International Nuclear Information System (INIS)

    In the last few years, there has been a significant increase in the demand for Uranium as historical inventories have been consumed and new reactor orders are being placed. Numerous mineralized properties around the world are being evaluated for Uranium recovery and new mining / milling projects are being evaluated and developed. Ore bodies which are considered uneconomical to mine by conventional methods such as tunnelling or open pits, can be candidates for non-conventional recovery techniques, involving considerably less capital expenditure. Technologies such as Uranium In Situ Leaching/In Situ Recovery (ISL/ISR - also referred to as 'solution mining'), have enabled commercial scale mining and milling of relatively small ore pockets of lower grade, and are expected to make a significant contribution to overall world wide uranium supplies over the next ten years. Commercial size solution mining production facilities have operated in the US since the late 1960s. However, current designs are expected to result in less radiological wastes and emissions relative to these 'first' generation plants (which were designed, constructed and operated through the 1980s) which typically used alkaline leach chemistries in situ, open to air recovery vessels and high temperature calcining systems for final product drying. Improved containment, automation and instrumentation control and use of vacuum dryers in the design of current generation plants are expected to reduce production of secondary waste byproduct material, reduce Radon emissions and reduce potential for employee exposure to uranium concentrate aerosols at the back end of the milling process. In Situ Recovery involves the circulation of groundwater, fortified with oxidizing and complexing agents into an ore body, solubilizing the uranium in situ, and then pumping the solutions to the surface where they are fed to a processing plant (mill). Processing involves ion exchange and may also include precipitation, drying or

  5. Evaluation of terrestrial plants extracts for uranium sorption and characterization of potent phytoconstituents.

    Science.gov (United States)

    Sharma, Sunita; Singh, Bikram; Thulasidas, S K; Kulkarni, Madhuri J; Natarajan, V; Manchanda, Vijay K

    2016-01-01

    Sorption capacity of four plants (Funaria hygrometrica, Musa acuminata, Brassica juncea and Helianthus annuus) extracts/fractions for uranium, a radionuclide was investigated by EDXRF and tracer studies. The maximum sorption capacity, i.e., 100% (complete sorption) was observed in case of Musa acuminata extract and fractions. Carbohydrate, proteins, phenolics and flavonoids contents in the active fraction (having maximum sorption capacity) were also determined. Further purification of the most active fraction provided three pure molecules, mannitol, sorbitol and oxo-linked potassium oxalate. The characterization of isolated molecules was achieved by using FTIR, NMR, GC-MS, MS-MS, and by single crystal-XRD analysis. Of three molecules, oxo-linked potassium oxalate was observed to have 100% sorption activity. Possible binding mechanism of active molecule with the uranyl cation has been purposed.

  6. Investigation report on criticality accident at the Uranium Processing Plant of the JCO, Ltd

    International Nuclear Information System (INIS)

    This report is a summarized one of investigation results on a criticality accident at the Uranium Processing Plant of the JCO, Ltd., carried out by the Nuclear Safety Investigation Special Group (SISG) of the Atomic Energy Society of Japan (AESJ). AESJ published a statement of the president on this accident on October 8, 1999, and decided to perform its investigation under SISG. SISG carried out some questionnaires for new trials together with conventional lectures of the well-informed. This report contains six chapters on critical safety and accident, process of the accident and elucidation of its facts, cause analysis and picking-out on problems, questionnaires on improvement proposal', questionnaires on 'duty of AESJ7, and future efforts on nuclear safety culture. At the last chapter, SISG discussed about some items on re-occurrence protection of the nuclear accident. (G.K.)

  7. Evaluation of terrestrial plants extracts for uranium sorption and characterization of potent phytoconstituents.

    Science.gov (United States)

    Sharma, Sunita; Singh, Bikram; Thulasidas, S K; Kulkarni, Madhuri J; Natarajan, V; Manchanda, Vijay K

    2016-01-01

    Sorption capacity of four plants (Funaria hygrometrica, Musa acuminata, Brassica juncea and Helianthus annuus) extracts/fractions for uranium, a radionuclide was investigated by EDXRF and tracer studies. The maximum sorption capacity, i.e., 100% (complete sorption) was observed in case of Musa acuminata extract and fractions. Carbohydrate, proteins, phenolics and flavonoids contents in the active fraction (having maximum sorption capacity) were also determined. Further purification of the most active fraction provided three pure molecules, mannitol, sorbitol and oxo-linked potassium oxalate. The characterization of isolated molecules was achieved by using FTIR, NMR, GC-MS, MS-MS, and by single crystal-XRD analysis. Of three molecules, oxo-linked potassium oxalate was observed to have 100% sorption activity. Possible binding mechanism of active molecule with the uranyl cation has been purposed. PMID:25946322

  8. Investigations of uraniumsilicide-based dispersion fuels for the use of low enrichment uranium (LEU) in research and test reactors

    International Nuclear Information System (INIS)

    The work presents at the outset, a review of the preparation and properties of uranium silicides (U3Si and U3Si2) in so far as these are relevant for their use as dispersants in research reactor fuels. The experimental work deals with the preparation and powder metallurgical processing of Al-clad miniature fuel element plates with U3Si- und U3Si-Al up to U-densities of 6.0 g U/cm3. The compatibility of these silicides with the Al-matrix under equilibrium conditions (873 K) and the influence of the reaction on the dimensional stability of the miniplates is described and discussed. (orig.)

  9. Preparation of uranium-plutonium carbide-based fuels simulating high burnup by carbothermic reduction and their properties

    International Nuclear Information System (INIS)

    Three types, hypostoichiometric, nearly stoichiometric and hyperstoichiometric, of uranium-plutonium carbide fuels simulating 10 at.% burnup were prepared by carbothermic reduction of oxide containing fission product elements. The carbides contained fission product phases such as the UMoC2 and the U2RuC2 type or the RECsub(1.5-2.0) phases (RE:rare earth). Composite theoretical densities of heterogenious carbides containing the UC, U2C3 type and fission product phases were calculated from the proportions and densities of these phases. By comparison of specific volume of the carbide between of 0 at.% and 10 at.% burnup, the solid fission product swelling rate of a carbide-based fuel was estimated to be 0.4-0.5 % per at.% burnup. (author)

  10. Speciation of uranium in environmental relevant compartments

    International Nuclear Information System (INIS)

    In the past, the chemistry of uranium was focused on its mining and milling for production of high pure uranium compounds as initial matter of reactor fuel elements for energy production and breeding of plutonium for weapons production. In this sense, the recovery of uranium and plutonium from the used reactor fuel elements was also technical realized. The increasing input of uranium into bio-sphere by mining and milling and industrial processes like production of cement, fossil fuels, and fertilizers has led to the realization of the importance of uranium environmental chemistry. For a better assessment of radiotoxicity and transport along the food chain knowledge about the chemistry of uranium is needed in all involved compartments. Starting from uranium content in geo- and bio-systems, about the determination of chemical behavior - the speciation of uranium - is reported in selected environmental compartments like seepage waters coming from mine tailings, different kinds of bacteria living in uranium contaminated soils, and relevant forage plants growing on these soils. For uranium speciation determination direct non-invasive methods are used like various laser spectroscopic methods, and X-ray absorption spectroscopy with synchrotron radiation source, The results obtained by spectroscopic methods showed that the speciation of uranium is dominated in surface waters by uranyl-carbonate complexes in opposite to the speciation in bacteria and plants. In these compartments the speciation is dominated by binding of uranium on carboxylic and phosphorous containing functional groups. It was shown, that in the investigated systems the speciation strongly depends on different physical chemical parameters like ionic strength, kind and amount of ligands, pH, Eh e.g. In experiments with living organisms it is necessary to characterize the state of the bio-system in dependence of the used parameters to compare the obtained results (ratio of dead or living cells of bacteria

  11. ATRIUMTM Fuel - Continuous Upgrading for High Duty BWR Plants

    International Nuclear Information System (INIS)

    AREVA NP is a supplier of nuclear fuel assemblies and associated core components to Boiling Water Reactors worldwide, representing today more than 60 000 fuel assemblies. Since first delivered in 1992, ATRIUMTM10 fuel assemblies have now been supplied to a total of 32 BWR plants in the US, Europe, and Asia resulting in an operating experience over 20 000 fuel assemblies. Among them, the latest versions are ATRIUMTM 10XP and ATRIUMTM 10XM fuel assemblies which have been delivered to several utilities worldwide. During six years of operation experience reaching a maximum fuel assembly burnup of 66 MWd/kgU, no fuel failure of ATRIUMTM 10XP/XM occurred. Regular upgrading of the fuel assemblies' reliability and performance has been made possible thanks to AREVA NP's continuous improvement process and the 'Zero tolerance for failure' program. In this frame, the in-core behavior follow-up, manufacturing experience feedback and customer expectations are the bases for setting improvement management objectives. As an example, most fuel rod failures observed in the past years resulted from debris fretting and Pellet Cladding Interaction (PCI) generally caused by Missing Pellet Surface. To address these issues, the development of the Improved FUELGUARDTM debris filter was initiated and completed while implementation of chamfered pellets and Cr doped fuel will address PCI aspects. In the case of fuel channel bow issue, efforts to ensure dimensional stability at high burnup levels and under challenging corrosion environments have been done resulting in material recommendations and process developments. All the described solutions will strongly support the INPO goal of 'Zero fuel failures by 2010'. In a longer perspective, the significant trend in nuclear fuel operation is to increase further the discharge burnup and/or to increase the reactor power output. In the majority of nuclear power plants worldwide, strong efforts in power up-rating were made and are still ongoing. Most

  12. Fuel staging tests at the Kymijaervi power plant

    International Nuclear Information System (INIS)

    The aim of this study was to measure nitrogen oxide (NOx) emissions and find the methods to reduce them in plants using coal and natural gas as fuel. The tests involved were made at the Kymijaervi Power Plant, Lahti, Finland. Coal and natural gas was used alone or mixed. With natural gas when using flue gas recirculation, the NOx emission level dropped from 330 mg/m3 down to 60 mg/m3. A negative side effect was that the flue gas temperature increased. At coal combustion and staged combustion, the flue gas recirculation had no significant effect on the NOx emission level. At coal combustion, the staging of combustion air halved the NOx emission but the combustibles increased strongly. With fuel staging, using coal as main fuel and gas as staging fuel, the NOx emission level was decreased from 340 mg/m3 to 170 mg/m3. At the same time the combustibles increased 2 %- units. Also the flue gas temperature increased a little. At the tests, the proportion of natural gas was rather high, one third of the fuel energy input, but it could not be decreased, because the gas flow ratio was already too low to ensure good mixing

  13. The Influences of Uranium Concentration and Polyvinyl Alcohol on the Quality UO2 Microsphere for Fuel of High Temperature Reactor

    International Nuclear Information System (INIS)

    The influences of uranium concentration and PVA on the quality of UO2 microspheres for fuel of high temperature reactor have been investigated. The UO2 particles were prepared by gel precipitation using internal gelation process. Uranyl nitrate solution containing uranium of 100 g/l was neutralized using NH4OH 1 M. The solution was changed into sol by adding 60 g PVA/l solution while stirred and heated up to 80 oC for 20 minutes. In order to find gels in spherical shape, the sol solution was dropped into 5 M NH4OH medium. The formed gels were small spheres, was washed, screened and heated up to 120 oC. After that, the gels were calcined at 800 oC for 4 hours, resulting in U3O8 spheres. The U3O8 particles were reduced using H2 gas in a N2 media at 800 oC for 4 hours, yielded in UO2 spheres. Using a similar procedure, the influence of uranium concentration of 150-250 g/l and PVA 40-80 g/l were studied. The qualities of UO2 particles were obtained by their physical properties, i.e. density, specific surface area, total volume of pores and pore radius using surface area meter and N2 gas used as absorbent, and the particle size was observed using optical microscope. The result showed that the changing of uranium and PVA concentrations on the internal gelation affected the density, specific surface area, total volume of pores and pore radius of UO2 particles. (author)

  14. The Effect of Wood Fuels on Power Plant Availability

    Energy Technology Data Exchange (ETDEWEB)

    Orjala, Markku (Markku.Orjala@vtt.fi); Kaerki, Janne; Vainikka, Pasi [VTT Processes, Jyvaeskylae (Finland)

    2003-11-01

    There is a growing international interest in utilising renewable fuels, also in multifuel applications. Main reasons for this are the objective to reduce CO{sub 2} emissions and meet emission limits for NO{sub x} and SO{sub 2}. On one hand cofiring, defined as simultaneous combustion of different fuels in the same boiler, provides an alternative to achieve emission reductions. This is not only accomplished by replacing fossil fuel with biomass, but also as a result of the interaction of fuel reactants of different origin (e.g. biomass vs. coal). On the other hand, utilisation of solid biofuels and wastes sets new demands for process control and boiler design, as well as for combustion technologies, fuel blend control and fuel handling systems. In the case of wood-based fuels this is because of their high reactivity, high moisture content and combustion residues' high alkaline metal content. Combustion and cofiring properties of fuels have been studied both in VTT Processes' test facilities and in industrial-scale power plant boilers. The formation of alkaline and chlorine compounds in biomass combustion and their effect on boiler fouling and corrosion have been monitored by temperature controlled deposit formation and material monitoring probes. Deposit formation monitoring at full-scale boilers provides unique information on the rate of deposit formation, the effect of sootblowing and consequent changes in heat transfer. Additionally, the data from deposit formation monitoring has been shown to correlate with boiler performance, which gives basis for studying the interrelation of: fuel blend characteristics; deposit formation; boiler performance. If biomass fuels are blended with coal or peat, following implications may be expected: increased rate of deposit formation, shorter sootblowing interval, cleaning of heat transfer surfaces in revisions may be required, bed material agglomeration (in fluidised beds), increased risk of corrosion, higher in

  15. Idaho Chemical Processing Plant spent fuel and waste management technology development program plan: 1994 Update

    International Nuclear Information System (INIS)

    The Department of Energy has received spent nuclear fuel (SNF) at the Idaho Chemical Processing Plant (ICPP) for interim storage since 1951 and reprocessing since 1953. Until April 1992, the major activity of the ICPP was the reprocessing of SNF to recover fissile uranium and the management of the resulting high-level wastes (HLW). In 1992, DOE chose to discontinue reprocessing SNF for uranium recovery and shifted its focus toward the continued safe management and disposition of SNF and radioactive wastes accumulated through reprocessing activities. Currently, 1.8 million gallons of radioactive liquid wastes (1.5 million gallons of radioactive sodium-bearing liquid wastes and 0.3 million gallons of high-level liquid waste), 3,800 cubic meters of calcine waste, and 289 metric tons heavy metal of SNF are in inventory at the ICPP. Disposal of SNF and high-level waste (HLW) is planned for a repository. Preparation of SNF, HLW, and other radioactive wastes for disposal may include mechanical, physical, and/or chemical processes. This plan outlines the program strategy of the ICPP spent Fuel and Waste Management Technology Development Program (SF ampersand WMTDP) to develop and demonstrate the technology required to ensure that SNF and radioactive waste will be properly stored and prepared for final disposal in accordance with regulatory drivers. This Plan presents a brief summary of each of the major elements of the SF ampersand WMTDP; identifies key program assumptions and their bases; and outlines the key activities and decisions that must be completed to identify, develop, demonstrate, and implement a process(es) that will properly prepare the SNF and radioactive wastes stored at the ICPP for safe and efficient interim storage and final disposal

  16. Internal dosimetric evaluation due to uranium aerosols

    International Nuclear Information System (INIS)

    The present work has like object to carry out the internal dosimetric evaluation to the occupationally exposed personnel, due to the inhalation of aerosols of natural uranium and enriched in the pilot plant of nuclear fuel production of the National Institute of Nuclear Research

  17. On the possibility of using uranium-beryllium oxide fuel in a VVER reactor

    Science.gov (United States)

    Kovalishin, A. A.; Prosyolkov, V. N.; Sidorenko, V. D.; Stogov, Yu. V.

    2014-12-01

    The possibility of using UO2-BeO fuel in a VVER reactor is considered with allowance for the thermophysical properties of this fuel. Neutron characteristics of VVER fuel assemblies with UO2-BeO fuel pellets are estimated.

  18. Determination of nitrogen in uranium-plutonium mixed oxide fuel by gas chromatography after fusion in an inert gas atmosphere

    International Nuclear Information System (INIS)

    A gas chromatographic technique has been developed for the determination of nitrogen in uranium-plutonium mixed oxide fuel after fusion in an inert gas atmosphere. When the sample and pure iron powder in a graphite crucible were heated to approximately 2500C by a resistance heating furnace, a large amount of carbon monoxide was evolved with a small amount of nitrogen and hydrogen. A gas chromatograph equipped with a pre-cut system was used for the separation of nitrogen from the carbon monoxide. Nitrogen separated by the gas chromatograph was determined by means of a thermal conductivity detector. Only 100mg of the sample was used, and the analysis requires about 10min. No specific skills for glove-box work are necessary. The relative standard deviation and detection limit (3σ-criterion) were less than 5% and 9μgg-1, respectively. The present method is not only applicable to the analysis of research samples but also to the quality control of uranium-plutonium mixed oxide fuel production lines

  19. Studies and research concerning BNFP: evaluation of spent-fuel-examination techniques for the Barnwell Nuclear Fuel Plant

    International Nuclear Information System (INIS)

    A study was made of various examinations which could be remotely performed on a production basis with spent fuel at the Barnwell Nuclear Fuel Plant (BNFP). These techniques could form an integral portion of fuel disassembly and canning operations. Their benefits accrue to either improved fuel storage, reprocessing, or both. In conjunctoin with these studies, evaluations have been made of the operational impact of receiving failed or canned fuel at the BNFP

  20. Issues and recommendations related to replacement of CFC-114 at the uranium enrichment gaseous diffusion plant

    International Nuclear Information System (INIS)

    The operating uranium enrichment gaseous diffusion plants (GDPs) in Portsmouth, Ohio and Paducah, Kentucky, which are operated for the United States Department for Energy by Martin Marietta Energy Systems (MMES), currently use a chlorofluorocarbon (CFC-114) as the primary process stream coolant. Due to recent legislation embodied in the Clean Air Act, the production of this and other related chlorofluorocarbons (CFCS) are to be phased out with no production occurring after 1995. Since the plants lose approximately 500,000 pounds per year of this process stream coolant through various leaks, the GDPs are faced with the challenge of identifying a replacement coolant that will allow continued operation of the plants. MMES formed the CFC Task Team to identify and solve the various problems associated with identifying and implementing a replacement coolant. This report includes a review of the work performed by the CFC Task Team, and recommendations that were formulated based on this review and upon original work. The topics covered include; identifying a replacement coolant, coolant leak detection and repair efforts, coolant safety concerns, coolant level sensors, regulatory issues, and an analytical decision analysis